ORNL-2387.txt

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C.84 —~ Reactors-Special Features of Aircraft Reactors

ORNL-2387, Parts 1-5
This document consists of 360 pages.
Copy/'l'y-:)f 273 copies. Series A,

~ Contract No. W-7405-eng-26

AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY PROGRESS REPORT

For Period Ending September 30, 1957

W..H. Jordan, Director
S. J. Cromer, Co-Director
A. J. Miller, Ass.i stant Director

DATE ISSUED

FEB 41958

0AK RIDGE NATIONAL LABORATORY
’ :Operated by
. UNION CARBIDE NUCLEAR COMPANY
DIVISIOI‘! of Union Carbide Corporation
. Post Office Box X
_ Ook R.dge, Tennessae

 

 

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S BERR L S A

i i ey Gl . sl bl i ki e

 

i s e

 

 

ORNL-528
ORNL-629

- ORNL.768

ORNL-858
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- ANP-60

ANP.65

ORNL-1154
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ORNL.-1227

" ORNL-1294

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ORNL-1896
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ORNL-2061
ORNL.-2106
ORNL-2157
ORNL-2221
ORNL-2274
ORNL-2340

 

Reports previously issued in this series are as follows:

Period Ending November 30, 1949
Period Ending February 28, 1950
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Period Ending December 10, 1950
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Period .Ern.c.ling June 10, 1952

Period Ending September 10, 1952
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Period Ending March 10, 1953
Period Ending June 10, 1953
Period Ending September 10, 1953
Period Ending December 10, 1953
Period Ending Mafch 10, 1954

" Period Ending June 10, 1954
~ Period Ending September 10, 1954

Period Ending December 10, 1954
Period Ending March 10, 1955
Period Ending June 10, 1955
Period Ending September 10, 1955
Period Ending December 10, 1955
Period Ending March 10, 1956
Period Ending June 10, 1956

 Period Ending September 10, 1956

Period Ending December 31, 1956
Period Ending March 31, 1957
Period Ending June 30, 1957

 

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C-84 ~ Reactors-Special Features of Aircraft Reactors

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ORNL.-2387, Parts 1-5

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.DIVISIOn of Reseorch cmd Development AEC, ORO

 

 

 
 

 

 

 

 

 

 

 
 

5

 

FOREWORD

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL
records the technical progress of the research on circulating-fuel reactors and other ANP
research at the Laboratory under its Contract W-7405-eng-26. The report is divided into
five major parts: 1. Aircraft Reactor Engineering, 2. Chemistry, 3. Metallurgy, 4. Radiation
Damage, and 5. Reactor Shielding. |

With the suspension of work on circulating-fuel reactors as of September 1957, program
emphasis will shift to research in support of the work of other organizations participating
in the national ANP effort. The major fields of the future ORNL effort will be research
studies of shielding, materials, and radiation damage and experimental investigations of

systems and components of power plants designed for the nuclear propulsion of aircraft.

 

 

 

 

 

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'CONTENTS

FOREWORD ...ttt eensisiacatese s rensnrste e sasssssasserstasssss nssssssssssossssmassnsssassassens st st sssssassesassansssssnsnseses vii
SUMMRY ............................................................................................... .‘...‘.....‘.,.'...' ........................................ xvii

1.1

1.2

PART 1. AIRCRAFT REACTOR'ENGINEERING

 

 

REACTOR AND FACILITY CONSTRUCTION........... reresiaeensaanss eetereeseee e st et bar b aRasnetrestontensaneans 3
Methods Development ........ccvvreeceninses cevnerannns veesarenen ravisresssasnsnenbiseantan asneta s brnasee s dent s ner e sansanestanearsrass 3
WEld T @St cccriirecrerieciriiecmnrmsesesscsnensstesesasnssssssssarsnsssssessasnsssineasass evenesteesnsetsresees pesvesessenessnsnennants 3
Equipment Inspection................. rreisesnnanasiirens veesenips st sasesi ssts s shn s ena s b e bR R s er e e sR bR st b r s 4
Component Fabrication and Assembly ........cccicciivninniccsenivenionnes eanereanenaranis resessssammensasens resseeessesesasne 5
Main Fuel-to-NaK Heat Exchanger........coueucunnccs ceuereranesaesessaeasesresnnensasarereres rreseverenerbeneseressrasans 5
Sodium-to-NaK Heat Exchangers ......... erevebreeranire e s b ase revesesaeseeaee s vaveransinsan reresersererseneanesnens -
ART-ETU Radiators......cuveervamrcvcorecnsns sterescreisonsisiennnsss tanan eeevassasesaiansresnaressesaesanasarersresnansaeasrase sanans S
Main Pressure Shetl and Liner ............ resserssanebivtpeenninneisinnanaataens sanes reneseinensiesatentasessbentsessessseantons 9
Beryllium Reflector-Moderator Outer Shell ...t 9
Boron Shield Containers — Shells IV and V ........... 9
Inner and Outer Core Shells ...ciinmiiiviiimiiimmsroensistaresinens reeebeasssereteresnsentasmoases 13
"Reactor North Head ...t pideesnenes veeresneressnsereanee e seastbers st bPRaOsSa L e Loy asen s e e et 13
Strut Load Ring.......... vorersaseserenmissonnsase ssernas revnrsssares e raseseresararasssnerasanan et anstnsens aeresenenas resseeenanas 15
Neutron Shielding .......cccoeeuureurenceusen. eesnensessspes eerermesreessness s et sanesRsranas vesesrsesesanes preseasessenestenns 16
Lead and Tungsten Shielding ......cccceuuusn. eessestnseninsssssnsreRest s R AR RO g AR R Sur s SrvSRS e radUS R SRR A SRS 18
Miscellaneous COMPONENntS ... iiirmmmismsmsiiessssmssmastanssistusssassessasssesssesss smssasssssasessasessens 19
Reactor Assembly .........cccceveivvcnnse estemenoasnsansensarieas ivesserreeinenrsuenss sesnrenertesnessnrresase tenistsbonsessstasnbestanees 21

- ETU Facility ......... donieienseseiisassibnnansia sares euiceessuisenissssnesidinseessusernssenen sire dosers seerespinasresessnsresssanasne senrs sasees 22
Reactor Support Structure .......... ikt usiesiai e seinsanniasu b sesan s SosssLan e aussesa s s RR LB s Re e b1 22

NaK Piping, Associated Supports, and Components ..... rsteseebessensesseeassrasse resmenaseanenneenesressnsnennive B
Electrical Equipment .....ciiivnviiccinvinininsn s innsiaseses mssssesaases reeeearteeterenresetantenassbere siet ot besbebattens 25
Auxiliary Services ......oocneennes Crsesnensnenssbane seperanriainER R RO 4 SRS NS KON SRS S AR IR LAS LSRR SRR SRR ARSI AR R SRS RS A 25
ART FQCiliy cuvveeereresmserassnssiesssassssssssmssssasessssssassesssnens reesertaraeraioseemitresstsonsestsinben nsisassiomase s sesaesssreeses 25
Design Activity...ciciinmcsiniinnee rerassssnsvenis seescesiasassises esmsisssnsrens Sestsumsensiasaisiosnatnsesninsiensirasssasansarenssns 25
CONSHUCHON PrOGIESS muviiiecsssiosiaivsmiseiusicssminsessissisminirssarssmmssssessssamesssansssssssrssisssssisssssssssasaisassans 27
Insta“ahon P!annmg................;.;..,_...‘;...;.,.',;'..,.;;.; ........ rersesivsssnssninens easssossassasens reseersessanisenssrsssninsres 27
ART DiSASSEMBY vovvuveresereervssasressserescsssissammsesssssssnisinssebens sssessssmsinsssiessssiisssmssiss eisssasemsisessssmasssssssonsoses 31
Reactor Disassembly e nainsimbinaesiasansbs Sssapsamasint nesats s s smistessiestesessasestsReIIASISY SRS SRS SRR SR CR LR bR - 31
Cutting Methods Evaluation ... cimiseniinssssssssens cersmssessosnsasmas sarasssasaneses wesesrrssessasinnes 31
Facfllfy Invesflgatlons eseiens ..... .............. esasiaeasaenrassaesrs sesiases 31
COMPONENT DEVELOPMENT AND TESTING serersie s siaebis s sEe s R RRs bR s e TR nes vemsenis 32
Pump Development TestS vmmrimininin eeseerdenrie s setn s s daege R iRt R beos SRR SR RSO RRS R PSR SRR sk s R SRR s 32

' Bearing, Seal, and Lubricant Tests.-’....;....,..;.;..';L ............. eerersai s seiaraittsessbaaseana et snsariasiseannnashens 32
— Aluminum North Head Water Tests . immiussmimmmssssimmmssisesinssimmmsosssassmiss it issmssssss W 32

~ Fuel Pump H|gh-Temperature-Performonce Tests eoviiodivieanintmes idhe s senaasebasnsas ssssne rerseesnrensrrseninsenson 34
 Fuel Pump Endurance Tests . ....coiminivinmmsiescsrsss iises et sespiesensassiaeseb s st sssassonseserenestResseaeen vienrans 35
ancry and Auxiliary NaK Pump Development.........; ........... verirener et ereasnssseses sasseseresas vessrsranasaes 35
Reactor Component Development Tests ......cccoviiiiiusininoiiinmin s sessssstssssssssmase snsaes 38

Heaf Exchanger and Radiator Development Tests ........................... 38

/4 '

 
 

 

 

 

 

Valve Development Tests .....cccveveeenrreeniivecnennes reierenceneeseeensensesseneerans 43
Sodium Circuit Water Flow Tests ... iiiiiininsientssninnesssssssssssmsssmsasssmmsressstnsorasisasasess 44
Outer Core Shell Thermal Stability Test.... ettt cstesncsssssenrennss 4D
Liquid-Metal-Vapor Condensers ......ccocveeececcivnnnncee. , ‘ Y 1.
Zirconium Fluoride Vapor Traps ...ceeoceveerrennccnrse e . - 48
ISIAND BelIOWS T@Stuuiiiiritirieieeiieteivesecesetetsecsasaesestsssasssbesssasnssassenssnsssosestan sssesssese s sansnesssass ssssns snsssas 48
Fuel Fill-and-Drain Tank Test.....coioicigiimimmsimssesiissssssamessissssgessssssessosssassassassssssmssansss 52
1.3. INSTRUMENT AND CONTROLS DEVELOPMENT ........... verseeessiseensrennaas cesesneees 54
ART Control Rod Drive Test. .. ettt s s sonsnss s ssssssnsss sessnese 54
Fuel-Expansion-Tank Liquid-Level Indicator. ... sencnnserssssensssensssssessasssessssons o 54
Bubbler Plugging Tests. ...t s s snscsctenn s sesassscnsoses senens ssesesssssesasssassassasass 54
Aluminum NorthAHeod Expansion Tank Tests woneirccrinciniinnnins 55
ON-Off LeVel Probes ... oeeeccierrereeecsenne cenesnsennsecansnsassnse ovsnsarssssensasosest sasssssasesnessansrsssnssnsssssessns senons 58
Magnetic: Flowmeters .................. eeestenestetesasestessrasaneeasste et eaesaenaensenee st rns sneaneeasanenns ' ..... 58
Liquid-Metal-Level Transducers............... reretetstesetesseenesasnee e ress b sensbaenes 59
ART ThermOCOUPIES «uvvereeeiriirsaaecireessesreesssmasssasessassassasssses ssssssssssassnsssras sstssassstessnse s ssens amansssansesssesssosns 61
Sheathed TherMOCOUPLIES ... et ceeteccecceeesecertesnesnecsssnsbassessesaranesesanonnensnns veererenes eeeninsennnnes 61
We Il ThermOCOUP ES o et te e tearrescree e snn e sasess e sseesasneesssssesessassasanenasasssaransases e 66
SUrface ThermOCOUPIES ... uiieeiciecccneietrtcreeetesteesaenesssasesessmasassessasantrsssnesassnsrsessassesnsssssntasassnss 68
Apparatus for Checking Thermocouples ........riveeeervnccrnnsnennas : 71
1.4. ENGINEERING DESIGN STUDIES ..covuceveeeitncnnennenesmsensssisessensensessns s sass s sssesesssssssssasssassssostasass 72
Applied Mechanics and Stress AnGlysis ...t reads s neaens 72
Thermal Stress Cycling Tests of Beryllium ...t ctscrissssssenassans 72
Thermal-Cycling Tests of a Welded Core Shell Model ...t 73
Stress Analysis of Island Bellows ... .ttt s rssae e e 73
Stress Analysis of Shield Support Structure ...................... 75
Stress Analysis of a Pressure-Measuring Instrument .........ocieviviiviiinininnicsmicsssissnncssssenseas 77
Apparatus for ETR In-Pile Tests of Moderator Materials .....c..ccveievevnnercrerccisennnans corvesranaseinsrarseasasns 78
1.5, DESIGN PHYSICS .....cooerreeteererenrneeesentensssesnenessssesssesssnsnssssseesesssssansesaessnssass sassessssssssssesassssssnsssasee e snsens - 80
ART Fill-and-Drain Tank Shielding ....cccccciiieiriine et ccescreecstestesnsseseanessasse snssnssessssssssnsesssssann 80
PeRetratIONS ..o.eeeeeecicceenereeereee s me s seaessesesesseesase s saesessssansasseseesassesssssesesenssseserssssssssasensessesen sanesensases 80
Overlap in Shield at Removable Top Section ........ccceovrrrecmeeerrererrinersnesressnnesens renerenesessrsnsasaenes 81
Incompleted Work ................. vrereessesessreeraiaeass eeevsesessseriseestasasssensessreesssassannnassnnssens sasssnn nnonanssnissasans 81
Dose Rate in Region of ART PUMPS ..cccceeerrreeninececcccnnenccncseessnne e ssenes erseseieeseneaeaeretatns saaseans 81
Decay of Gamma Activity in U235.Containing ART Fuel for Short Times . '
After Shutdown ........ eteerseeteestetsseaetesatareaaetetesenas s s ehebasa s bens Sese bR se st R bR ere s R arR s s eannrann eevesereseseesenenses sasenes 82
Decay-Gamma Heating in ART Fuel Drain Valves...... et eeneresesene rrererebesebeneseresrsrestrasrasiens e 82
L.6. MATERIALS AND COMPONENTS INSPECTION .......oovceerrerrrarreannnns restensnssissnaseesaessarassenanes eeisrenanenens 83
Material INSPECHiON ....cocvveerererecverncrreesrrnresmraeesvaseneresssessssessneass reveveresatersasasresneesare asresssperaneesesbnberaes N 83
TUDING oottt e sese s ar s ssne st e as s nsts e s s sasnstene s nsse sasrneresa st e s hn e saesesnanesben raninesaranens 83
Pipe, Plate, Sheet, Rod, and Fittings cc..ccerirrcrenenrencrreeeserssrsivanessesnsassesssesssesssnnssees rveenessenes 83
Embrittlement of Inconel by Penetrants ......cvoicioncrnrinirnnnrsessesensseseseessessesnsssssnoreinaseens 84
Chemical Analyses of Materials.....ovnininnininninns caessesases arnis IR ESORESUFSRUSORES 4 PO L AR SRS 0 renneaeeanes . 8
Electrical Calrods .......... ' ........................................................... cereersssaseans reeeeanesrepesaans 84

 

 

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Weld Inspection .......ccocveccueveenennenns Maeeesesesissesersresanares fereeersaseeeneareaneesnerenans eresasesimsrsssas sessstessisnsessasisansay

Inspection of Reactor Shells ..... eeestersesesaesesara s ettt et e saes st et et she s e e a s s R sebebeeReraeRbena e e arereseneane nareitseres
Inner Core Shells (Shell 1)..eiiiiviiciireee fetaresttet st te et et en sttt e rnser s beennt stk e rasbeua st sorace
Outer Core Shells (Shell 11 ..ot sesbrresmasa e sessenass sssssessisssuessans sereen
Shell Hl et vervensteneresaanaes Lt e e bt st sttt et s seressr s sabasesareseae reeuensieereeeaenens
Shells 1V and V............. reestea et e ensnans veerenneesnsaterans eerens feensetesaseinanantstes esanentesaeretsstesnrsananbereessntrees

Inspection of Components Received from Vendors......... S veeeereesarenes treeereesetessaeesrrestesserebesrae penessen
York Corp. Radiators and Radiator Materials...........cinieuesvecnirsieeenesenesesssessesesssessenessnees
Griscom-Russell Co. Heat Exchangers ................. reverasanesaene
Black, Sivalls & Bryson Heat EXchangers ...t cssnsnesssseseeenesesssnssnsersssses
Midwest Piping Company Forgings and Weldments .................... reereniesenestenressnans resreere e snaeae
Process Engineering, Inc., Tanks and Bellows Expansion Jomts......_....; ...................................
Fulton-Sylphon Company Bellows ......... eebeeesseereesasanotessteteee et rs st e e ba s eRbe s e s e s b bR e n s eusbeabes a0t bbae

- Hoke, Inc., Inconel and Stainless Steel Valves ...t

Metal ldentlflcahon MEter coveieeiereeecsieeereeeeeaesoien tretesieresaerersssterenaneneresssssarasttreren reeveneneeeeeseasaerasaessnsessas

Instrumentation and Controls Inspechons sebaisesansassresusestiastans e eeserbesRON e RO TSRO SRR SRS SRR RSO PR SO R RS O RS Re e b OO 0O
Nickel-Copper Welds in Liquid-Level Probes ...........ccommveinicrnrcincnr e sreneesissens s
- Thermocouple Welds...........c..coee.ee. vedvaeeseseratteesresReranesr et eaes ot tanas e At et eeaseshes aenr e se bt e seraransans nries

Procedures for Radiographic |nspechon of Weldments on ART lIsland and -
REFIEEION woriiieierreeircrercre sttt redeseiaimn e e e ses sassnssesaresssbes e astnssrssesensessassansas teverereeestensentenerenesnsenens

1.7. HEAT TRANSFER STUDIES......coiinercenpensicsnessscssenssssseesenssenaas reess sttt ar e e s s s rse s asaes

Thermal-Cycling Research.......ccccveerecvevievnnnenes eeveneeseerirassereesnenaraisasaseest sens s anetes uesaasternsesasasssrasassrananes
Pressurized System ......... veereriueeeteerenane sessaenes veveresaresasnnstesnsssen ieesrsess b bbb s e tene
Pulse-Pump System ......ccovervrnene, ceavesesnessssssasansene Fesspersesnesninsareresssnenasesess beresaserssansessrtsessnsnssnnenens
Thermocouple Development......_.....-. ....... eereraeeraeserasnnsaen eterisaisseniesraar besase st seaaaste s se et e et e e e et ane

ART Hydrodynamics c.ueerermnssssesesisessesesssssnessssmsnnns cres sttt s s s st et s banesa e nes renrerorisererseenss
Full-Scale Core Studies..........conuen... deeneeesertsssrenasaises bnnestissensrasne satassesinares asenaratensesseseesueanesesens s sueren

- QUArtEr-SCale Core STUAIES ... vt esisesssseesare sereesasabesssness e sasssasastsesassnensensaneesssss
Instantaneous Velocity Profile Measurements ...........ccoreieriiivinveiceseevenvnnnneenssseriseessvessnmene sevese

Fused Salt Heat Transfer .. iciinniarivenee rsisesscssesssssesstesesorns eveeresssnesesesessesnnn reevesseetaessrenssenrenssesens

Heat Transfer Experiments .......cooccivrimnionnecsisesensenes vevereereresnrssnenrarenasaneres vetreserer s s s st
ART-Type Core with Screens ...... eeeensavanas sreeesiesrtsenearerstan et ne e ser st st e antson bas s res s aSRT Rt RS e e R bes
Vortex Tube...cccevrnrranene vorresrran vevaeiesanatens \ibdeberineesires saasusre sanineneaest sesioenen teeenerersentiesra e et e nrasenes

Liquid-Metal Volume-Heaf-Source Expenmenf ceitessii e ri s sassrsaneee cerrrteseseamenns erereeserenesnsraenees

Mass Transfer .t rrensnsasarsntnasens versieassar s asasssas sresans

Physncal Properties ..........ummermmmssmesisens sersesbemsiensenstasas st e e iR e s ssabene terreessenseseass veeiesasesnnasyonsessaseas
"Enthalpy and Heat Capacnty vere s rreneaaerneas veeveuiresnereasessbesseninai aseareniensasrrrnne asensseroras s ansansaseasen
Thermal Conductivity.. . uummmmsiviicrismnmssissmsmssssssnsssssss e vertoraeuss saosssente sesesraessesesbeseatsanens vovones
Propemes of Zirconium Fluorlde Vapor Deposifs ................................. cersearsanessasersennensrenes -

PART 2. CH EMlSTRY

2.1, PHASE EQUILIBRIUM STUDIES .c.ocercorsncssissmstmssmsissrsssisesiossssiesmsscs |
The System KF-UF , .......cco.c... eeeeesririneeen etae e AR 818 R A SRRk AR SRR
The System NaF HIF ; oonre it veeseons

 

 

 

93
96

96
96

xi

 
 

 

 

2.2

2.3.

2.4.

2.5.

xii

 

 

 

The System KBF (sNaBF | w.c.oucenieec i,
YHrium FIluoride Systems ... iicicreeeccrctcenentetornnsssmststnssessassssstossssisssssssnssssessmsisssorssnsassaseases

CHEMICAL REACTIONS IN MOLTEN SALTS ....conminnrimcsenenisans et senae bt sare R e R sesstassensessaens
Equilibrium Reduction of NiF, by H,, in NaF<ZrF ..o remreieseeranasane
Activity Coefficients of NiF, in Molten NoF-ZrF , Solutions einertessenesasstessseenssene eererasarnsanssrnsnesneene
Thermodynamic Considerations Relating to the: Acnwty OF NiF 5 oo tincsncsecsseneanns
Reduction of UF , by Structural Metals ...t resersesararsrsnsassn
Stability of Chromlum Fluorides in NGF-LiF-KF ..ot
Reduction of FeF, by Cr° in RbF-ZrF, reeentenasnanans eeersesmesensasntseane s ssasanans
Solubility of Structural Metal Fluorides in RbF-ZrF oot
Activities in Alloys ... rcncieneccinecnnercsnseessanssss s sasssssasess

 

 

Emf Measurements in Molten Salts .... reeesesesesranesareseaenssaete e aresssaranbe beebereseens oensenessennnranas
Daniell Ce|_|s in Fluoride Melts 7 rereerementeserebeeetessttenehENatseIeNT RS seta ennessrans .
Emf Measurements of Chloride Melts...covvcnerinneeriecsccnnnn reveasan

Solublhty of Hehum, Argon, and Xenon Gases in NaF-ZrF !
Solublhty of HF in NaF-ZfF Mlxtures vivevessases wsaessssesenssessaisssnassnsensanns eesstessuseseseriersasessmtsesiesssasaensre
Solubility of HF in NaF-KF- LIF MEXBUPES coccecienecirecccnerece st nrcaseasesesseeasnesnnsesenssssnassensesnsasassesesasanns
Solubility of CeF ; as a Function of Solvent Composition .......cocccrsvscecrcnnene. ereeseseeseneasesserenesesenns
Precipitation of Cerlum as Ce, 0 from Molten LiF-KF Eutechc ....................................................
Stability of UO,F , in Fused Salts e i as e s R e eenes S

Oxidation of Mixtures Composed of Sodium and PotasSium .......cceeeveeeeeereenesresenenensnnecesesssssinesens

PHYSICAL PROPERTIES OF MOLTEN MATERIALS ...ttt in s
Yapor Pressure of HfF4 eeves rersaseesesensaseneeses eeeteeesesbees st st ten e st eae e e see s e e e e ade e sema tmmeanena s s es s sases

 

 

 

 

 

Vapor Pressure of BeF ...l eseetesaeesteseatesassasesesteaesarsesatesae st naseeersesesnsmreeras reeernnenersnenen
PRODUCTION OF PURIFIED MIXTURES ..oooeroeieeetiveeviniseneesessesessensssesssssassnarssssnnresteesmessonsorsesssssnens

Preparation of Various Special Flucrides ............... evereressseseressesaseraartestaeea bR e anae e sarmasas e srasane eveae
Preparation of YF 4. S Verereviertessesteseastsssestessertasensrenntresasasesnarenests
Preparation of Other Fluoride s ..o setccreeresestesseeseeseesssnsasssnsssssssnessssssssnnssennns

Production and Dispensing of Fused Salts and qumd Metals ..ot e
Expenmentol Batch Production ...ttt crresnes erresse s ceeseasessans ereeseeereesneeesrannans
Dispensing and Servicing Operations ... rircrvicnnionneni s csesssssesssens eerenmmennes
Materials Available ..o enenirasasraerenssrranenssanerass renesnsnrensresnssanas

ANALYTICAL CHEMISTRY .o restereseneasseraenrasssererenertneste sanersnssesans rensernneenerneearasnrnenserons
Determination of Traces of NaK in Air ......... tiereestseeinnesansssersnissatesssrsisneiens sreseedaensenesasesnensnens desemmrassans
Determination of Oxygen in Metallic Lithium............ euessareneteaeresraeatsteasaraeaeea snes aertseetunersnastsartsaseses
Determination of Oxygen in Fluoride Salts and in Metals ....................................................... S
Determination of Nickel in Metallic Sodium ......cccnvreicninnininnniniisicinnecnas e resersssnens eenarne

Formation of Carbides by the Reaction of Pump Lubricants and NaK at
Elevated TempPeratures ... inenresmserenmersssstsnsssonmsesssnsssssssssessesssssassas sassassrsasssans srasassssonsasians

Determination of Aluminum in Mixtures of FIuoride SAlts ..o eeereceeceieereeseresereessessesessessseesessssnen

 

 
 

 

 

 

 

oA

3.2.

3.3,

 

 

 

Solvent Extraction of Titanium and Nloblum from Acnduc Soluhons with -

- Tri-n-octylphosphine Oxide......... reaeenerurasseses aiuspesessasivevaniosyasusarens feeserensnersonssstesasiaestessanesanssnnses sesnenanss 156

~ Extraction of Sulfuric Acid .c.vceveeerecnreeaensens R e nane R eerernesses s steses SO 156
Extraction of Titanium. ... it rasisss et sonsossisssn sesssisscs sosssiosasmsssssssreses earernen 156
Extraction of Nicbium ................ rresansrovianesnseenesives erreringansie s drsnrit brensnensesiusiesasianttsmess asensesasarasastbes 157
Installation of a Microbalance in'a Vacuum Drybox.......; .................................. crmeseraess s rsiesaasssaen 157

PART 3 METALLURGY

) NICKEL-MOLYBDENUM ALLOY DEVELOPMENT STUDlES .............. 161
Material Development .........cceiioveeeniiinenn. ......... eeereeeraseeseraraenes 161
Properties of INOR-8 rinemseresiydenien .. 162
Properties of INOR-9 .........ccccceun. Fbasseunssuessinssiinsssrnst s s et deRiRra RO r ke eSS b arins St arasan s s sassanataas 162
- Studies at University of Tennessee...,..'-_..'.5...‘.....‘..;'..-‘.-...._...;....._.................4.' ........ eserer e esassas e enne 163
Studies at Battelle Memorial InSHtue ..ot snncaniosinsansiaisessssenssssesesns rerenteerenienrieesessnesasaennes 163
Studies of StresssRupture Properties at New England Muterluls chorctory.........................;.. 163
INOR-8 Alloys Produced by Superlor Tube Compqny et erisreisshes SRRt s kb ass st b s sessan s 163
Status of Production Heats ceersrrinsnyeres easeentaenisu e esast e phrendisnsasnsunsases preesniassanssseassosnssssnassseseres 164
. Compos: te Tubnng...._;..,.., ...... cevreen ..... . ....... , ......................... 167
- Jonnmg Expenmental INOR- 8 Alloys ,....‘ ................... 168
Stress-Rupture Propernes of Nlckel Molybdenum Bose Alloys ina Fuel | '
Envnronment....;..,.._..._ ......... diesseienntagianenes seeneiasnesiens fopaseishasberanesrstasnsesptiases bonogscnensiasesoses e taesasass 171
~ Corrosion Studies ........c..... ........ .................. 176
Forced-Circulation Loop Test of a Nlckel-Molybdenum Bcse Alloy o
Exposed to Fuel 107........ccou.... ieviiuayaieninsranentionesentabasnesenbbede asseranaeg sers sensaesase eeeresresenesesaenesbnnans 176
Thermal-Convection Loop Tests of Nlckel-Molybdenum Base Alloys |
Exposed to Fuel 107 .o.iviiimmiiiviiomessmsassisssins ieeieriusnraes et ssr sy bess s dn e aee e sasnsasssr s s bosts 179
Compai'lblhty of Nlckel-Mo|ybdenum Base A“oys w:fh Molybdenum resstrse s ese st 181
DEVELOPMENT STUDIES OF INCONEL vrvciovorrcissiomesesssssississmesssssssesssss e 182
Strain-Cycling and’ Stress-Relaxahon Studses.....;.‘..‘...r;.».’..v....'.':...'...........,". ....... eueeserenssesss sttt saras 182
Biaxial Creep Studles ........ e eeeemssues S S vieae e aone e e Ao epip e et etmeen e e ettt s sra SRS Rr0S 185
Relative Tensnle Properhes of 1nconel Plute and Inconel Weid Deposuts s 192
_ Brozmg Alloys for Inconel Jo:nts...,.7'._-.';.;.;"._.;.'i;;..;_;.;.,-;.,:'...'.;a'.,..,.f....g....'.',.,i...._.,..;........; ........... ceasesesrsiseniaans 196
 OXIAAHON STUAIES ..coirirerismisiens onsiesisssiesiitisiriunins s st ssssissasssisesssssasisssesassssstsssssssrass st ass snsssssssons 196
: Corroslon Studles .... ..... S 196
'Effecf of Gram Slze on Corroslon of Inconel by Fuel 30 iceresiin, et ...... eeviensensene 199
,WELDING AND BRAZING STUDIES ...... s e enrnsirsssnes 202
Exammahon of All-Welded inconel lmpeller That ercked in Serv:ce eveiensdeters ............... - 202
. Brazed-Jomt Crackmg Tests....Q.."Q..Q..’...;;...;.';:';;.;.'.'.'.-."__.'.’-..‘.;..’..;......,i ..... , 202
' Fabncohon of ART Fuel Fnll-and-Drcm Tanksv..:..'fi..f....;';.,;.L,a;;:.;'.L.;;...~;;;';';..;;'.;..;.;.;.;r.,..;..,..a..-._; ...... revasons oo 204
IR Correlahon of Rad:ographlc ond Metallcgraphtc Determmcflons of Porosnty S
in Tube-to-Heoder Welds ..ot nniannes, e il st e e s R s 207
Fabrication of Sodwm-to—NuK Heot Exchangers ,.. ............................ 208
lnveshgctlon of Material for Radiator Headers.;..,..'...'..'......'..... . ..... e nien s s bR ceean 208
Fabrication of a Small Semucnrcu|ar Heaf Exchonger.....,.-....-;....;....; ................ reresvereeniestennsresnearesansteas 209

 

Xiki

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3.4. CORROSION AND MASS TRANSFER STUDIES ................ reeeesterestesssassesasrensanrereenastnes rvereneeresaiesesaenians - 215
Corrosion of Yanadium in Fluoride Fuel and in Lithium. .. i e esesienssenssnessssis assessns avees 215
Molybdenum and Niobium in Contact with Lithium...ocoececcucucenveircecnnnnen. R ...... 215
Tungsten-Nickel<Copper AHoy in SOGIUM werrvoeevooreeersreseeeesssssesss s coeesesssseesseesssesesssesse i 216
Tungsten Carbide- and Titanium Carbide-Base Cermets in Sodium......c.ccinivccnivnnvcvsiciinnsncinee, . 218
Experimental Studies with Molten Lithium ... ccesstrsnesscesne e sevasesesesesssssnoss 219
Inert Atmosphere Chamber for Thermal-Convechon Loop Testing of

Nonoxidation-Resistant Materials ... iivicininincicnncnnenesnenccssiesosens cereerisasraseesseinnssssassssesss 220

Forced-Circulation Loop Tests of Sodnum in Hastelloys, Incone! and ‘
Stainless Steel ........orrecccinnenrneneeenans estemereneesans rasteeseenese e aas e et rase R e ee R s st st s e b e e e etessesbesstan 222
| Diffusion of Nickel in Liquid Lead w..ueemerrmmessrsrimsstssssissssrmseesstesenes st 224
PUFIfiCation of SOGIUM w..eeeeeeeeeneeceeeemreesssenesesesessosessseerasesennns rreseessreraeseseneserass i serasase sorasa st e smannass nnanns 225
Mass Transfer in AQUEOUS SOIUIONS .....ccieecerervennnisieessssssessisssssesssessssssssssasieessanssisssossasssensssnesssssseses 225

3.5. MATERIALS FABRICATION RESEARCH.............. e en s s s st 228
Yttrium Mefal PrOdUCHION oot a s e seesaan s ses s s e sares st s snsnansatassesns saseasesninnasenesanestanenan 228
Hydrogen. Difquibn Th'rough MELALS ..o cciiinec et te s ce e rrren e sae s es s ars e se s e e s s s ere e nanana s st nens 228
Volatility of BeO in Steam at Moderator Temperatures ..........cccvccveccncnienecesuenenenssnsessoresessmensascssses 229

; Fabrication of High-Density BeO ...t insnnsnsessessssssnsssonsesssssesssisass 230
BeO Thermocouple Inserts and lrradiation Test Specimens .......cccecvervrrvnrrrnrnrecnreeccnsences eeerereens 231
~ Thermal Stress Resistance of Ceramic Cylinders ettt st e 231
Fabncahon of Metal Hydrldes ................................................................................................................ 231
3.6. METALLOGRAPHIC EXAMINATIONS OF ENGINEE RING TEST COMPONENTS
| AFTER SERVICE......o oo cetteerertrctrtessesseressssessasesasassssessassssesmesssessssnsss sstenere sssesssssobesnssesensossenssssme anes 232
ART Prototype Test RAIGIOF ...cu.eeeiiiireiciireneinneesnesieicsnenesss st st ptesesassssssesssessass et sasss sssssssnnmenssisssons 232
York Corp. Radiator No. 16 ... icivciciinnrecreienctenernesssssessesnesssssssssenssresssesesssssssenssssnssnssssssassassasn . 235
Process Engineering Corp. Heat Exchanger No. 3, Type SHE-7 ...t 238
Process Engineering Corp. Heat Exchanger No. 2, Type SHE7 cctirrrininrcnermisnssesnsnsssassasnanes 238
Inconel Centrifugal Pump That Circulated NaK ......oceeeeincvcirnirnrincicicsicnssssinssnisneenee. - 241
Welds of Test Components ................. 241
Thermal-Convection Loop Welds ...t srnsenens 241
Forced-Circulation Loop Welds ... ans s 242
Heat Exchanger Welds ...ttt et 242

3.7. NONDESTRUCTIVE TESTING ..o ceccrtrncrsscer e sasasieesasssnssnessaans Svevesaeseseseessetesrartesnanssntesenesaress 247
Eddy-Current Measurements of Metal Thickness ......ccovoveviiniicnnininccnnccnsicnnenene cenvnsineeniens 247
Mathematical Analysis of E ddy-Current Probe Coils.......c.coivivrinicinniincccsinncnnaeens revanes 247
INSPEction Of TuUBING. . ittt e crertsesees s ssasianssanesasaestssssssssesesisssasssensnensssssssssmssssssonsss SAT
INSPECHION OF PiPe .cciiiriiiiiirreieirencesssestreesenesontssess s inssnsse s sebesaseses et s sesesrsassessrassssasess sanserbesess suinesntsanssren 250
Plate INSPECHON .oucveveeriririnscrrsstses e resssttete e asis st sasaesisesssassss sveasabsssssssnesass nsssesesassasnsion stanaanes weenes 250
0re Shell INSPECHON e veeeeeeeeeeeeeeesioresesesessesesesssessssstts sesssssresssss sabesesssasssssssssssassssssssaesessaossssnssassansesens 250
The Detection of Unbonded Areas in Plates with Resoncrlc.::“l_l'ltrasound cierreeereresarssenmmasnteneanaensene 250

xiv W‘

 

! g

 
 

 

 

4.1.

| PART 4. RADIATION DAMAGE
RADIATION DAMAGE oot eseeeeeeoeeseeeses s e 257

Examination of Irradiated Components and Mafenols reerneererarenesteteebersensenebesasererasans rreneraenneesssenennnns 257
MTR In-Pile Loops........ erveneens eteriesiee e aetereaaebastestebbe el esaers et ses s sesreraesebasasebes erereseeseensereressaesaseas 257
Moderator MGteridls ...t ss e esesnese s sesesessenssesasserssnssessesens eararasenanes . 257
ARE SPeCIMENS et st civsrseeessssssss e essses s ess st sbe st es s ee e aseese s resererenrenanraenenes 262

Creep and Stress-Rupture Tests of Inconel ....................... bttt reesasrsesanastssasssasaraens 266
MTR Experiments .................. S hsnnenresnsrasranned reeesiresrseeensaebestsasaseraansarssasse eereereetnebaesesee sasarane 266
LITR Experiments .....covciviiciinmeiinsinsinnensiiinnsssnsennnsen, crerereenines reveienrerenesasaete erenernnans rerserenesans 270

LITR Vertical In-Pile Loop.......ic.cooiieriveeerees ereetisndiivseniomrens iessncrsnns eessererenseaeenssessssa s saesasssae seseennass 270

MTR Statlc Corrosion Tests .ooeeeeicserssessens mevsssssssasnn osasi rvetereesneene rerererseassstoesrsesnensaaneraesestrsetaerantser 272

Flux Momfcrlng of MTR Static Corrosuon Capsu|es.................................». .......................................... 272

‘Relative Contribution from the Ni®%(,)Co%? Reaction to the Co‘0 :

Disintegration Rate in Irradiated Inconel sesbessesssidan renpinrrhsnenedenaiisnerisnssnsseanrinsatentsarasansnsararsasisoser on 272
ETR Irradiation of Moderator Materlals ensr s anatsss e ittt s ass s e anens ceresrencanienes 275
Effect of Irradiation on Thermal Neutron Shleld Mc:tencls veevesesrennaens eesebessesenssasnens TPRTURRRN 275

BOoron Nitride o vururrnirirssesssssissssssesissesssnsnsssissasness srnmesieressssansensabagssmanenas dorasteresnesnastasremasstsaneass 275
. Boron Carbide w.....civermerrnsinnns. ......... feemsessenansaviasnesarcannsanerssasessrns evesesressrenennns 277
Cermets ...cccpeeeeeeene etrernnee e s et sesensasen shes sareson et e s b e R e e e b e w 277
o 'vEffects of Radlotlon on. Electronlc Components.......-s ........... veseessaesesneraseseshessaessessansessonduoeraessnntosnansasnne 284
Experimental Results......ccouviiviiionsiiineiinnna, trasenseniasaisansinsrenies cernanaes sesiparesasnaiesns sresasssesestususease 284

5L

5.2.

5.3.

Experimental Reflnemenfs ninissrpeisiesiiasssnsssnurinsios dibipioreeey e e e R s s a R R e R R 292

PART 5 REACTOR SHIELDING :

"'L!D TANK SHIELDING FACILITY ........... s 297

| Radlcmon Aflenuaflon Meusurernents in Plam Water, Borated Water, and Oll i e 297
Study of Advanced Shleldlng Materlals ,,. ............................. 300
‘BULK SHIELD!NG FACILITY ..................... 303

| The Suppress:on of Capture Gamma Rays by L.lthwm in. Leod......’7.‘...._,.,....;.........._; ................ csssrsartenes 303
_‘-TOWER SHIELDING REACTOR-!I ...'_.i_Z.'.._..'....'(-.‘.",;_‘.’_;..,..;.;.;...,'.;,;...;' ........ s s R 304
Mechcmcal Desngn ........-.,;...’;;,...-._..,.._...-.,.A..,...,.._..,‘.v.._;-..w.x_.'..,.'l.jr..,‘.,....,_..,-.,., ........ ereesarsrssseansiannn cvrehernsesssensranasensesersecs 304

- Nuclear Calculchons ..;........,».-'..’.‘;;......;.;.,,'...._.'.~....'.".‘..';.-..= ...... ereressisanenastseisenetsisontarsrteiterenes corersireasssisisans w304

- Fuel Element Develooment terers serens e e e e e smasnnt s iinsnissesrsessrarisasdsianesas veseeneees 304

| Developmenf of the Control Dewce et b ssiesep st srasesaerassaranes resssnessatinsessasaenesanronasnssesensenersstantssssants 304
o Calculation of the Gamma-Ray Heahng in the: Lead Shleld resrise R AR R RR SR b RS 1 307
" Heating at the Core-Shield Interface ......coviumrmcrenirsins eessessansaiunsesiessasmtesestssensasesntensasasares sussnsasns 308

~ Heating Within the Shield Lurustsmmsnsssesarivivh e ssbesiessasesare s esasssainsenrassssepnenseRmonpassaneness restnieaisersssnstbainins 312

- ConclUSIONS cuvnicvvesessssssrisnsenns wbemaeanrnens devssistssstassasseneresstesaronsesanses erss s s a e seranses 315

 

Xy

 

 
 

 

 

 

e alherFa R T WD
Ly ket

  

5.4, SHIELD DESIGN ......oouootitireeerreerecesteeensssetsrsssesassesssmsessaiesessasss st sassesssens sassssesesesesassesessatstssssssssssacsessesnsses
Meonte Carlo Calculation of Gamma-Ray Penetration of Aluminum Slabs ...covcevcevnncireene
Monte Carlo Calculation of the Gamma-Ray Penetration of Lead-Water Slabs..................... N
Monte Carlo Calculation of Gamma-Ray Dose Rate Buildup Factors for Lead .

AN Water Shields ...t e e s e E e s s

 
 

 

 

 

ANP PROJECT QUARTERLY PROGRESS REPORT

SUMMARY

" During preparation of the reporfs given here
of work on the Aircraft Reactor Test (ART) and
Engineering Test Unit (ETU), it was onnounced

" by the AEC that work on the ART and ETU was

to be suspended. The' necessary steps “have since
been taken ‘to place all fabrication and con-
struction work in a standby condition. The reports

- presented here have not been modified to reflect

the work stoppage. The next report in this series
will summarize the work on the ART and the

ETU and give a picture of the status of the

molten fluoride reactor program at the time it
was suspended.

PART 1. AIRCRAFT REACTOR ENGlNEERING

1.1. Reactor and Fcclllty Construction

Methods for assembly and inspection of the

ETU and the ART were studied further, and the
necessary special checking fixtures and in-
spection tools were designed and fabricated.
Component fabrication and assembly continued, and
an experimental main: fuel-to-NaK heat exchdnger

was received. ' The completion of this unit dem-

onstrated the feasibility of manufacture of the

‘heat exchanger. The two sodium-to-NaK heat.

exchangers received previously were installed in
the ETU north head. Modifications that will
facilitate installation are being incorporated in
the ART units. The NaK-to-air radiators required

for the ETU were received and installed, and
production’ of the ART. radiators is “under way.
Rings of. brazing mater:ol to be used in the
'_fabrlcatlon ‘of these units were produced at ORNL o
ot arate of lOOOOperhour. R L

A ptessure shell lower forgmg was recewed _";.’ i
and: the - ‘upper pressure’ shell forgmg is nearly',fi
“complete.. -The design of the pressure’ shell lmer—f :
 was completed, and: preparahons for forgtng and
" machining are under way. The beryllium reflecfor-: o
. moderator outer: shell is’ belng fabricated from.
a ‘machined’ weldment. ‘As d result of recent exs
- _penments, ‘the method for manufucturmg the: ‘boron- -
.shleld-contummg shells -has been" chunged from
“shear spinning to deep drawing. The core shells

for the ETU have been completed, as previously
reported, and machining operations are under way
on another set of core shell weldments.

'

/J

Assembly of the ETU north head is well under
way, and the retort for stress-relieving the assem-
blies was completed and tested. The strut load
ring assembly was essentially completed.

_All .the boron carbide tiles for neutron shielding
are on hand and the cons and lids for the tiles
are being fabricated. Work is proceeding satis-
factorily on the production of the required stain-
less-steel-clad copper-B,C cermets. Design work

_is proceeding on the shielding external to the
- pressure shell,

Fabrication .and procurement of auxiliary equip-
ment are well under way. Layouts of the equipment
to be installed in the ETU cell are about 75%
complete. Installation work on the ETU facility
is progressing as scheduled, and many of the
major components are now available for instal-
lation.

Design work continued on components outside
the cell of the ART facility. Models were con-
structed of the special equipment room, radiator
pit, ~and radiator penthouse area to assist in
detail design and to demonstrate equipment,

_assembly, ond changeout procedures. All lump

sum contract work on the facility has been com-
pleted.

Developmental work on methods and equipment
for ART disassembly was continued. A full-scale
experiment and demonstration of an optical method

for taking measurements through hot cell windows

is being prepared. Additiona! cutting tools were

.. tested, and a list of tools and fixtures that will

be required for removul of the ART from the cell

o was. prepured

.2. Component Development nnd Tesiing

The design reactor sodium pump speed was

_';_fr:mcreused from 3000 to 4000 rpm because of an
. ‘incredase “in. the heud requtred ln a 1000-hr test
“run at’ the hsgher speed, it ' was found that seal

leakage of the rotary cssembly was never greater

,f"ithan at the lower speed. Qil sparging system
tests “showed  that. with" a: “lower shaft annulus

helium - flow of 500 hters/day and a low oil

| ‘|eakage rate, satisfactory sparging could be

achieved with intermittent gas flow. A sparging
procedure for ART and ETU operation is being
prepared '

 
 

 

 

 

d

 

The reactor fuel pump rotary element that was
aitered for irradiation testing was installed in
a gamma-irradiation facility in the MTR canal.
The test is under way with the rotary element
operating ot a speed of 2700 rpm. The average
dose rate is 106 rep/h.

Additional tests of the twin fuel pumps installed
in the aluminum north head have demonstrated
satisfactory degassing characteristics over a
wide range of operating conditions. Installation
of the twin sodium pump loop in the aluminum
north head was completed and tests were initiated.
Preliminary data indicated satisfactory performance.

Additional performance data were obtained for
a fuel pump operating at high temperatures with
the fuel mixture NaF-ZrF ,-UF, (50-46-4 mole %,
fuel 30) as the circulated fluid. Curves showing
the minimum surge tank gas pressure and minimum
suction vs flow for cavitation-free operation were
plotted. In priming tests it was found that the
pump primed and gave full head and flow per-
formance when the liquid level was above the
floor of the expansion tank.

Examination of o primary NaK pump that was
stopped when the NaK level in the pump tank
rose and flooded the oil catch basin revealed
that grease formed by the interaction of NaK
and oil had clogged the seal loading springs
and inhibited their operation. Further development
work on the seal region of this pump is planned.

As a result of experience with NaK pump hot
test loops, modifications of the cold traps were
made. A nozzle was devised to admit a controlled
water-air mixture to the cooling coil and thus
effect a gradual transfer of the cooling load
from air to water. Tests of auxiliory NeK pumps
were initiated.

A semicircular heat exchanger designed to
simulate the tube stresses in the ART main
heat exchanger is being installed for testing,
as is a 25-tube hect exchanger that is to be
tested with the fuel mixture NaoF-ZrF, -UF,
(56-39-5 mole %, fuel 70). A test of a 100-tube
heat exchanger was terminated because of a
failure of the furnace inlet line. This heat ex-
changer had experienced 168 thermal cycles in
421 hr of high-temperature operation, ond there
were no evidences of failure. A first-approximation
thermal-stress calculation had predicted a life
of 35 to 45 cycles for this heat exchanger.

York Corp. ART test radiator Ne. 1, which
developed a leak after 870 hr of operation was

 

N L mam e
e sa ey
o et = .
. - .
oLt

examined. Despite immediate shutdown to limit
damage by fire, the damage was sufficient to
destroy evidence that might have indicated the
nature of the failure.

The excellent performance of the cold traps
in heat exchanger and radiator test systems has

led to the elimination of plug indicators from

the ART and ETU NaK systems.

Further tests of the fuel dump valve designated
ORNL-1 indicate that the valve will be satis-
factory for the intended service at 1300°F. As
a result of further operating difficulties, the NaK
dump valve has been redesigned to minimize weld
and thermal stress distortions. .

Sodium circuit water flow tests were completed
during the quarter. In these tests the proper size
for the orifice in the control roed sodium cooling
passage circuit was determined. Also, tests were
run to establish the proper location of o flow
divider for directing sodium downward into the
reflector and upward to the island entrance region.

The second test of a one-fourth-scale outer
core shell model for determining dimensional
stability under thermal cycling conditions was
terminated after 339 cycles, that is, 39 cycles
more than scheduled. Visual inspection disclosed
no damages to the core shell, and it is currently
being measured.

Tests have demonstrated that satisfactory liquid-
metal-vapor condensers are now available for ART
and ETU use. Satisfactory zirconium fluoride
vapor traps are now also available.

A test of the ART island bellows under cyclic
strain, temperature, and pressure conditions re-
vealed that the two-convelution bellows would
be inadequate for ART service conditions. The
design has therefore been modified.

Since a failure of the fuel fill-and-drain tank
might have serious consequences, a test tank
is being built for testing under simulated service
conditions. This test is thought to be necessary
because the ART may impose stress and tem-
perature conditions greater than those for which
design properties are known with any certainty.

1.3. Instrumentation and Controls Development

A test was completed of the ART control rod
drive mechanism that simulated ART operating
conditions, except that there was no radiation
present. During the first 1129 hr of the 3000-hr

test, the rod was withdrawn and inserted 1161

times. This portion of the test was followed

 

SR TSR
xviii W

v Tt

. w7
ettt SR

v Ay

B N

et o W T

y

 
 

 

 

 

 

by a series of periods in which the tod was fixed
at positions spaced 5 in. apart for a minimum
time of 200 hr ot each position. The remainder
of the 3000-hr test consisted of further rod cycling
in the same sequence as that used for the initial
phase of the test, The test system has been
dismantled for examination, and if results of the
examinations are satisfactory, it may be concluded
that the ART control rod and its actuating equip-
ment have met all the specifications that can
reasonably be tested without the actual nuclear
tests. '
Additional tests with helium-bubbler-type liquid-
level indicators in the fuel expansion tank of
the aluminum mockup of the ART north head
confirmed - the ‘‘dishing’’ effect on the liquid

surface in the expansion tank of changes in pump

speed. In tests in which measured amounts of
fluid were added to or removed from the expansion
tank, it was found that the level-measuring device
responded accurately.

Examinations of on-off level probes that fauled
at high NaK temperatures in NaK pump bowls

showed that the units fcnled because of internal -

oxidation of the coppet wires and that such

"oxidation occurred when the copper-to-lnconel.

welds were made. The probe has therefore been
redesigned so that the copper-Inconel junction
can be brazed instead of welded. Units of the
new design are being fabricated. ' '

Calibration” and testing of the 2- and 33’-m.-
magnetic flowmeters for use in the ART . und
ETU were continued. All the 3/-m.-magnehc
flowmeters required for the
dellvered and tnstalled

‘Several . ORNL des:gned resustance-type liquid-
use in NaK. were

etal level transducers for
‘completed

~units - mstalled in NaK pump bowls fo obtdin

drift data and to observe wetting effecfs. - Other
_probe materials - ‘and special cocmngs or. plates* _
are to be tested in an effort to improve the wetting
charactenshcsr S:mllar units are being fabricated'

~for use in the ART sodmm expansnon tank, -

Tests of thermocoupl_es were contmue_d
data were obtained for Inconel-sheathed Chromel-
Alumel and platinum, platinum-10% rhodium thermo-

couples exposed to fuel 30 for various periods

ETU have ‘been:
Tests have indicated -
that conhnuous temperature and magnehc flux
momtormg of these units will not be requlped

Tests have “been- run with these

‘moderator

Drlft ii |

 

m

at various temperatures. A new closure technique
in which the junction is made by brazing rather
than by welding is being investigated.

A well type of thermocouple was designed for
temperature measurements of high-velocity high-
temperature ltiquid metals in Inconel piping. The
pressure drop created by this thermocouple was
found to be less than 2.0 psi at 1200 gpm. Drift

tests of sample units in flowmg NaK have been

started.

1.4, "El_lg:ineering Design Studies

Analyses and experimental investigations of the
effects of stress on various portions and com-
ponents of the ART system were continved. In
one test, the resistance of beryllium to fotigue

‘cracking under the temperature conditions an-

ticipated in the ART is being studied. In another
another experiment a 1{‘--sccale model of the outer
core shell made with the inner core shell weld
pattern is to be tested under strain cycling

‘conditions. This test will determine the adequacy
of the welded reactor shells,

"A  stress analysis of the ‘island expansion
bellows at the design condition of 300 cycles

of 90-mil axial deflection at a temperature of

1250°F revealed that the bellows would be sub-
jected to considerable plastic deformation at
the outer bend of the convolutions. . Correlation
with the strain-cycling data for Inconel indicated
that the fatigue life of the bellows would be less
than the contemplated 300 cycles. The resuits

. of the analysis were confirmed in a recent test
~in which the bellows failed after 80 cycles. The
bellows has therefore been redesugned

A detailed stress -analysis of th_e support

structure for the lead shielding of the ART was
‘completed, and an analysis of the water bag was
 initiated. A
~measuring device for use in the ART was made
“to establish the - operating limits of temperature
',and préssure for applications in which dlmensnonal

A stress analysis of a pressure-

stablhfy of the unit is important.

- An oppurotus for’ smuhuneously teshng six
specnmens in' the tempercture range
of 1500 to 2000°F in the ETR is being designed.
The moderator materials to be tested initially
are beryllium oxide, yttrium hydride, and beryl-

lium metal.

W Xix

B et ———————

‘'L

 
 

 

 

 

 

1.5. Design Physics

Calculations were made of the shielding re-
quired for the ART fuel fill-and-drain tank. The
lead shield was designed, insofar as it was com-
pleted, so that the dose rate at any point on its
surface was less than 0.2 r/hr after 100 hr of
operation at 60 Mw, 9 days cfter shutdown. It
was found that the pipes for carrying the NaK to
heat and cool the fuel could not be allowed to
penetrate straight through the shield, ond a con-
figuration that would satisfactorily reduce the
dose rate at such penefrations was developed.
Similar studies indicated that it would be necessary
to overlap adjoining pieces of the removable top
shield of the tank.

In connection with the problem of replacing o
fuel or sodium pump, a preliminary estimate was
made of the dose rate at a point 62.5 in. above
the reactor midplane and 22.5 in. from the axis,
that is, about at the bottom of the fuel pump
motor under the same after-shutdown conditions as
those specified above. The dose rate was found,
for the present shielding geometry, to be around
800 r/hr, and the radiation was mostly - from
residual fuel in the heat exchangers viewed
through the NaK pipe penetrations in the north
head.

The rates of decay-gamma energy emission from
the ART fuel for the first 180 sec after reactor
shutdown following 500 hr of operation at 60 Mw
were calculated for six gomma-ray energy groups.
The results were used in calculations of the heat
deposition rates in the bellows and stem of the
ART fuel drain valves. A total temperature rise
of 106°F was obtained for the maximum allowable
duration of the fuel draining operation.

1.6. Materials and Components Inspection

Materials for the ART,ETU, andtest components
were inspected with the intended use as the
criterion of acceptability. Results of metallo-
graphic examinations were received which in-
dicated that pickling in phosphoric acid could
be used satisfactorily for the removal of magnetic
particles. It was also found that polishing could
reduce rejections caused by surface scratches.
Results were obtained of metallographic exami-
nations which showed that no embrittlement or
attack occurs upon exposure to the materials used
in penetrant inspections.

 

[ ™

 

Inspections of sample pieces and completed
units submitted by vendors were made, including
radiators and radiator materials from York Cerp.,
heat exchangers from Griscom-Russell and from
Black, Sivalls & Brysen, forgings and weldments
from Midwest Piping Company, tanks and bellows
expansion joints from Process Engineering Corp.,
bellows from Fulton-Sylphon Company, and Inconel
and stainless steel valves from Hoke Valve, Inc.

Procedures are being developed for the in-

_ spection of seven critical welds on the ART and

ETU that can be inspected only through sections
of beryllium. Normal radiographic techniques
cannot be used because beryllium scatters the
incident x-ray beam and causes general fogglng
of the film.

1.7. Heat Transfer Studi_es

The effect of stresses and cyclic frequency on
the strength and corrosion characteristics of
Inconel in an Nc:F-ZrF“-UF4 (50-46-4 mole %)
environment was studied in the pressurized-
system thermal-cycling apparatus. Data were
obtained for straight tubes which indicated that
the extent of corrosive attack on Inconel is
dependent on the frequency of the thermal cycling,
wnh the critical frequency in the vicinity of
é to 1 cps. The effect of prior strain was in-
vestigated in tests of bent tubes. In all cases
the attack was greater in the bends than in the
straight sections of the tube, but unexpectedly
no difference was detected between the attack
on the tension and compression sides of the
bends. As in previous tests, fine-grained tubes
showed intergranular attack and coarse-grained
tubes showed general attack. Inconclusive tests
of machined welds were run and additional tests
of inspected and approved welds are to be run.

After two runs in which the operating periods
fell short of the scheduled 100 hr because of
difficulties with liquid-level control of the high-
frequency pulse-pump thermal-cycling loop, the
loop was operated for 100 hr at 1 cps with the
fuel mixture NoF-ZrF,-UF, (56-39-5 mole %)

as the fluid medium. Outside surface temperature’

oscillations of 164°F were attained at the inlet
end of the test section. Metallurgical examination
of the test specimen has not been completed,

but standard dye stain and radiographic exami-

nations show no superficial damage.

i¥

&)

 
 

 

 

 

 

. L e et s
[
LR et
PR

Temperature measurements made with the nickel-
plated junction of an Inconel-nickel gunbarrel
thermocouple flush with the wall of a thick-walled
Inconel tube exposed to sinuscidal temperature
oscillations in water showed good agreement with
results of theoretical calculations. A similer
Inconel-nickel gunbarrel thermocouple that was
not attached to a tube was tested in static NaF-
ZrF -UF, (50-46-4 mole %) at 1200°F for 300 hr.
It was determined that rapid removal of the
metallic junction between the central nickel
wire and the Inconel by the flucride salt precluded
the use of this type of thermocouple for measuring
rapid surface temperature transients in a fluoride
environment.

The full-scale Plexiglas model of the ART core

was used for studies of the effect of inserts
and ramps in the core volute and of thin baffles
in the core region on the core flow. No significant
improvements in flow were noted. Velocity and
pressure-loss data were obtained in water flow
tests of the 10/44-scale model of the 21-in. ART
core with an ART type of entrance header, a
core inlet collimator,
The data show . significant flattening of the
velocity profiles throughout the core. . Simulated
one-pump operation with the screen-filled core

indicated that the screens and collimator effec-.

tively eliminated circumferential asymmetries in
the flow. Measurements were made of the pressure
variation in the inlet header. Velocity profile
photographs were obtained of the flow in the
10/44-scale model of the ART by using the
phosphorescent particle technique.

Measurements -
transfer coefficients for

(40-7-53 -wt %) flowing through electrically heated

tubes. = The date agree with the standard cor-

* relations. Preli iminary results for forced-convection )
~ heat trcmsfer with KCI-LiCl (41.2-58.8 mole %)

flowing in a ‘stainless steel tube fell below the
standard correlations.
filled half-scale volume-heat-source core model
show markedly reduced temperatures and fluc-
_tuations " along the island wall but little change
at the outer wall from earlier results obtained
with . a vaned entrance system. Initial heat

transfer data for source-vortex flow “indicate a

fivefold increase in comparison with straight,
turbulent flow when the data are compared on
the basis of equal system energy expenditure
per unit of heat transfer area.

 

A ?;yé'tem~ for

and five core screens.

- the sample thickness.

_were completed of the heat

NaNO,-NaNO,-KNO,,

3 .. gradient quenching experiments revealed, however,
several differences. from the phase behavior pro-

Data from the screen- -
- of the system NoF-HfF .

_are isostructural with NaF-ZrF
" responding stoichiometric' formulas.
_ however, fewer phases in the NaoF-HfF 4
. than .in the NaF-ZrF system.

Sl et T

determining heat transfer in
flowing mercury containing a uniform volume heat
source is being constructed. Experimental dota
on mass transfer in alkali metal systems were
compared with results obtained with two theo-
retically derived expressions. It was concluded
that a more general hypothesis that combines
the mechanism of hot-zone attack with a
nucleation-deposition mechanism in the cold zone
is needed to explain the experimental results.

Preliminary results indicated a value of 0.21
Btu/lb:°F for the heat capacity of LiCl-BaCl
(32.2-67.8 wt %) and of 0.24 Btu/Ib:F for LiCl-
SCl, (22.5-77.5 wt %). Measurements with o
modified variable-gap apparatus yielded a value
of 0.34 Btu/hr+ft:°F for the thermal conductivity
of the salt mixture NaNOz-NaNos-KNoa (40-7-
53 wt %). New data were obtained for the thermal
conductivity of NaF-ZrF -UF, (50-46-4 mole %)
by using a calorimetric dewce. The values were
considerably below the thermal conductivities pre-
viously obtained (1.0 to 1.5 Btu/hr-ft-°F) by using
the variable-gap apparatus.

The apparent density of ZrF, vapor deposits
was measured and found to vary inversely with
An attempt to measure
the thermal conductivity of a ZrF, deposit in-
dicated a value of 0.045 Btu/hreft-°F.

PART 2. CHEMISTRY
2.1. Phase Equilibrium Studies

- Further studies of the system KF-UF, confirmed
the liquidus curve presented previously. Thermal-

posed earlier. The formulas of the equilibrium

“phases are 3KF.UF,, QKF,UF‘, 7KF.6UF,, and

KF.2UF,.

Refmed phase equnllbr:um studies were made
Seven NoF-HfF phases
phoses of cor-
There are,
system

The 'system KBF, -NuBF was studied as part
of a search for Iow-melhng salt mixtures for use
as reactor coolants. It was found to be a simple
binary system with o eutectic containing about

90 mole % NaBF 4 that melts at about 357°C.

xxi

 

It

 
 

 

 

 

 

SNV e e vl g

BN e
A

TRt s ey

el v e @

The same unidentified compound was found in
both LiF-MgF,-YF; and LiF-ZnF,-YF, through
petrographic and x-ray examinations. Preliminary
indications are that the unidentified material is
an LiF-YF; compound.

2.2. Chemical Reactions in Molten Salts

The determination of the  activity coefficient
of NiF2 in the molten mixture Nc::l:--Z'rF4 (53-
47 mole %) by the study of the equilibrium
constants for the reaction

NiF,(@) + H,(g) == Ni(s) + 2HF(g)

was concluded. The investigation revealed that
NiF, in solution, under the experimental con-
ditions used, behaves in a very different manner

‘than that predicted by the free energies of

formation for solid NiF,. This result was un-
expected because good agreement had been
obtained, in a previous similar study, between
the experimental ond calculated behavior of
FeF, in the same solvent. Thermodynamic con-
siderations of the unexpectedly large activity
coefficients of NiF, in fuel mixtures indicate a
probable error in fime estimated free energy of
formation of Nin.

Studies were made of the reduction of UF,
by V° at 600 and at 800°C in NaF-ZrF, (50-
50 mole %) and in NaF-LiF-KF (11.5-46.5-
42 mole %) and of the reduction of UF, by W° in
NaF-ZrF (53-47 mole %). The results showed
that V° is not stable in contact with UF, ot
either temperature in either medium. In marked
contrast to earlier experiments with NaF-LiF-KF
as the reaction medium, W° was found to be
stable with respect to UF in the reaction medium
NaF-ZrF,. '

Additional studies of the stability of chromium
fluorides in NaF-LiF-KF were made with both
CrF3 and Cer. For the tests with CrFa, chromium
metal was added.  Since the experimentally
determined values for Cr** did not give a reason-
able balance of Cr*t, Cr***, and total chromium,
it was concluded that the experimental values
were probably inaccurate.

Additional experimental studies of the reduction
of FeF, by Cr® in NaF-ZrF, (53-47 mole %) in
which higher FeF. concentrations were used
further indicate that reliable values for the Ce*-

" to-Fe _H ratio cannot be determined in this reaction
' ‘medium. A study of the reaction with RbF-ZrF4

Vet et g, b

(52-48 mole %) as the reaction medium was also
made. Large Cr**to-Fe** ratios were found,

and there was a difference between the calculated

and experimentally determined chromium -con-
centrations.  Additional experimental “work is
planned to resolve these discrepancies.

"The studies of the solubilities of the structural
metal . fluorides CrF,, FeF,, and NiF, in molten
fluoride mixtures were continued. Data are now
available. for the solubilities at 600 and 800°C
in the following mixtures: NaF-ZrF, (53-47
mole %), LiF-ZrF"4 (52-48 mole %), KF-ZrF4'
{52-48 mole %), and RI::F_-ZrF4 (52-48 mole %).

Studies of the activities of the metallic con-

stituents of the nickel-molybdenum alioys being

developed .to contain molten fluoride salts were
initiated. Preliminary experiments are under way
to determine the feasibility of measuring th
activities by an emf method. o '

Ratios of activity coefficients of NiF, to FeF,
in NaF-ZrF4 (53-47 mole %) and in KF-LiF
(50-50 mole %) were obtained in Daniell cells
ot 650°C. Similor measurements were made of
Cr-Ni couples in NaF-Z¢F, (53-47 mole %). The
values for the Ni-Fe couple were in agreement
with results obtained previously, and, if the
activity coefficient for FeF, is taken to be 3,
the activity coefficient of NiF, obtained from
these measurements is (1.8 + 0.3) x 103, based
on the solid as the stondard state. The data
obtained for the Cr-Ni couples indicated that
the container material affected the results, which
were thus inconclusive.

Various experimental difficulties involved in an
emf study of the effects of various compositions
of the solvent LiCI-ThCl, on a solute in dilute
concentration were resofved, and 'PbClz was
selected as the solute. Cooling curves were
run on various LiCl-ThCl, samples to study
the solvent composition limits at 700°C, and
important features of the phase diogram of the
system were established.

Determinations were made of the solubilities of
argon and of xenon gases in molten NaF-ZrF 4

(53-47 mole %) as functions of pressure and tem- '

perature. The data show that in this solvent the
solubilities of the gases follow Henry’s law,
they increase with increasing temperature, and
they decrease with increasing atomic weight of
the gas. Similar studies of the solubilities of
HF gas in NoF-ZrF, have shown that the gas

 
 

 

 

(3) less stable oxides.

limation. “pressure of ‘ZrF .
- the volue for ZrF was obout three ‘times that

species ‘was monomeric..

follows Henry's law and that the solubility de-
creases with increasing temperature and increases
with increasing NaF concentration. Preliminary
dota on the solubility of HF in" NaF-KF-LiF
reveal that the solubility “is roughly 50 times
greater than in NaF-ZrF ;o

The solubility of CeF in NaF-ZrF, is bemg
studied as a function of composmon of the
solvent. The solubility has been found to de-
cregse . as the proportion of. the tetravalent con-
stitvent is decreased, -
more like NaF or BoF2 than like ZrF_4. . Data
were. also ‘obtained  which . indicated. that: the
evoporohon of - ZrF durmg ‘these experiments

ond the CZeF3 behc:Wes'

) .

* o qupaapytpET——,

between the values indicates that BeF, is
monomeric, it should be possible to measure
activities from vapor pressures over BeF,-rich
solutions.

2.4. Production of Purified Mixtures

Approximately 6000 g of YF, was prepared and
MgF, -and LiF were added to make about 8000 g
of the processed YF ,-MgF,-LiF mixture for use
in the preparation of yttrlum metal. About 5 Ib

~ of CrF; was prepared, and most of it was con-

does - not - exceed the voccurocy .of the defer- |

minations (£5%). .
The possibility of prec:pttohng the rore-eorfh

~ fission products as oxides is being consudered_

as a means of- reprocessmg fuels based on the

NaF- KF LiF fuel solvent.: Three ‘methods of oxide
formohon ‘are bemg studled that involve reactions

with (1) corbonofes, 2) alkeli hydroxides, and
In connection with this
study, experiments” are being - conducted to in-
vestigate the: stoblllty of UO F
mixture,. -

In studies of the oxldcmon of mlxtures composed

of sodium and potassium, it was found to be quite-
“difficult to control-all the variables which affected

the oxidation process. The results obtained were

not precise, but it appears that the ratio of

sodium-to-potassium: in the oxides decreases with
increasing potassium concentration.

2.3, Physical Properties of Mohen Materiuls o

verted to CrF,.
prepared. :

Experimental batches of various fused salt
mixtures were prepared for use in phase equi-

A small batch of VF3 was also

- librium studies, physical properties testing, and

corrosion testing. Conversion of 10 Ib of low-

~ zirconium HfCl, to HfF,- was completed with a

) total recovery o? 96.3%.

ina fused sult. _

The subhmoflon pressures of HfF ¥ were de-'—_ :

termined for comparison wuth the known" sub-

the factor of 3 difference in ‘the vupor pressures
may-be ‘used’ as @ basis for separotmg zirconium .
from - hofnlum ing’ high-temperature {above 925C)
process and’ for the produchon of ZrF, from o -
stortmg rmxture of zirconium. und ‘hafnium oxldes.
New meosurements of the vapor -pressures of -

It was found that -

_were -drilled. - _ 7
‘correspond precisely to the images of the quartz
‘end windows of the absorption cells.

One hundred filling and draining operations were
performed during the quarter that involved 785 kg
of liquid metals and 1229 kg of salts. A series
of experiments for determining the amount of fuel

‘which will be held up on various internal surfaces

of the ART was initiated.
2.5. Auolyti‘cal jC_-heuli_stry of Reactor Materials

The instrument for the detection of NaK in air

‘by measurement of the absorption of sodium

resonance radiation by sodium atoms was modified
by the addition of an additional light stop which
will exclude the continuum radiation of the heated
furnace from the photomultiplier tube. This light
stop was effected by means of a plate in which
two apertures, approximately 0.08 in. in diameter,
‘The circular holes in this plate

In order to

- achieve the precise . focusmg and alignment of
of HfF , ot a given temperoture. RURE fhought thot -

BeF “vrere ‘made by ‘using “the Rodebush-Dlxon

method

The values obtamed ‘were - in “good

ogreemenf with literature * values “in the range

800 to 1025°C that were obtained by using the

carrier gas method and assuming that the vapor

Since the -agreement . .

- complex the iodide and,
.chloride rather than iodide system. An excellent

‘the optical system that is required, a rigid optical
_bench fabricated of 18-in. channel iron was in-
: corporoted into the apparatus.

"Studies of the butyl |od|de-|odme method for

the - determination of oxygen in metallic lithium

‘revealed that several interfering reactions occurred
‘during dissolution of the metallic lithium in

ethereal. solution, - Mercury was - odded in order -

to eliminate the mterference of the excess iodine
that remained

‘after reaction of the lithium.
Anhydrous mercuric chloride was then added to
in effect, produce a

 

sae b 1o La

 

o [P

 

 

I
‘“‘,1 xxi

 
 

 

 

 

recovery of oxygen was obtained when 0.5 mg
of Li,O was added to a simulated sample solution.

The apparatus for the determination of oxygen
as oxides in fluoride salts and in metals by the
fluorination of the oxides with KBrF, was com-
pleted and assembled. A detailed procedure for
use of the apparatus was formulated and is
presently being tested. Zirconium oxide was
selected as a standard for calibration of the
apparatus. -

The method for the determmahon of nickel as
its chelate complex with 4-isopropyl-1,2-cyclo-
hexanedionedioxime was applied to samples ob-
tained ‘from studies of the solubility of nickel
in molten sodium. Since the samples are contained
in molybdenum buckets, some molybdenum is
also present in the sodium. In order to avoid
interference by this molybdenum it is necessary
to dissolve the sodium sample in water rather
than in methanol. ' A special apparatus, fabricated
from fluorothene, was designed in order to carry
out this dissclution safely and efficiently.

A series of tests was initiated of the possible
reaction between pump lubricants and alkali
metals in an effort to determine whether the
reaction can produce acetylene. Initial tests
showed that *‘Gulfspin-35"" and NoK in equdl
volumes reacted to produce considerable acetylene,
with an accompanying significant increase in
pressure. Approximately 2% of the alkali metal
apparently was converted to the carbide.

Additional work was done on the determination
of aluminum in mixtures of fluoride salts with
pyrocatechol viclet. A study was made of inter-
ferences that would normally be encountered in
corrosion tests with fluoride salts and the newly
developed nickel-molybdenum base alloys. Iron
and titanium interfere seriously, but nickel,
chormium, molybdenum, and niobium do not
interfere. '

Studies of the application of tri-n-octylphosphine
oxide for the solvent extraction of metals of
interest in the ANP program were continued.
The effect of sulfuric acid concentration on the
extraction of titanium with TOPO was demonstrated
in experiments in which only 6% of the titanium
was extracted from | M sulfuric acid in contrast
to 99% from 7 M sulfuric acid. Niobium was found

to be extracted gquite well by TOPO from hydro-

chloric acid solutions, relatively well from sulfuric
acid, and essentially not at all from nitric or
perchleric acids.

 

The weighing compartment of a Cahn Electro-
balance was installed in o vacuum drybox. By
means of this microbalance it is possible to weigh
trace quantities of materials. that must be pro-
tected from either atmospheric contamination or

“moisture.

PART 3. METALLURGY

3.1. Nickel-Molybdenum Alloy Developmeni
Studies

Developmental work has continued on a new
container alloy for the fuel mixture NaF-KF-LiF-
UF, (11.2-41-45.3-2.5 mole %, fuel 107) at service
temperatures up to 1800°F. Emphasis has been
placed on determining the optimum alloy com-
position, on determining whether the most promising

alloy, INOR-8 (composition range: 15-19% Mo,

6-8% Cr, 4-7% Fe, bal Ni), can be produced as
a ‘‘commercial’’ item, and on determining the
fabricability of duplex tubing of INOR-8 clad
with stainless steel for use in heat exchangers.
Since corrosion resistance, weldability, and creep
properties have not yet been completely evaluated,
two other compositions, INOR-8 modified and
INOR-9, are also being studied as alternates.
The modified INOR-8 alloy contains 2 to 4 wt %
niobium to increase creep strength, and INOR-9,
which contains no chromium but is strengthened
with niobium, should be corrosion resistant at
1800°F.

The physical properties of INOR-8 have been
determined, and means for producing sound ingots
are being investigated. Studies of oxidation have
shown that unless the alloy contains about 6 wt %
chromium the oxidation rate is not acceptable.
Tests of corrosion resistance and mechanical
properties are continuing.

Preliminary tests have shown the oxidation
resistance of INOR-9 tc be .about the some as

“that of Hastelloy B and the creep properties to

be about the same as those of INOR-8. It is
attacked by fuel 107 to a depth of 1 to 3 mils
in 1000 hr at 1500°F.

Studies of produchon heats of INOR-B have
indicated that it is a ‘“commercial’’ alloy. Upon
completion of orders which have been placed with
various vendors, sufficient material will be
available to fabricate about 30 forced-circulation
loops, one in-pile test loop, and specimens for
a rigorous program of creep and weld tests.

i

 

 

 

il

 

 

"

 
 

 

 

©Twe" thermdl-convection - loops |
-',exper;mental mckel-molybdenUm elloys were’ clso__.
’opercted for. 1000 hr with fuel 107 and a maximum -
{hot-leg tempefature of 1500°F, The hot leg of
|oop 1136 (16% Mo—2% AI-'IS% Tl-bu[Nl) showed' -
heavy " surface . roughemng ‘and. pits to. a ‘depth of
In addition, very large, |rregu|qr-shc|ped '
voids were observed at depths as much as 9 mils
but they appeared to be -

 

.- . ;. o Pt
PR W

Since all nickel-base alloys exhibit mass
transfer when used as containers. for flowmg
sodium ‘at high - temperatures, it appears that
composite - fubmg will be- required for service
temperatures of < 1650°F - “and . dbove in fluoride
fuel-to-NaK -heat exchangers. It has been demon-
strated that acceptable tubing can be produced
by coextruding composite billets of the nickel-
base alloy and type 316 stainless  steel. The
weldability of the composite tubing and the effects

of diffusion of the ~components are being in-

vestigated. |

- Tests have shown fhat alloys of the INOR- 8
type can be brazed in a dry-hydrogen atmosphere
with Coast Metals brazing alloy No. 52 and similar
brazing olloys by using conventional techniques.
Sound weld deposits  can be made, and joints
have been found to have satisfactory properties
in both the as-welded and the aged conditions.

Evaluation studies of the creep charucterishcs
of the nlckel-molybden_um alloys under @ constant
load (8000 psi) in contact with fuel 107 at 1500°F
were continved. The data indicate that INOR-8
with 0.06 to 0.1% carbon meets the desired
strength requirements.

A forced-circulation loop fabrlcoted of an -ex-
perimental nickel-molybdenum alloy (17% Mo—6%
Fe-bal Ni) completed 1000 hr of operation with
fuel 107 ot o maximum fuel-to-metal interface
temperature ‘of 1760°F. Attack of the hot leg
was confined almost enhrely to grain boundaries
and reached a depth of 4 mils ot the hottest
point. A thin deposit of metal crystals was

present over most of the cold leg, and there were -
widely dispersed - clumps of metal deposits up -

to- 6 mlls in thickness.’ The deposited materm!

,contamed slzable quantities of nickel and - iron.
The outer surfoce, which was exposed to_air, -
‘developed a. ~heavy umform ‘oxide - film; and-.-.'-r.;_i )
~oxidation - wns found ‘along grcun boundarles to o

" The program for the mveshgahon of the strain-
: cycle properties ‘of fine-and coarse-gramed Inconel
- was- continued. . Results of tests of tubular speci-
_mens " at 1600°F
'sfrem-cycle properhes ‘ot 1600°F are “similar to

-""_depths of as much as 9 mils..

5 mlls.

below - the surface,
fabrication defects rather than areas of _c_mock.

 void formation to a depth of 3 mils.

" 1500°F.

- roughened,
‘moderate surface pits and some general subsurface
voids to a depth of 1.5 mils were observed. No
~mass-transfer deposits or layers could be seen
in either Ioop.

fabrlcafed o?i:?fi‘:

 

No deposit was found in the cold leg, but there
was light surface roughening. An analysis of
the fuel after the test indicated considerable

- reaction of the fuel with the aluminum and slight
‘reaction with the titanium.

The other loop (No.
1155, INOR-8 plus 0.5% Al~6% Cr~5% Fe-0.5%
Mn-0.06% C) showed no cold-leg deposits, but
the hot leg showed heavy surface roughening and
surface pits, with heavy intergranular subsurface
Fabrication
defects in the form of cracks and voids to a
depth of 7 mils were also found.

A summary of the thermal-convection loop results
obtained thus far indicates that the maximum
attack has not exceeded 3 mils in 500 hr at
Increasing the time of the test from
500 to 1000 hr causes the attack to increase
T to 2 mils.

In order to investigate the compatibility of
molybdenum with nickel-molybdenum alloys in a
commen fluoride fuel circuit, two Hastelloy W
thermal-convection loops with molybdenum inserts
in the hot legs were operated with fuel 107.

 One loop was operated for 1000 hr ot 1500°F,
- and the second operated at 1650°F for 517 hr.

Numerous fabrication defects were found along
the surfaces of the molybdenum inserts, but there
was no evidence of attack. The Hastelloy W
surfaces exposed to the salt showed heavy
surface pits ond shallow void formation to a
depth of 3 mils in the loop operated at 1650°F
and 2 mils in the loop operated at 1500°F. The
cold-leg samples of Hastelloy W were moderately
‘and in the loop operated ot 1500°F

3.2, Developmental Studies of Inconel

in fuel 30 mdncate that the

those ot 1500°F and’ show that, as in previous

tests, ‘coarse-grained- lnconel is greatly affected
“by this environment. -
‘the effect of prior strain cycling on the creep

“The evaluation study of

strength has indicated that the creep strength
and rupture life are reduced by the first 100

S R A

 

XXv

 

ko e

 
 

 

 

 

 

 

cycles and that 300 cycles does not further
reduce the creep strength; however, the total
elongation and rupture life after 300 cycles are
less than ofter 100 cycles.

A test program was initiated for evaluating the
strain-cycle properties of Inconel weldments in
orgon aond in fuel 30. Dota from initial tests
compare favorably with standard tubular specimen
data. _

The strain-cycle properties of hot-pressed,
reactor-grade beryllium are also being investi-
gated. Results of initial tests at 1250°F illustrate
the difference between a brittle metal and a
ductile clloy.

Results of strain-cycling tests of carburized
inconel at 1500°F indicate that carburization
does not impair the strain-cycle properties.
Relaxation tests of fine-grained Inconel at 1100°F
were completed, and the results were found to
be those expected on the basis of previous tests.

In order to obtacin information on the behavior .

of Inconel Under multiaxial conditions, ¢ biaxial
creep-test program was initiated. Data obtained
in initial tests demonstrate that the axial strain
rate is independent of the state of stress and
dependent only upon the stress in the oxial
direction in tests in which the ratio of the
axial to tangential stress is greater than Y.

Tests were run in order to determine the
relative strengths and ductilities of Inconel
weld metal and base material. Both the yield
and vultimate strengths of the weld metal are
greater than those of the parent metal at tem-
peratures from 1400 to 1800°F, whereas - the
ductility of the weld metal is lower.

Several new brazing alloys were tested for
oxidation resistance ond found to be satisfactory.
Screening tests of a series of high-nickel-content
brazing alloys for use in fabricating NaK-to-fuel

heat exchangers were completed. It was found-

that porosity of the fillets was related to the
boron content of the brazing alloy.

A study of the effect of grain size on the cor-
rosion of Inconel by fuel 30 was made with the
use of thermal-convection loops containing hot-
leg sections subjected to annealing treatments
intended to produce grains similar to those formed
during the furnace brozing of heat exchanger and
radiator tubes. Preferential grain-boundary attack
such as that found in the examination of heat

XXVi

|5

 

exchangers occurred to a limited extent in the
coarse-grained sections of the loop. In reviews
of previous test results indications were found
that relatively deep attack can occur. under
certain conditions of grain-boundary orientation in
coarse-grained material. ' '

3.3. Welding and Brazing Studies

An examination of an all-welded Inconel impeller
that cracked in service revealed that the fobrication
procedure provided only a limited joint length
where the vanes were welded to the top plate,
ond the joint design and lack of accessibility
permitted only very shallow weld penetration.
A modified fabrication procedure is now being
used in which the joints are brazed for additional
reinforcement and to improve reliability. |

Tests were performed which indicated that cracks
in the braze fillets of tube-to-header joints of
NaK-to-air radiators could be healed by rebrazing.
Such cracks result from transverse shrinkage and
distortion of the tube sheets during the deposition
of the longitudinal welds of the headers.

Tests were run to study welding and brazing
procedures for making the tube-to-tube sheet
joints in the heads of the ART fuel fill-and-drain
tank. In order to produce joints free of root
cracks, a low restraint, trepanned type of joint
was designed. The brazing alloy will be contained
in annular sumps machined in the tube sheet
around each tube. When the tube sheet reaches
the brazing temperature, the brazing alloy will
flow into the space between the tubes and the
tube sheet through three holes drilled from each
sump into the tube hole. A relatively low rate
of temperature rise will be used to aveid brazing
alloy liquefaction before the massive header
sheet reaches the brazing temperature.

A study was initiated to ascertain whether,
radiographic and metallographic determinations of
porosity in tube-to-header welds could be cor-
related. The limits of detection by the radio-
graphic method are being determined as an initial
step in this study.

Fabrication of a small semicirculor heat
exchanger for experimental tests was under-
taken. The initial planning and procedural
experimentation have been completed, and the
first of two heat exchangers will be completed
soon.

W

 
 

 

 

3.4, Corrosion and Mass Transfer Studies .

The possibility of using vanadium as a con-
stituent of a container materidl for lithium is
being investigated. After 100 hr ot 1500°F in
a seesaw furnace apparatus, a considerable
amount of mass transfer was found in the cold
zone of a vanadium capsule filled with fuel 107.
A static capsule showed only ‘slight surface
roughening of the bath zone after 100 hr at 1500°F.

Seesaw furnace tests were also conducted on
molybdenum ond niobium cups enclosed in the
hot zones of type 316 stainless steel capsules
while in contact with lithium at 1500°F. Exami-
nation of the surfaces of the refractory metals

following the tests showed no attack, and no

mass-transfer deposits of niobium- or molyb-
denum .could be detected either metallographically
or spectrographically on the sfamless steel
cold-zone walls. :

‘Mallory 1000, which is a tungsten-mckei-copper
alloy, was tested for possible use as a transition
layer between tungsten carbide-cobalt cermets and
Inconel in disk and seat brazed joints of sodium

or NaK valves. Specimens exposed to sodium for -
100 hr ot 1500°F in seesaw furnace apparatus

lost 1.3% in weight and ‘were completely pehe-
trated. A ‘heat treatment at 1920°F for l/2 hr

prior to testing in ‘sodium did not change the -
The quantity of copper.

corrosion resistance,
(4 wt %) present in Mallory 1000 .is sufficient

apparently to cause this material to be of doubtful

value for long: exposure to sodium or NaK at
1500°F. '

Tungsten carbide- and :
cermets were tested in sodium to determine their
suitability as seat ‘components of valves for
sodium or ‘NaK service at 1000°F.
1200°F the specimens were unaflacked
- Experimental

were - continued.

500-hr tests without pluggm'g. The lowest rate
of weight loss ‘was 60 mg/in.2 in 100 hr, “which
krepresents ‘quite heavy and rup:d ‘attack.  In
future. studies primary emphasis will be piuced
“onrefractory metals and corrosion inhibitors.

A stainless steel inert-atmosphere chamber in
which thermal-convection loop tests can be con-
ducted was used for tests of unclad niobium

t7-

~titanium - corbldeebase'

ln tests ot,

‘studies -of corrosive aflock of
containers by molten Ilth:um_ at high temperatures -

, - Thermal-convection loops fab-
ricated ‘of stdinless steel have been: tested, and-
only two of ‘eleven loops completed scheduled

' centratlon, .
“centration " is less fhcn 0 5wt% at the hlghest

thermal-convection

SRR

loops. Both loops tested failed because of leaks
that developed in the saddle-weld areas. Loops
with welds fabricated by a different design are
to be tested. Preliminary tests of specimens
of refractory metals suspended in the chamber
indicate - that the chamber will provide cdequate
protection for the metals and thus eliminate the
need for oxidation-resistant claddings.

Forced-circulation loops of both Hastelloy B
and Inconel were operated for 2000 hr with sodium.
The maximum sodium temperature in these tests
was. 1500°F, and a 300°F temperature drop existed
between hot- and cold-leg sections. Metal deposits
were present in the cold legs of both loops;
weight “measurements of the deposits indicated
a buildup of 21 g in the Hastelloy B loop compared
with 20 g in the Inconel loop. Comparison of
these test results with those of 1000-hr tests
under identical operating conditions indicate that
the rate of mass transfer during the second 1000-hr
period is much slower than the rate during the
first 1000-hr period.

A Hastelloy W forced-circulation loop which
circulated sodium for 1000 hr ot a maximum
temperature of 1500°F showed mass transfer
deposits that were approximately 17% greater

in .weight than the deposits observed in a

Hastelloy B loop under similar conditions. The
deposits were found by chemical analysis to

contain 97% Ni and 3% Cr.

A type 347 stainless steel forced-circulation
loop after operation with sodium at 1500°F showed

only slight traces of mass-transfer deposits.

The loop was operated for 1000 hr with a tem-
perature drop of 300°F. '
-Studies of the kinetics of the process of dif-

_-fuswn ‘of nickel in- liquid lead were continued
.as part of a fundamental investigation of the
. mnss-transfer process. .

Results obtained thus far
indicate that. the diffusion coefficient for-the
dtffusmn of nlckel in lead is a function of con-
~even though the saturation con-

temperufure used. -
A vacuum still’ for producing hngh-pumy sodlum

" has been completed and is being leak checked.
With this system the sodium can be transferred
,from the receiver to the pot and redistilled with-

out opening the system. -.Studies have been
initiated on the mechanisms of mass transfer in
loops by using o benzoic

xxvii

 
 

 

 

 

acid and water system to simulate a liquid-metal
system. The mass-transfer process in this system
has been shown to be hot-zene controlled.

3.5. Materials Fabrication Research -_

Experimental = studies were continued of the
production of high-purity yttrium metal by the
reduction of a YF,-MgF,-LiF mixture with lithium
to yield an yttrlum-magnesmm alloy.  Slight
modifications of the procedures used during the
reduction step. improved the yield. A vacuum
heat treatment of the yttrium-magnesium alloy is
being  investigated as o means of disfi"ing off

the magnesium and excess lithium.

Equipment is being built with which to measure
the rate of diffusion of hydrogen through potential
cladding materials for hydride moderators. The
equipment will also be used for measuring the
rate of loss of hydrogen from clod hydrided-metal
assemblies and for monitoring hydrogen losses
during thermal cycling and hydrogen migration
tests.

A study was made of the available information
on the volatility of BeO in steam. The data
indicate that under some conditions it would
be entirely possible to corrode and transport
large quantities of BeO in steam in relatively
short times. Several hot-pressed BeO specimens
were fabricated for physical tests.

An apparatus is being designed for the de-
termination of the heat throughput required to
cause initial failure in ceramic cylinders. The
data obtained with this apparatus will be used
in evaluating the thermal-stress resistance of
ceramic materials.

Dense zirconium hydride samples were prepared
to establish procedures for hydriding ytrium
metal in the new hydriding epparatus. The
equipment is now ready for production operation.

3.6. Metallogrephic Examinations of Engineering
Test Components After Service

A metallographic examination was made of the
York Corp. ART prototype NaK-to-air test radiator
which foiled after 870 hr of testing, but no
specific cause for the failure was found. The
microstructures of the tube that failed and
adjacent tubes did not indicate stress failures,
and there was no evidence of incipient failure.

 

xXxXvili ‘%

 

York Corp. test radiator No. 16, which also
failed during testing (after 1438 hr) was so
extensively. damaged by fire that the metal-
lographic data were inconclusive. Five tubes
were found to be plugged. Chemical analyses of
two types of plug material indicated mass transfer
in the one case and fuel constituents in the
other. The presence of fuel components indicate
that there had been o leak of fuel into NaK at
the heat exchanger which was alsc being tested
in the test system.

Examination of the heat exchqnger (Process

Engineering . Corp. heat exchanger No. 3, type
SHE-7) did not reveal any open cracks, but the
fuel probably lecked through severcl tubes in
which there was complete grain-boundary pene-
tration. A metallic deposit was found on the
fuel side of o tube from the hot-header area that
was shown by spectrographic analysis to be
predominantly .nickel; the heat exchanger was
fabricated of Inconel. The fuel (No. 70) circulated
in this heat exchanger was NaF-ZrF, -UF (56-
39-5 mole %).

A similar heat exchanger (Process Engineering
Corp. No. 2), which had been removed from a
test system because of a cold trap failure after
1845 hr of testing, was also examined. This heat
exchanger had been circulating NaF-ZrF,-UF
(50-46-4 mole %, fuel 30) and NaK. Grain-boundary
voids were found that completely penetrated the
tube walls in the hot-header area. There was
some general corrosion attack on both the fuel
and the NaK sides of the tube walls, and the
usual gold-colored metal deposits were found on
the surfaces exposed to fuel in the cooled area.

An Inconel centrifugal pump that had circulated

NaK for 4400 hr at about 1200°F was examined
after it was found that lubricating oil had been
lecking into the NaK stream. Carbide precipitates
were found on all wetted parts.  The extent of
carburization was not sufficient to damage the
heavy sections, but all thin sections, such as
bellows and diaphragms, were sufficiently affected
to require replacement.
- A metallographic study was also made of the
fused-salt - corrosion resistance of saddle welds
from a large number of thermal convection loops.
No evidence of intergranular corrosion of im-
properly oriented grains was found.

 
 

 

 

 

 

 

 

3.7. Nondestructive Testing

The recent application of eddy-current methods
has involved the measurement of thin-metal
sections and of cladding layers on metallic
laminates. Also, a mothematical study is being
made of the impedance changes in a probe-coil
over a plane metal object of apparently infinite
thickness. This study will be extended to the
case of a probe-coil over a plane metallic object

of finite thickness. The results of continued

efforts, both the experimental and the theoretical
studies will be used to develop techniques of
thickness measurement with induced eddy currents.

The ultrasonic and the eddy-current inspection
of pipe, plate, and tubing for the construction of
the ETU and the ART continued. The rejection

rate for these materials was normal.

Two core shells were inspected for wall thlck--.

ness variations, for lamination-like discontinuities,
and for crack-like imperfections. No defects were
found, but one of the shells was rejected because
the section thickness was less than the specified
allowable dimension. '
Ultrasonic resonance techniques were used to

detect unbonded areas in Inconel-clad copper-

boron carbide plates. Unbonded areas greater
than the area of the ultrasonic transducer were
easily detected from either side of the laminate,
and .unbonded areas as small as 1/8 in. in diameter
were detected with reasonable reliability.  All
unbonded areas detected with resonance ultrasound
were confirmable by stripping the cladding from
the core, and, in the course of stripping the
cladding, no unbonded areas were located- that
had not been revealed by fhe mspechon. .

, PART 4. ‘ RADIATION DAMAGE
4 Radictlon Dumage

Exommuhon of ‘the radioactive - portlon ‘of in-
pile loop No. 6 has mdlcated that the maximum -
depth’ of corrosion was. 3 mils. This Inconel
forced-circulation loop was o'pérated in the MTR.

‘Three. capsules contaamng moderator materials

(graphlte in nickel, ZrH, in ‘molybdenum, and

BeO in Inconel) that had been “irradiated in the
MTR were examined, The: dimensions of the
capsules were not upprecmbly affected by the
irradiation. The BeQ siugs were removed from
the capsules, and no evidences of swelling or
cracking were found.  The cans will be removed

from the graphite and ZrH, slugs on the remote

~milling machine.

Specimens from the ARE serpentine fuel tubes
were examined. Subsurface void formation to a
depth of 3.5 mils was found on the fuel side.
The sodium side showed a slight mass transfer
deposit and some subsurface void formation.
A specimen taken from the inlet to the fuel-to-
helium heat exchanger showed subsurface voids
to a depth of 4 mils.

Out-of-pile creep tests dre under way to provide
for comparison with data obtained-in the MTR.
Difficulties with temperature control were ex-
perienced during the MTR irradiations, and the
apparatus is therefore being modified.

Leaks of undetermined origin caused three
stress-corrosion experiments in the LITR to be
discontinued, and therefore an apparatus is being
constructed that will be instrumented to check
the point of failre. Thermal stresses that
occurred during irradiation may have caused
opening of brazed joints or tubing connections.

The forced-circulation loop which was operated
in a vertical hole in the LITR for 235 hr is being
disassembled. The supposition that the pump
became inoperative because of bearing seizure
was substantioted. No significant amounts of
ZrF, vapor deposits were observed in the pump.
The loop will be examined in detail to determine
the extent of corrosion and the performance of
of the various components.

The average thermal-neutron-flux exposures of
several static corrosion capsules irradiated in
the MTR were calculated from the Co%® dis-

~_ integration rate of the Inconel container material.
'~ The values were in substontial agreement with

those calculated from fission-product yields in
the fuel contadined in the capsules. The con-

" tribution. to the Co%% disintegration rate in

irradiated Inconel arising from the nickel con-

. stituent through the Niw(n,;:v)Co‘so reaction was

investigated and found to be negligible.
Preparations  were continued for irradiation in

" the ETR of moderator materials for use in high-

temperature reactors. Fabrication of test speci-

. mens was started.

Hot-pressed boron nitride, whlch is bemg con-

 sidered for use as o neutron shielding material,

was fested and found to be undamaged by
irradiation at a temperature of 1800°F to a thermal-

neutron dosage of 1.3 x 102% neutrons/cm?.

%

 
 

 

I

SERE——

Post-irradiation examinations of two hot-pressed
boron carbide tiles are underway. Plates of BN-Ni
(10.3 wt % BN) and CaB ;-Fe (7.6 wt % CGB‘)
cermets clad with type 304 stainless steel were
irradiated in the LITR. There were no structural
changes as a result of irradiation to an average
B burnup of 19%, but both . materials were
damaged by irradiation to an average burnup
of 38%. Stainless-steel-clad B JC-copper cermets
were also irradiated. Damage to these samples
was found to be related to burnup, temperature
of irradiation, and the thermal-cycling history.

The general results that have been obtained
in the investigations of the effects of nuclear
radiation on semiconductor devices indicate that
the barrier phenomenon should be studied under
- closely controlled conditions, and the equipment
for such a study is being developed. Experimental
results obtained from gamma irradiation of grease
coated and uncoated point-contact germanium
diodes were analyzed, and curves describing the
behavior were prepared. A preliminary experiment
was performed with transistors to determine the
actual amount of radiation domage introduced
into a sample by bombarding the sample at liquid
nitrogen tempercture. The damage observed was
somewhat different than that of samples bombarded
at room temperature, but refinements must be
made in experimental techniques before con-
clusive results can be obtained.

PART 5. REACTOR SHIELDING
5.1. Lid Tank Shielding Facility

The latest and best data are presented for
attenuation from a fission source of fast neutrons,
thermal neutrons, and gamma rays in water,
borated water, and o hydrogen-saturated oil.
The thermal-neutron flux data presented are
corrected for flux depression by the foil detectors,
as well as for self-absorption and self-shielding
of the detectors.

The first test in a series of tests for investi-
gating the shielding properties of Hevimet,
stainfess steel, and lithium hydride preceded
by a beryllium reflector-moderator and backed
by borated water has been completed. The
thermal-neutron flux measurements beyond the
various .configurations are presented. Gamma-ray

and fast-neutron dose rates were presented
previously.
XXX

] W

penetration of lead and water shields.

5.2. Bulk Shielding Facility

In the continuing study of methods of reducing
capture - gamma-ray production in shields, the

effectiveness of distributing lithium throughout
a lead shield was investigated.

It was found
that the lithiated lead was approximately. as
effective as boral-covered lead.

5.3, Tower Shielding Reactor-ll

Nuclear calculations and development experi-
ments have been continued in order to fix the
design of the Tower Shielding Reactor-11 (TSR-I).
An additional feature in the mechanical design
is being investigated which will allow a beam
of radiation through the shield of the reactor -
to sweep both a vertical and a horizental plane.
On the basis of the design now anticipated,
calculations have been performed for predicting
the gamma-ray heating of the TSR-Il shield.

5.4. Shield Design

A Monte Carlo calculation of the energy spectrum
and angular distribution of normally incident
0.662-Mev gamma rays which are transmitted
through 3-, 6-, and 9-in.-thick cluminum slabs
has been performed on the Oracle. Plots of
the transmitted collided energy fraction have
been fitted with analytical expressions.

A total of 512 problems has now been computed
in the Monte Carlo calculation of the gamma-ray
Typical
results of the heating of pure lead slabs 4 mean
free paths thick are given for 3-Mev gamma rays
incident on the slabs at .angles of 0, 60, 70.5,
and 75.5 deg.

In the Monte Carlo calculations for lead and
water shields, dose-rate buildup factors were
also calculated for normally incident gamma
rays on all-lead slabs and on all-water slabs,
as well as on composite slabs of the two materials
having varying percentages of lead content and
totdl composite slab thicknesses of up to § mean
free paths. The values of the buildup factors
determined for the pure matericls in these cal-
culations have been compared with the values
obtained for infinitely thick slabs with the NDA
moments method.  Those determined for the
composite slabs have been compared with values
obtained by the use of formulas in which the
buildup factors independently computed for lead
and water are combined to determine the value
for a composite slab.

 

 
 

 

Part 1
AIRCRAFT REACTOR ENGINEERING
S. J. Cromer

 
 

 

L

 

 

[

 
 

1.1. REACTOR AND FACILITY CONSTRUCTION

W. F. Boudreau

METHODS DEVELOPMENT
G. D. Whitman
C. K. McGlothlan G. W. Peach
Weld Tests

Weld shrinkage and distortion tests, as previously
described,’ have continued, with approximately
50 tests having been completed. Shrinkage tests
were completed on the equatorial weld joints in
the two /-m.-thlck core shells that form the
reactor fuel annulus, and a nominal value of 0.060
in. was established for transverse shrinkage. The
weld tests were also completed for the redesigned

 

'G. D. Whitman et al., ANP Quar. Prog. Rep. March
31, 1957, ORNL-2274, p 67.

 

A. P. Fraas

pressure shell equatorial joint, and a nominal value
of 0.200 in. was established as the allowance for
transverse shrinkage. Work in progress during one
of the pressure shell weld tests is shown in
Fig. 1.1.1. | '

Additional tests are under way to determine the
weld shrinkage ond acceptability of the joint de-
signs of the equatorial welds on the '4 -in.-thick
outer reflector shell, the /16-|n.—th|ck shielding
cover shells, and the é-m.-tl-nck pressure shell
liner. It was found that improved results could be
obtained in making the equatorial butt weld for the
pressure shell liner by reducing the metal thick-
ness at the land so that the root pass could be put
in with a lower welding current. Improved results
were obtained on the other special closure welds

UNCLASSIFIED g
PHOTO 28671

Fig. 1.1.1. Pressure Shell Weld Test.

 
 

 

ANP PROJECT PROGRESS REPORT

by design modifications, including changes in

dimensional tolerances, and the development of
new welding procedures.

Equipment Inspection

-Main Fuel-to-NaK Heat Exchanger Checkmg
Fixture. — A special checking fixture was designed
and fabricated to be used to dimensionally check
the main heat exchanger tube bundles and to thus
establish whether they can be assembled into the
reactor. The fixture is shown in Fig. 1.1.2, with

 

 

Fig. 1.2 Fuel-20-NoK Heot Exchanger Checking
Fixture with Wooden Mode! of a Tube Bundle Positioned
for Checking.

a wooden model of a tube bundle positioned for
checking. The spherical section of the fixture
establishes the outer surface of the reflector
assembly, and the pins located in this spherical
section define the area into which the tube bundle
channel must fit. The dummy header blocks at the

top and bottom define the parting planes between
the headers, and the notched rings at top and
bottom establish the nozzle positions.

Two heat exchanger checking fixtures of this
type, which were fabricated by the Bunell Machine
& Tool Company, were available in June; one was
delivered to ORNL and the other was delivered to
the heat exchanger fabricator, Black, Sivalls &
Bryson. Since then it has been necessary to return
the units to the Bunell Machine & Tool Company
for reworking, because the header end plates have
been redesigned to give more cleurance between
the headers.

Reactor Component Inspectlon. ~ Special tools
were designed and built to dimensionally check
the B,C tiles required for neutron shielding. An
lnspechon gage designed for checking the chord
fength of the various pieces is shown in Fig.
1.1.3, and a device used for inspecting the top

~and bottom edges of the tiles, which were machined

as conical surfaces, is shown in Fig. 1.1.4,

UNCL ASrND
PaTo W7

 

 

|llll|||||lr'|v|1|||||ll1|!
‘ INCHES

Fig. 1.1.3. Boron Curblde Tile Chord-Length ln-»
spection Gage.

Mechanical inspection forms have been made up
for all the reactor shells, the filler plctes, the
north head, the load ring, the beryllium sections,
and other miscellaneous parts. For the most part
these inspection forms are scale drawings of parts
or assemblies with lists of the critical dimensions
which must be recorded for assembly fitup or for
general information.

|2

 
 

 

Fig. 1.1.4. Boron Carbide Tile Coniecal-Surface In-
spection Gage.

COMPONENT FABRICATION AND ASSEMBLY
Main Fuel-to-NaK Heat Exchanger
J. Zasler R. B. Clarke

An experimental unit of the main fuel-to-NaK
heat exchanger was completed at Black, Sivalls &
Bryson and shipped to Oak Ridge. The manufacture .
of this unit, which is shown in Figs. 1.1.5, 1.1.6,
and 1.1.7, has proved all the major fabrication
methods and established the feasibility of monu-

facture of the heat exchanger. This experimental

unit was completed prior.to the assembly of a

prototype unit through the use of partially- mSpected
material and some imperfect parts.

The seventh channel  preduced was: machmed'

successfully to within dimensional tolerances and
is ready for assembly into the prototype unit. A

modification was made to the fool-octuafmg mecha- -

nism on the large boring mill used for channel
machining to allow the tool to rock as the channel
moves past it. The tool thus operates at a constant
angle as it is moved vertically. All parts have

PERIOD ENDING SEPTEMBER 30, 1957

been completed for the prototype unit, and assembly
can be started. A change in the method of fabri-
cating the header assemblies has been adopted
which eliminates several operations. As a result
of the experience gained in the brazing of the
experimental unit and in additional test work, it
has been decided that it is no longer necessary
to hand-clean the oxide from the tubes after they
are hot-formed in the resistance-heating fixture.

Sodium-to-NaK He_gt' Exchungers
W. S. Harris

The first and second sodium-to-NaK heat ex-
changers, which were delivered during the previous
quarter, were successfully installed in the ETU
north head. It was determined that meodifications
could be made that would facilitate the installation
of subsequent heat exchangers, and the modifi-

-cations are being incorporated in units 4 and 5.

During the fabrication .of unit 3 it was determined
that a dented fube could be cut out and replaced
successfully.

ART-ETU Radiators
M. M. Yarosh

Both of the radiators required for installation in
the ETU facility were received in Oak Ridge during
the quarter. The first unit, received during the
week of July 22, was inspected and then installed
in the facility. The second unit was received the
week of September 2, and inspection was started.

As had been somewhat anticipated, fabrication
difficulties were encountered in the production of

- these early units. "It was found possible to intro-

duce corrective measures both to bring the units

. up to an acceptable standard for their intended use
and to ellminafe the. dlfficultles from later pro-
_ductlon. o

" Distortion of the first umf during assembly

“welding resulted in cracks in the back-braze alloy
 of 69 of the tube-to-header joints. The unit was

repaired by placement of additicnal alloy ot the
back-braze joints and putting the unit through a
second brazing cycle.  All cracks were filled by
the ddditional brazing operation.

~ During the fabrication of the second unit three
tubes were damaged as a result of distortion of
the unit during assembly. An assembly sequence
was worked out to eliminate such troubles in future

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

l HOTE 42032

 

 

Fig. 1.1.5. Two Views of Experimental Unit of Main Fuel-to-NaK Heat Exchanger.

i

%

H

 
 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

G HRRPIER
PHOTO 42033

 

| CONFIDENTIAL U

i

 

Fig. 1.1.6. Two Views of North Header of Experimental Unit of Main Fuel-to-NaK Heat Exchanger.

 
 

 

 

 

 

LB A R e e i

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

Fig. 1.1.7. Two Views of South Header of Experimental Unit of Main Fuel-to-NaK Heat Exchanger.

 
 

 

 

 

radiator assemblies. This procedure is being em-
ployed on the first ART radiator assembly and
appears to be working successfully. 7

It was decided that the brazing operation could

be improved if the vendor was supplied with the
type of brazing rings previously produced at Oak
Ridge in pilot quantities. A facility for producing -

brazing rings was therefore set up at ORNL, and
ring production of approximately 10,000 per hour
was achieved. Through the week of September 6,
a total of approximately 750,000 rings was pro-
duced.

Main Pressure Shell and Liner
G. D Whitman -
C. K. McGlothlan G. W. Peach

The first pressure shell lower forging was re-

ceived from the Ladish Co. and is shown in Fig.
1.1.8. This ETU part is to be machined in the

B WNCLASSIFIED
PHOTO 29411

Fig. 1L.L8&, 'Pressure Shell Forging. -(-Conhd-enfid
-Il ll ) : .

PERIOD ENDING SEPTEMBER 30, 1957

Y-12 shop, and tooling for this operation is being
fabricated. The part is scheduled to be compleied
by April 1958.

It has been decnded to fobricate the dormers in
the upper pressure shell as separate components

. and weld them intc the shell. This work is to be

done in the Y-12 shops. . It is anticipated that the
first pressure shell upper forging will have been
received at Oak Rldge before the end of the report
period.

The lower pressure shell liner half is to be
forged and machined by the Ladish Ce. This work
is currently in the design end tooling fabrication
stage. The first shell half is scheduled to be

_struck in October 1957.

The upper pressure shell liner is to be fabri-
cated from a cold-formed weldment and will be
welded to the north head assembly before final
machining so that accurate dimensional control
can be maintained. This first weldment has been

fabricated by the Steel & Alloy Tank Co. and has

been received.
Beryllivm Reflector-Moderator Outer Shell
G. W. Peach

The beryllium reflector-moderator outer shell
(shell 11l) is being fabricated from a machined
weldment. A cold-formed weldment of the lower
half of the shell is shown in Fig. 1.1.9 as received
from the Steel & Alloy Tank Co. This lower half,

which is nearly complete, is shown again in Fig.

1.1.10 during the machining operation. No par-
ticular ' difficulty has been encountered in this

- ‘operation, and the design tolerances have been

held.

A shell Il upper weldment has been received
-and will be welded to the strut load ring assembly

before machining so that weld shrinkage and dis-

" “tortion can be controlled and the finished shell
moy be oecurofely positioned in the assembly.

Boron Shield Contemers - Shells IV ond V

G W. Peach
The ‘method for manufacturing shells IV and V,

" the boron shield - containers, has been changed
from shear spinning (Hydrospin). to deep drawing.
. Experimental draws have been made on steel and
‘Inconel “sheet by forming 35-in. hemispheres in

order to evaluate the forming problems. These
tests substantiated the practicability of the method.

 
 

ANP PROJECT PROGRESS REPORT

UNCULASSIFIED
PHOTO 29088

 

 

o

UNCLASSIFIED
PHOTO 29257

R ’ g N »

§
i
i
i

 

Fig. 1.1.10. Shell lll Being Machined. i i t

10

 
 

Tool engineering is now under way for the manu-
facture of shells IV and V by this method.

The application of the deep-draw process to
these pieces is an unusual one because of the
large ratio of part diameter to thickness. Con-
ventional drawing practice'_indicdies a maximum
diameter-to-thickness ratio of 100/1.  The parts
in question are 0.085 in. thick and 45 and 50 in.
in diameter, and thus the diameter-to-thickness
ratios are 500/1 and 600/1, respectively.  This
involves a considerable extension of conventional
limits. The principal difficulties encountered in
the early stages of the test program were buckling
in the equatorial region of the hemispheres because
of insufficient blank hold-down pressure and rupture
at the center of the blank from excessive tensile
stress on the blank. . '

 

PERIOD ENDING SEPTEMBER 30, 1957

Equipment used for the experimental drawing
operations at Kaiser Metal Products, Inc., is
shown in Fig. 1.1.11, in which may be seen the

‘punch, the draw die, and the hold-down ring of the

2200-ton hydraulic press. A typical rupture defect
on a 0.074-in.-thick, 35-in.-dia steel shell caused
by excessive punch pressure is shown in Fig.
1.1.12. A typical pucker defect in a 0.062-in.-
thick, 35-in.-dia Inconel shell is shown in Fig.
1.1.13. Specimens of shells formed by three dif-
ferent methods are illustrated in Fig. 1.1.14. The
specimen at the rear of Fig. 1.1.14 is 0.074-in.-
thick, .35-in.-dia cold-rolled steel that was formed
without backing material. As may be seen, the
material puckered. The middle specimen of 0.074-

in.-thick, 35-in.-dia cold-rolled steel was backed

with 0.132-in.-thick hot-rolled steel during forming.

<UNCLASSIFIED
. PHOTO 41303

g

Fig. 1.1.11. Hydraulic Press for Deep-Draw Process.

11

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

The slight pucker is not evident. The front speci-
men of 0.074-in.-thick, 35-in.-dia cold-rolled steel
was backed with 0. 169-in.-thick hot-rolled steel
No defects were present.

 

 

— 5o

Fig. 1.1.12. Ruptured Shell Produced by Deep-Draw
Process. Cenfidertict—witircoptien)

UNCLASHFIED
PHOTO 4194

 

Fig. 1.1.13. Deep-Drawn Shell Showing Pucker
Defect. (Eonfidential-with~eaptian)

12

 

Fig. L1114, Deep-Drawn Shell Test Specimens.
(Conftdanriotwithtopten)

The experimental work showed that by reinforcing
the blank with backup steel sheets, the desired
hold-dewn pressures to avoid puckering could be
maintained without rupturing the blank. The best
results were obtained on 35-in.-dia hemispheres
with 0.062-in.-thick welded Inconel blanks re-
inforced with ‘r’/sz-in.-fhick hot-rolled steel sheet.
There were indications that good results could
be obtained wnth less remforcmg but ‘that this
would require ‘‘cut-and-try’’ balancing of hold-
down and press-punch pressures.

The decision to discard the shear spinning tech-
niques for manufacturing shells IV and V was
based on the following:

1. Try-out work demonstrated that adherence to
the spinning mandrel would be very difficult and

- perhaps impossible to maintain.

2. In order to even approach adequate dimensional
control it would be necessary to preform and pre-
machine the blanks.

3. An extensive amount of development work
would still be necessary to establish the most
satisfactory shape for the spinning rolls.

 
4. Intermediate annealing steps’''might distort
the parts and make them d|ff|cult to reullgn on the
machine. '

5. Inconel stock requirements appedred to be
higher than originally anticipated and it was un-
certain that material costs could be held within
reasonable limits.

While it is felt that fhe spmmng process has
very definite advantages in ultimate quality of
reactor parts, much basic knowledge of the forming
process needs to be developed to overcome the
difficulties experienced with ART parts.

Inner and Outer Core Shells
G. D. Whitman G. W. Peach

As previously reported the inner core shell and
island assembly for the ETU has been completed.
Four more inner core shell weldment halves have
been received, and two of these weldments are
shown in Fig. 1.1.15. Upper and lower halves of
an outer core shell weldment have also been
received, and machining operations have been

 

PERIOD ENDING SEPTEMBER 30, 1957

started.” The lower-half weldment prior to machining
is shown in Fig. 1.1.16. Machining operations are
50% complete on the lower half of the outer core
shell and 15% complete on the upper half of this
shell. These parts are stress relieved after the
rough machining operation so that distortion in the
final machining operation is minimized. The shear
spinning process for the production of these and
other shells was cbandoned because of the dif-
ficulties described previously and because satis-
factory forging, welding, and machmmg techniques
have now been developed.

Reactor North Head
C. K. McGlothlan

The two sedium-to-NaK heat exchangers received
from the Griscom-Russell Co. have been fitted into
the ETU notth head, and welding operations have
been started. The heat exchangers are shown in
place in the lower deck weldment in Fig. 1.1.17.
In Fig. 1.1.18, one deck plate may be seen in
position for welding.

 

Fig. 1.1.15. Inner Core Shell Weldment, TEome
fidontialawith ciom)

Fig. 1.1.16. Outer Core Shell Weldment, Lower Half.

13

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

Fig. 1.1.18. North Head Assembly with One Deck Plate in Position for Welding. Secrot-with-capntion)

14

 

1,

ty
 

Three lower deck skirt forgings have been re-
ceived from the Ladish Co., and rough machining
has been completed on the ETU part. The forging
is shown as received in Fig, 1.1.19'and during the
machining operation in Fig. 1.1.20.

Three rough-machined fuel-expansion-tank tops,
forged from bar stock, have been received from the
Ladish Co., and one of these parts is shown in
Fig. 1.1.21.

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

The installation and successful testing of a
facility for producing dry hydrogen gas for a
reducing atmosphere for the stress-relieving oper-
ations on the north head assemblies has been
completed. A trial run was made in which the
north head retort was filled with scrap Inconel
parts.  The scrap items were all successfully
cleaned at 1850°F. The stress-relieving retort
installed in the furnace is shown in Fig. 1.1.22.

Strut Lead Ring
W. E. Thomas

The strut and ring assembly has been welded to
the load ring forging. The inside contour of the
load ring has been finish-machined in the passage
vnder the conical divider and requires one more
finish cut on the other surfaces. Inconel cover
plates for the copper-B,C cermet layer on top the
load ring have been formed and are ready for
installation. All sodium return possages have
been drilled. This assembly is shown in Fig.
1.1.23. Two conical dividers have been fabricated
and machined. One of the dividers has been welded

UNCLASSIFIED
HOTO 29:

 

Fig. 1.1.20, Lower Deck Skirt Being Machined.

15

 
 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
PHOTO 29702

 

Fig. 1.1.21. Fuel Expansion Tank Top, Rough Machined. -{Confidential with caption)

 

UNLL AYWT I:O
AL e,

Fig. 1.1.22. North Head Stress-Relieving Retort
Installed in Furnace.

16

to the pan face—pan rim—sodium inlet pipe sub-
weldment and heat treated, and it is now ready for
assembling in the load ring.

Neutron Shielding
C. K. McGlothlan

Boron Carbide Tiles. — All the tiles, 3386 pieces,
consisting of 21 shapes, have been received from
the Norton Company, and they have all been
inspected and accepted. A total of 1251 pieces
is required for each reactor assembly, and therefore
the total quantity on hand includes a number of
spare pieces. A hot-pressed spherically shaped
tile blank and the finished tile produced from such
a blank are shown in Fig. 1.1.24, and one of the
shapes produced for the load-ring assembly is

_shown in Fig. 1.1.25.

Cons and Lids for Boron Carbide Tiles. ~ The
cans and lids for the tiles are formed from 0.0035-
in.-thick type 430 stainless steel and an inter-
mediate layer of 0.0015-in.-thick copper foil. These
materials have been fabricated by General Plate

Div. of Metals & Controls Corp. and delivered to
the ERCO Division of American Car & Foundry

Co., for fabrication of the cans and lids. The
tooling for this job is 75% complete. A total of

>

 
 

PERIOD ENDING SEPTEMBER 30, 1957

Fig. 1.1.23. Strut Load Ring Assembly.

 

UNCLASSIFIED
PHOTO 29703

 

UNCL ASSIFLED
PHOTO 29260

 

 

 

FT T T T T T T T T T T T T

2 - 3
INCHES

o 1

Fig. 1.1.24, Boron Carbide Tile Blank and Finished Part.

4

S

&

 

17

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

UNCLASSIFIED
PHOTO 29261

 

 

 

2 3 4
INCHES

,lllllll,1|l|||1'||llrlli||||'|.|.l.‘; 'l‘l'l"

 

5434 pieces is required for the ETU and ART

Fig. 1.1.25. Boron Carbide Tile for Load Ring Assembly.

involve 48 different shapes.

Work is progressing

 

reactors; the first 6 pieces have been received and
are being inspected. One of the tile, can, and lid
assemblies is shown in Fig. 1.1.26.
Stainless-Steel-Clad Copper-B,C  Cermets. -~
Approximately 15% of the rolled cermet sheet
material being fabricated by Allegheny-Ludlum
Steel Corp. has been completed ot a price of
$1.07/in.2 for 0.100-in.-thick material. An esti-
mated 80,000 in.2 will be required for the entire
job which includes a small amount of 0.250- and
0.312-in.-thick material ot a cost proportional to
the increased thickness. The ERCO Division of
American Cor & Foundry Co. is fabricating the
required cermet shapes from the rolled cermet
sheet material. Tooling for this job is 95% com-
plete. Seventy fabricated pieces have been re-
ceived and accepted. The 1050 pieces required

18

satisfactorily on this job. The controlling factor
in the completion of this order is the rate of
cutting of the material by the Elox (electrical-
discharge) process. _

The electrical discharge method has been the
only practical way found to trim the cermet parts
accurately, and the work has been scheduled on
a multishift basis to expedite delivery of the parts.
Some of the cermet shapes required in the load ring
assembly are shown in Figs. 1.1.27 and 1.1.28.

Lead and Tungsten Shielding
A. M. Smith

Details of the cooling coils in the equaterial
section of the lead shield have been determined,
and the design of the steel inner shell support
 

 

 

"

 

Fign ]. 1! 27.
Plates.

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
PHOTD 29796

 

 

 

 

I l'l'lllll'l'_‘_'"|rT“"
o 1
INCHES

Fig. 1.1.26. Tile, Can, and Lid Assembly.

520 ANGLASSIFIED
B9z

 

 

 

Load Ring Cermet and Inconel Cover

assembly has been modified by increasing the
number of support ribs to reduce the stress level
at operating conditions. Changes have been made
in the method of welding these ribs to the shell to
avoid buckling problems. A vendor is now fabri-

‘cating this support assembly.

Work is continuing on the design of the north
and south lead shield sections. It has been decided
to support the tungsten shielding from the NaK
manifolds, and the prehmmary layouts have been
developed

P

Miecelluneous Components
A M. Smith ¥. E. Thomas

Reactor Spacer Rings.,- Parts for two spacer
ring assemblies have been machined. One set of

- parts that is ready for ossembly is shown in Fig.

7'1 1.29.

Recctor Filler Plates. -Two sets of flfler plates

f'are in various stages of completlon._. .One design
- change was made to prov;de a better flow profile.

The ETU filler plates, with the exception of plate

~No. 6, have been finish-machined on the mating
‘surfaces, the sodium passages hove been cut, and-

‘the bolt ond dowel holes have been drilled. The

plates have been bolted together and set up for
contour machining of the inner and outer surfaces,
as shown in Fig. 1.1.30. Plate No. 6 for the ETU

19

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
PHOTO 29695

 

 

 

T T T T T T T T
. 1 2

4 5 6

 

INCHES

 

Fig. 1.1.28. Load Ring Cermet Pieces.

 

UNCLASSIFIED
PHOTO 29741

Fig. 1.1.29. Spacer Rings.

has been finish-machined on the inner surface
only. All ART filler plates, except No. 6, have
been finish-machined on mating surfaces and the
sodium passages have been cut.

NaK Manifolds. — Design of the NaK manifolds

for the ETU has been completed, and the drawings
were transmitted to the vendor. Several minor

20

changes of ART manifold drawings are being
mede. Sufficient material for the manifold nozzles
was obtained despite the unusually close dimen-
sional tolerances specified as a means of reducing
stresses.

It is presently excepted that a prototype of one
of the ETU manifolds will be fabricated by October
1, 1957.

 
 

 

i
i
i
i
i
i
i
i
i

 

(]

 

 

PERIOD ENDING SEPTEMBER 30, 1957

UNCL ASSIFIED
PHOTO 29704

Fig. 1.1.30. Filler Plate Assembly.

ETU Cell end Accessories. - Layouts o'f-f-he-

equipment to be installed inside the cell are esti-
mated to be 75% complete. A ‘‘negater’’ spring
arrangement to aid in support of the dump valve

actuators has been ordered. Preliminary drawings

have been prepared of the mockup shielding and
of the balconies or platforms that will be provided

within the cell.  Fobrication of the fuel fill-and-

drain tank supports was completed.  Other equip-
ment such  as . ZrF -vapor trops, ‘the auxiliary
sodium exponsion tank and its associated piping,
and the fuel overflow - line are being designed.
Estimates of cost aond material needed for the

fabrication of the various units have been made. |

~cold-positioning the island.

REACTOR ASSEMBLY
G. D. Whitman W. E. Thomas
~ The top and bottom neutron shielding cans for

the top of the island have been fabricated and
assembled for installation in the ETU reactor.

‘Two design changes have been initiated which

will make final assembly easier. Welds on the
sodium expansion tank are to be displaced to a
point more distant from the sodium pump barrels
in order to eliminate the need for boring the pump
barrels after final assembly. A second change is
being made to eliminate a possible hazard while
In the previous

21

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

arrangement, a nut and bolt located between the
pressure shell and the filler plates were to be

used, but, if the nut and bolt were to become in-.

operative during assembly, it would be necessary
to grind out the pressure shell equatorial weldment
and replace the threaded parts of the support struc-
ture. The redesign eliminates the hidden nut and

bolt by providing threads on the pressure shell and

a plug and plunger arrangement that can be re-
moved and replaced externally.

E.TU FACILITY
G. D. Whitman
P. A. Gnadt A. M. Smith

Installation work on the ETU facility is pro-
gressing at the rate originally predicted. The
status of construction as of August 15, 1957, may
be seen in Fig. 1.1.31, which shows the control
room enclosure, the NaK pipe main support columns
and cell mockup, and the main piping of the furnace
and isothermal NaK circuits, with the stationary
units of the four pumps mounted at the top of the

structure. In the background just above the center

of the support steel is a portion of the main air
duct in which one radiator is already installed,
In the foreground are the two reactor support
pedestals.
cross bracing and walkways on the structure.

The induction regulators and associated conduit
installed in the basement area may be seen in
Fig. 1.1.32, and the heater distribution panels,
transformers, and associated switches may be
seen in Fig. 1.1.33, with the induction regulators
in the background. Present estimates show an
expected ETU facility completion date, exclusive
of the reactor and its associated cell equipment,
of September 1, 1958. Completion of the cell
equipment is scheduled for December 1, 1958. A
description of the construction and design status
of the facility as of September 1, 1957, is given
below.

Reactor Support Structure
A. M. Smith

Details of the reactor support platform were
completed, and fabrication was started on both
the ETU ond the ART structures. Plans have
been made for the machining of the columns and
the stress-relieving of the entire support platform.

22

The workmen are shown installing

Column footings . for both ETU and ART support _
columns hclve been completed ~

| NGK Piping, Assb‘ciufed Supporié,‘dr'lacq‘:mfipn‘enfs

P. A. Gnadt

Suppo_rf- Structures and Shielding. ~ All main
supporting steel has been set in place. Installation
of cross bracing, gratings, shielding supports, and

- hanger supports is approximately 65% complete.

Work is now in progress on this portion of the
facility. It is planned that the shielding plates
will be installed during the summer of 1958.
Design drawings are complete for this work, except
for minor modifications. The shielding plate design
is scheduled for completion early in 1958.

NaK Piping. — The installation of the furnace
circuit main piping is complete to the cell wall.
Pumps, furnaces, throttle valves, and flowmeters
are installed. One cold trap has been installed.
Present plans are to install the remaining cold-
trap circuit, thermocouples, liquid-level-measuring
devices, fill-and-drain tanks, and associated lines
and valves after completion of the auxiliary and
isothermal NaK-circuit main piping. All material

for this work is on hand, except the drain valves,

drain tenks, and continuous-level-measuring de-
vices. The drain valves are scheduled for delivery
late in 1957. The drain tanks and continuous-
level-measuring devices are scheduled for delivery
October 1, 1957.

The main piping for the two isothermal NaK
circuits is 95% installed. Pumps, throttle valves,
and one flowmeter are in place. The status of
material for the remaining work is the same as
noted above for the furnace circuits.

A radiator has been installed for one heat dump
circuit. The remaining radiator is scheduled for
shipment approximately September 1, 1957, and it
will be installed in the air duct as soon as it is
received. The two pump bowls for these circuits
are scheduled for delivery September 15, 1957.

The main piping for these two radiator circuits
is preformed ond will be installed as soon as the
pump bowls and remaining radiators are delivered.
The flowmeters are on hand and will be installed
with the pump bowls and the radiators. The status
of the cold-trap circuits, thermocouples, level
devices, drain valves, and drain tanks is the same
as mentioned above for the furnace and isothermal
circuits.

 
 

 

PERIOD ENDING SEPTEMBER 30, 1957

o
w
w
w3
w
<
—
Q
z
3

PHOTO 29616

 

et 5T T

R
oy

% WWH

 

 
 

 

 

7. (Seeret-with-saptiom

195

Status of ETU Facility Construction as of August 15

1L.3L

1.

Figu

23
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

Fig. 1.1.32. Induction Regulators and Associated Conduit Installed in Buselfienf Area of ETU Facility, {(Gem

 

 

 

 

Fig. 1.1.33. Heater Distribution Panels,
of ETU Facility, (Confidertiel-with-copriom

24

-

 

 

 

 

 

 

 

 

 

AW UNCL ASSIFIED
2 Fho

 

 

Transformers, and Associated Switches Installed

TO 29613

in Basement Area

%

f

 
 

 

 

 

The cold-trap and plug-indicator circuits for the
six NaK systems are prefabricated and ready for
welding .into the main piping circuits. It has been
decided, however, that the plug-indicator circuits
can be eliminated, and they will not be installed.

Main Air Duct. — All major components for this
system are on hand. The section of the duct for
housing the two radiators is installed. The re-
maining sections will be installed following the
installation of the radiators.

Electrical Equipment
P. A. Gnadt

The emergency electrical system, consisting of

the 300-kw diesel generator unit and associated
switchgear and supply cables is 95% complete.
Design work for the normal electrical distribution
system is 90% complete, and will be 100% com-
pleted by November 1, 1957. All completed design
drawings were released to a cost-plus—flxed fee
contractor on June 1, 1957. :

The contracter has installed all vohoge regu-
lators, transformers, motor control centers, heater
distribution equipment, and 85% of the cable trays
in the basement area. Installation of cables be-
tween the induction regulators and the distribution
cabinets is approximately 50% complete. Accept-
ance tests for approximately 125 variable trans-
formers revealed excessive winding temperatures
at the specified ratings. These units are being
returned to the vendor for modification or replace-
ment. These variable transformers are for use as
voltage controllers for o portion of the heater
circuits. The cost-plus-fixed-fee contractor has
completed the control room enclosure, installed

the cable trays ubove the control room; installed
all conduit entrances to the control room through

the roof and floor, and is presently setting the

steel support structures for the NaK motor con-

trollers, resistors, and associated cable troughs.
The variable-frequency drives for the reactor

pumps have been ordered, and delivery is expected
late in calendar year: 1957 ‘The motor-generator

sets ussocmted with the pump drive motors will
be ‘installed by the cost- plus-flxed fee contractor.
The pump’ drive units will be installed after the

reactor - and pumps have: been mstulled in the_'.-

facility.

De5|gn ‘work on pcmel onouts ‘and controls has ‘

been started by the Instrumentation and Controls
Division. Preliminary ponel layout drawings and

PERIOD ENDING SEPTEMBER 30, 1957

a set of preliminary block diagrams describing
‘contro} functions have been prepared. The control
panels are to be available by approxlmately August
1, 1958.

Details of the heater designs are bemg prepared,
.and installation and wiring-connection drawings
for the Calrod ond clamshell heaters are to be
completed by December 16, 1957. Heaters and
surface-type thermocouples will be installed after
completion of the NoK piping assembly.

Auxiliary Services
A. M, Smith

Piping drawings for the auxiliory services
(helium, lube oil, air, and water) in the basement
area have been revised. The installation work for
this piping will be done by a cost-plus-fixed-fee
contractor. The removal of some of the existing
pipe and installation of the modified pipe runs were
delayed until August 28, 1957, to allow the elec-
trical forces to complete their overhead work in this
area. Layouts of the auxiliary piping above the
track floor have been prepared. A preliminary bill
of material for items required on this part of the
system have been prepared, and procurement of

some items is now in progress.

After considerable difficulty in obtaining a leak-
tight system, the type R lube package (Figs.
1.1.34 and 1.1.35) delivery was made on June 28,
1957. This equipment is stored in the basement
area and will be installed by the cost-plus-fixed-
fee contractor. Failure of the subcontractor to
deliver leak-tight pumps to the lube package vendor

- has delayed the completion of the type K lube

package. Delivery of this package is now scheduled

“for October 1, 1957,

ART FACILITY
F. R. McQuilkin
Design Achvntyf |

Ma|or design work on components ‘outside the

. cell of the ART facility is to be completed by
 April 1, 1958. The work outlined by these designs

is to be done by ORNL forces. Design work is

‘proceeding on the special equipment room (equip-
.ment for NaK coolant system fuel fill-and-drain
tank), the radiator pit, and the radiator-penthouse
- area (NaK coolant system equipment for fuel and

sodium systems). Incomplete l/‘z-scclle models of

25

 
 

 

 

 

 

ANP

26

PROJECT PROGRESS REPORT

———
o =

 

 

 

Fig. 1.1.35. Type R Lube Package for ETU Facility.

i@l unCLASSIFIED

PHOTO 29614

n._..
ittt e

 

 

 

 

©

#

 
 

 

 

 

these areas, which are being made;to assist in
detail design and to demonstrate equipment, as-
sembly, and change-out procedures, are shown in
Figs. 1.1.36, 1.1.37, and 1.1.38.

Figure 1.1.36 is an elevation view taken looking
southeast at the radiator pit. The circular pattern
in the lower left corner represents the cell; the
special equipment room is in the background. The

air duct, which has not been assembled, will be

located in the upper right corner above the radiator
pit. The tank at the right is the drain tank for the
radiator drain pans. The five vertical tanks are
the main and aquxiliary NaK dump tanks.. The
horizontal tank is the special NaK dump tank.
The equipment structure to the left .of the drain

tank ‘is the NaK purification system for the main

and auxiliary coolant system and consists of the
cold traps, plug indicators, flowineters, valves,
and piping. The rack will be revised to omit the
plug indicators. In the ceiling of the radiator pit
are the main and duxiliary NaK piping, with the
flowmeters and the heat-barrier-door operators.
Figure 1.1.37 is a view looking down into the
radiater pit with the air duct removed. The cell
wall is at the left; the NaK piping and off-gas
penetrations may be seen. The smaller boxes
represent the NaK flowmeters and the larger boxes
represent the heat-barrier-door operators. - The hot
NaK piping penetrations for routing to the radiators
are located between the heat-barrier-door operators.

Figure 1.1.38 is a view looking south from the

cell into the radiator pit at the right and the
special equipment room at the left. The special
NaK pump and motor, along with the motor cooling

duct, are located at the top. The two large tubes -
at the center of the special ‘equipment room-are .
the spectrometer tubes. Below the tubes in the
foreground is located the special NaK pUl'lflCGhOfl.
system.” The. specml heat dump annulus and main
"blowers, wnth trcmsmon ond ducf are’. locoted in
S ,the background ' - S
" The major |tems of plping and eqmpment for fl'le'_ =
specml equnpmenj‘ room have been ossembled A
prehmmury demonstrahon by use. “of “this model,--
indicates - fhat “the reqmred equipment change-out
funchons can -be accompllshed Meanwhile the
equipment _is belng rearranged in the radiator pit -
“to provrde befler uccess for eqmpment mmntenance_

and change-out.
Design work is in progress on the cell-evucucflon
system, main blower back-flow louvers, smoke-

PERIOD ENDING SEPTEMBER 30, 1957

generator equipment, water filter, control-room
lighting, and additional switchgear. Detailed
schedules for design and construction of package 3
(all werk that is remaining exterior to the cell)
are being prepared. The remaining capital work,
which is scheduled for completion by June 30,
1958, consists primarily of additions and medifi-
cations to electrical facilities, cable trays, and
instrument-room air conditioning. Shop fabrication
of noncapital items has been started in preparation
for construction work to begin at the facility

" November 1, 1957. ltems scheduled to be started

at an early date consist of the NaK purification
equipment, the NaK dump tank installation, valve
rack equipment, air duct structure, and NaK piping
in the special equipment room and radiator pit.

Construction Progress

Lump sum contract work on the ART facility has
been completed. This work was contracted in four
stages, as determined by the design program. The
first stage, package 1, included major building
alterations, additions to existing buildings, cell
installation, installation of main air duct, instal-
lation of 1500-kva substation, and installation
of 480-v main switchgear. The second stage,
package A, included installation of auxiliary
service piping. The- third stage, package 2, in-
cluded installation of dlesel generators and facility,
electrical motor control centers, spectrometer room
electrical system, and the spectrometer air-con-

ditioning system. The fourth stage, package 3A,
“which was completed during this report period,
~included installation of an electrical system for
- -supplying power to the pipe and equipment heaters,
" a dry-air plant and facility, NaK pump ‘motor con-
-trollers, a lube-oil fill-and-waste system, and a
: hydrauhc system for Iouver operahon

Insfnflnflon Planmng

F’rellmmary plunnlng is in progress for instal-

. lation of facnllty equipment for the ART. A revised
. schedule of time estimates and completion dates -

has been prepared and is being reviewed with
regurd to the design und procurement programs.
The - revused schedule proposes  that installation

- of Inconel piping ‘and -auxiliary equ:pmenf in -the
“main, auxiliary, and special heat dump areas should

commence during November 1957. It is estimated
that 250 man-months of craft labor will be expended

27

 
8z

 

 

 

 

Fig. 1.1.36. One-Twelfth.Scale Model of ART Radiater Pit. (Swcretr-wirhcaptiond

 

~;

140dIY SSITA20Ad L23r0dd dNV

 

 
 

\ . 0!3\K RIDGE NATIONAL LABORATCRY
LLAl“lIJ-.!l.lljlllillll?ll]?lll-lrlll?l]l?lll"l)ill

m o

 

1.1.37.

Model

 

 

 

 

 

 

of Radiator Pit with Air Duct Removed, ~(Seeret.with captien)

UNCLASSIFIED
PHOTO 41475

192
Lelil

LS6L ‘05 ¥IIWILJAS ONIANT @013 d

 

 
 

   

 

 

  

0t

 

1Y0d3Y SSIYV0dd LD33roAdd dNV

1

] ”,-‘_-' ",
X )

| 4
!
w

i

 

 

 

 

 

 
 

 

 

4

y

to accomplish the work planned ‘for fiscal year

1958.

ART DISASSEMB LY
M. Bender ; _
F. R McQunIkln, A A Abbaflello
Reactor Dlsussembly '

Methods for taking accurate measurements fhrough '

hot-cell windows were mveshgafed further. A test
performed with a low- power cathetometer gave

accuracy in an acceptable range and no appreciable -
varigtions over the surface area could be defected
These results indicated ‘that such measurement

methods should be explored in detail. Accordingly,
optical tooling was studied and greater familiarity
with these instruments and their adjustment was
obtained by attending an optical tooling course.

A design has been evolved which utilizes standard
equipment for taking measurements. of radioactive -

parts from outside a hot-cell window.

A full-scale experiment and demonstration of
optical tooling has been planned. The use of an
optically flat reference bar mounted adjacent to

and parallel with the exterior window surface will

permit aligning the instruments with adequate
precision.

the accuracy of the measurements. The preliminary
tests with a cathetometer indicate that small vari-

ations may be acceptable, but further tests will

be required to establish limits on the v::riat‘"io'ns.'
Optical tooling has the advantage of requiring no

_ space wflhm the hot cell.

' s:mulraneously
controlled for operation in a hot cell.

Tests at the manufacturer's laboratory
indicate that these optical tools are: capable of
meeting the accuracy requirements, but vcmqtlons _
in the glass of the hot-cell windows may reduce

 

PERIOD ENDING SEPTEMBER 30, 1957

Further, since the
instruments will be outside the hot cell, they will

“not become contaminated.

Cutting Methods Evaluation

 Further experimentd work was done with cutting

_ methods for hot-cell use. Two methods were tested

and discarded: the ultrasonic process proved to be
extremely slow, and the Elox process was some-
what erratic in operation on Inconel, although it

-might be .applied to hard, elre'ctrica“y conducting
.. materials.

" A Heliarc cutting torch was found to
be capable of severing multiple layers of metal
Such @ torch could be remotely
However,
the -amount of activity released would probably

~limif its use to the reactor outer shells which have
" -the greatest bulk but where the activity level is
relatively low. Other cutting equipment has been

received and is to be tested. Among these items
are a portable bandsaw and a metallizing gun for

more- complete evaluation of metal replication

techniques. A stud gun has also been received
for  full-scale testing of the cnrmdge-acfuuted
sealant injector.

Facility Investigations

Equ:pment was added to the / ,-scale model of
the hot cell, such as the reccfor supports, the
water bag, miscellaneous tools, and handling
equipment. The model has been extremely valuable
in visualization of the handling problems.

A list was prepared of the tools and fixtures that
will be required for removal of the ART from the

cell. 'Tl"le_ list includes items that must be built
~into the cell and items that are portable.

31

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

1.2. COMPONENT DEVELOPMENT AND TESTING
| ~ H. W. Savage

PUMP DEVELOPMENT TESTS
W. B. McDoneald A. G. Grindell
Bearing, Seal, and Lubricant Tests
D. L. Gray W. K. Stair?

An additional seal test was conducted to determine
the effect of increasing the reactor sodium pump
speed from 3000 to 4000 rpm, as necessitated by
an increase in the required sodium head. In a
1000-hr period at 4000 rpm, the seal leakage of
reactor pump rotary assembly was never greater
than that previously observed ot the lower speed,
and inspection revealed no deleterious effects.

A plastic catch basin was fabricated for the
reactor pump rotary assembly test, and a mechani-
cal shakedown stand was fitted with glass ports
to facilitate observation of the oil-sparging system.

At a helium flow of 2500 liters/day down the shaft

annulus, an oil carryover occurred which increased
when the oil-sparge-line gas flow was stopped.
When the shaoft ennulus flow was reduced to 500
liters/day, no oil carryover was observed when the
oil sparge gas was flowing, and oil leckage was
adequately sparged by a continuocus flow of 50
liters/day through the oil sparge line, which
effected a 23-ft vertical lift of the oil. With a
lower shaft ennulus flow of 500 liters/day and a
low oil leakage rate, satisfactory sparging was
achieved by intermittent gas flow. Work is con-
tinuing to establish a detailed sparging procedure
for ART and ETU operation.

The reactor pump rotary seal assemblies have
consistently passed the static 50-psi gas-pressure
room-temperature test which is imposed prior to
installation in a rotary element. However, the
Graphitar No. 39 seal nose mountings in brass,
SAE 1020 steel, and type 316 stainless steel rings
have been found to leak at operating temperatures
between 200 and 280°F, with no appreciable dif-
ference being noted for the different materials. It
is now apparent that further improvement in the
mechanics of assembly and more severe acceptance
tests are required.

A redesigned positioning collar of the lower
Durametallic NaK pump seal has functioned satis-

 

1Consultant from the University of Tennessee.

32

factorily in a series of tests simulating an ART
reactor cell catastrophe. Various conditions were
simulated in pressure tests of the NaK pump
bearing housing and seal assembly, including oil
system pressures from 0 to 250 psig, both with oil
and with gas in the bearing housing. No failures
occurred with gas pressures of up to 250 psig for
either stationary or rotating seal elements. In one
test with a stationary seal, a 200-psig oil pressure
was applied in the lubrication system without
leakage to the pump tank, which was at 15 psig.

The reactor fuel pump rotary element, aitered for
irradiation testing, was installed in a new gamma-
irradiation facility at the MTR. The facility con-
sists of an 8-in.-dia tube surrounded by racks of
fuel elements immersed in the reactor canal. The
header used for the experiment may be seen above
the cancl parapet in Fig. 1.2.1. After approxi-
mately one week of shakedown operation of the
rotary element, partially spent MTR fuel elements
were placed in the grid to supply the test radiation.
Operation of the experimental equipment was stopped
temporarily for repair of a defective ion chamber,
and a general inspection revealed the equipment
to be in good condition. Testing has been resumed
at the rated pump speed of 2700 rpm, and the
average dose rate is 108 rep/hr.

Aluminum North Head ther Tests
J. W. Cooke? H. Gilkey?

Fuel System. — Operation of the twin fuel pumps
installed in the aluminum mockup of the ART north
head has demonstrated effective removal of air
from the test loop over a wide range of operating
conditions. Degassing times for removal of 0.1
scfm of air ranged from 8.5 min with 3’/2 in. of fuel
in the expansion tank and the pumps operating at
2100 rpm to 1.5 min with 2 in. of fuel in the ex-

pansion tank and the pumps operating at 3000 rpm.

No ingassing occurred at speeds from 0 to 3000

rpm when the pump speeds were equal and some

variation in relative speeds could be tolerated
above 500 rpm. '

 

20n assignment from Pratt & Whitney Aircraft.

f

 
 

“

i’

 

PERIOD ENDING SEPTEMBER 30, 1957

An eight-channel pressure-recording instrument  fluctuations increased with an increase in the
was used to measure the fluctuations of the dis- pump speeds and o decrease in the expansion-tank
charge and suction pressure taps and the flowmeter  liquid level. The maximum recorded pressure
orifice taps for both pumps. As was expected, the  fluctuations at 3000 rpm and a flow rate of 780 gpm

FIED]
| NRTS-57-3919
"4

P
s L ST,

 

 

 

 

 

 

 

Fig. 1.2.1. Reactor Fuel Pump Rotary Element Irrediation Installation in MTR Canal.

33

 
 

 

 

ANP PROJECT PROGRESS REPORT

per pump, with l/ in. of fuel in the expansion tank,
were 10.5 psi for the pump discharges and 0.8 psu
for the pump suctions.

The performance of one of the fuel pumps was
comparable with that obtained in single-pump tests,
but the head obtained with the other pump was 7%
(2.5 to 3 f1) below that for the single pump at
2700 rpm and 645 gpm. This difference can be
explained by minor dimensional differences in the
impeller and volute.

Sodium System. — Fabrication of the twin sodium
pump loop for testing in the aluminum north head
mockup was completed. Preliminary data showed
that the system readily degassed with the pumps
at equal speeds above 3000 rpm and with the liquid
level in the expansion tank above the ports that
connect the expansion tank to the pump impeller
regions. The pumps shared the pumping lead

equally when operating ot equal speeds. The

loop flow resistance was considerably higher than
anticipated, with most of the higher resistance
being attributable to the sodium-to-NaK heat ex-
changer pressure drop. Examination revealed that
the high heat exchanger pressure drop was caused
by blockage of inlet and flow passages by flakes
of an epoxy resin that had been applied to the
inner surfaces of aluminum components.

Fuel Pump High-Temperature-Performance Tests
P. G. Smith

The fuel pump high-temperature-performance test
loop in which NaF-ZrF,-UF, (50-46-4 mole %,
fuel 30) is circulated was placed in operation again
on June 26, 1957, and has since been operating
continuously. For this series of tests a nuclear
radiation thermal barrier was assembled with the
pump rotary element in order to simulate, as nearly
as practicable, the temperature distribution at re-
actor zero power operation during priming and
purge tests; a bubble-type liquid-level indicator
was installed in the pump fuel-expansion tank; and
four of the seven Moore pressure-measuring devices
were replaced with Taylor instruments.

The head-vs-flow performance tests at 2400,
2700, ond 3000 rpm were repeated, and the dis-
parity in total head between the tests with fuel
and with water was again found to be 1 ft, as
previously reported.3 The total head obtained

 

 

 

3p. G. Smith and H. C. Young, ANP Quar. Prog. Rep.
June 30, 1957, ORNL-2340, p 34.

34

with fuel ot flow rates of 450 to 950 gpm ranged
from 2 ft below that obtained with water to 3 ft
above. The head-vs-flow performance curves ob-
tained with fuel are presented in Fig. 1.2.2.

UNCLASSIFIED .
ORNL~-LR-DWG 25799

 

60

3000 rpm .

 

50

| N

40 S

2400 rpm \' \-
NN

20 \

 

 

HEAD (ft)
o
o

 

 

 

 

 

 

 

 

{0

 

o 500 600 700 800 900 1000
FLOW.(gpm)

Fig. 1.22. Head-vs-Flow Performance Curves for an
ART Fuel Pump Circulating Fuel 30 in a Test Loop.

s ol ror)

Cavitation tests at different operating conditions
(600, 645, and 700 gpm ot 2400, 2700, and 3000
rpm) made during the previous quarter were re-
peated. From the date obtained, curves showing
the minimum surge tank gas pressure and the
minimum suction pressure vs flow for cavitation-
free operation were plotted (Fig. 1.2.3).

Priming tests were made with the fuel at a tem-
perature of 1200°F after having filled the loop to
levels of 0, 9/1 , 2, and 3 in. gbove the expansion
tank floor with the pumps stopped. In all cases in
which the fuel was above the floor of the expansion
tank, the pump primed and gave full head and flow
performance. Without fuel above the floor of the
tank the pump head was about 4 ft low. During
these tests the effects of ZrF, vapor on pump
performance were investigated at various liquid
levels in the expansion tank and for various shaft-
annulus helium-purge flow rates. With the pump
stationary, a purge of 500/liters of helium per
day down the pump shaft was maintained for 46 hr
without fuel above the expansion tank floor, @
purge of 50 liters/day was maintained for 71 hr

L8

 
 

 

9

MINIMUM SUCTION PRESSURE (psig)

" MINIMUM SURGE PRESSURE (psig)

 

with the fuel / in. above the floor of the expan-
sion tank, und a purge of 50 liters/day was mgin-
tained for 24 hr with. fuel at the 2- and 3-in. levels
in the tank. During each of these tests the pump
was rotated by hand one turn each hour. No detri-
mental effects on pump mechanical or hydraulic
performance were noted that could be attributed to
ZrF , vapor.

UNCLASSIFIED
. ORNL—LR—-DWG 25800
32 -1

2,

 

 

30

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6' . . . . - - : .
60 - e - - - 700
. . FLOW({gpm} . =
| Fig. 1 2.3. 'Cur"fee "Shovririnrg ' Minim(mi “Surge and

Suction Pressures vs Flow for Cavitation-Free Oper-
ation of an ART Fuel Pump with Fuel 30. {Seeret-with
£Laption”

PERIOD ENDING SEPTEMBER 30, 1957

The bubble type of liquid-level indicator was
used intermittently during these tests, and the
level indications agreed with those of two spark
plug probes to within 1/16 in.

Fuel Pump Endurance Tests
P. G. Smith

The fuel pump being tested for endurance with
fuel 30 as the circulated fiuid was stopped on
June 4, 1957, after 3550 hr of operation. The
wetted portions of the pump had been thermally
cycled 652 times between 1100 and 1400°F, The
pump lower-seal oil-leakage rate continued to be
5 em3/day, and there was no measurable leakage
from the upper seal. Examination of the impeller
revealed that it had been operating in a region of
cavitation (Fig: 1.2.4). Data on input power at
2700 rpm vs pump surge tank helium pressure,

~which were obtained at the midpoint and at the
end of the test, are plotted in Fig. 1.2.5. The

slight difference in the maximum power input in
these two runs may be attributed to the nonrepro-
ducibility of the instruments used for obtaining
the power data. The 4- to 6-psig range of surge
tank pressures during this test was below that
required for cavitation-free operation.

Disassembly of the hydraulic drive motor revealed

failure of the output shaft bearings prior to termi-
- 'nation of the test. The pump and the test stand
will be put back into operation for further endurance

testing in the near future.

" Primary and Auxiliary NaK Pump Development
H. C. Youhg 5
‘The primary NaK pump test that was terminated

J. N. Simpson

) _durmg the previous. ‘quarter® when the NaK level
- 'in the pump tank rose and flooded the oil cutch
o fbasm has been exammed in order to determlne the
oo -cause” of the- -operating difficulties encountered.
-_-_";':'_The exammoflon of the lower seal has’ mdtcated
- that grease formed by the interaction of NaK vapor
-f;-und_ oil clogged the seal loading springs and
_:,;,eventuall'y" lnhiBifed their operation.
seal faces separated possibly because of normal
. pump . wbrcmons or because of an increase in the

~ When the

 

4P G Smnh ANP Quar Prog Rep. June 30, 1957,

'ORNL.-2340, p 35.

50n assignment from Pratt & Whitney Aircraft.

4. c. Young and J. N. Simpson, ANP Quar. Prog.
Rep. June 30, 1957 ORNL-2340, p 37.

35

 
 

9€

 

   

Flg. 1.2.4. Impelier of Fuel Pump That Was Enduronce Te

 

e T s . e UNCLASSIFIED
; b : PHOTO 28996

sted with Fuel 30 at 1100 to 1400°F for 3550 hr. {SecTErwithrception}

 

13043y SSIFYJ03d LI23r0dd dNV

 

 
 

 

9

 

* "GNCLASSIFIED
ORNL-LR~DWG 2580

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

es e o
) e 8
28
./ DATA TAKEN 3-19-57
) A A A Amd i -
Pl N ' : ‘
/ "~ DATA TAKEN 6-3-57
o ®
= /
& 27 A
=
Q
a
.—
=
% A
Q.
=
>
a
b—'—l
\\SURGE PRESSURE RANGE (4 TO
26 6 psig) DURING OPERATION FROM —
l 1-8-57 TO 6-4-57 (3550 hr)
f | |
~ PUMP SPEED: 2700 rpm
0 4 8 12 48 20

PUMP SURGE PRESSURE (psigy

Fig. 1.2.5. Input Power vs Surge Pressure for Fuel
Pump Operating with Fuel 30. fs'ecm-whl‘r*cupfl‘un)

pump tank gas pressure to obove the lube o:l_‘fi_
pressure as a result of voponzohon of the oif that

feaked into the NaK, the clogged sprlngs were

unable to reseat ‘the seal faces and thus more. oil

“leaked into the NaK.. Further: development work'
 on _the seal region of the pump is planned in an
-~ effort to eliminate these seal problems. o

After removal of the pump from the’ ‘test |oop b
(No. l) a cover plote was installed on the - pump'
tank opening, the atmosphere was changed fo argon, .
and the loop was left for three weeks before it was
An explosion occurred inside the loop -
approximately 45 sec dfter removal of the throttle
NaK sprayed through the throttle - valve

opened.

valve.
opening and caused a lost-time injury. As a result
detailed loop cleaning procedures were prepared

PERIOD ENDING SEPTEMBER 30, 1957

for this and subsequent loop cleaning. The results

of tests of the reaction of NaK and lubricating oil
that were made in order to investigate the cause
of the explosion are reported in Chap. 2.5, ‘‘Ana-
lytical Chemistry."’

Since the oil leakage might have carburized the
Inconel components of the test loop, Inconel
samples were token from several areas and sub-
mitted for metallurgical examination. The maximum
depth of carburization was found to be 0.005 in.,
which was not considered sufficient to be harmful
in further test operations. The loop has been
cleaned and reassembled and is to be put back
into operation.

Operation of NaK pump hot test loop No. 2 was
stopped for removal of four of the first set of 3'/-
in.-IPS electromagnetic flowmeters inserted for
calibration and installation of four uncalibrated
flowmeters. When an attempt was made to resume
operation of the loop, it was found that the main

loop flow was restricted to very low values, that

is, 100 to 150 gpm at 3550 rpm with the throttle
valve wide open. Heating of the loop to nearly
1200°F finally restored full flow. The only subse-
quent interruption of loop operation has been an
unscheduled shutdown for replacement of the cold
trap because water was leaking from a ruptured
cooling coil into the loop drip pan. It is believed
that the thermal shock that occurred when a full
stream of cold water (60°F) was admitted to the
cooling coil while the cold trap was at a fairly
high temperature (450°F) hastened cooling coil
failure. A nozzle was therefore devised to admit
a controlled water-air mixture to the cooling coil.
The nozzle has proved useful in effecting a gradual

_transfer of the cold-trap cooling load from air to
~ water on all NaK pump hot test loops, and its use
has ‘been suggested for the ETU cold traps. The

total operating time for the pump in loop No. 2 is

1850 hr, which brings the total running time for
5pr|mory NoK pumps operating with NaK at elevated
temperatures to 6250 hr.

" "Auxiliary NaK pump hot test loop No. 1. was

converted to ‘operation with hot NaK (56% Na)
- during the previous quarter, The initial operation
. of ‘this foop has been devoted to a study of the
"oxide level in.a new NaK loop operating ot an
elevated temperature. The test loop is constructed

mainly of 4-in. sched-40 Inconel pipe, and it holds
approximately 42 gal of NaK, For the oxide-level
study the loop was heated to 1500°F without cold

37

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

trapping of the NaK, the cold trap being insulated
and uncooled, and plug-indicator break tempera-
tures were recorded as a measure of the oxide
level in the NaK. Data on plug-indicator break
temperatures were also tcken to observe changes
in the oxide level caused by draining the 1500°F
NaK into the system dump tank, cooling the NaK
and the loop to room temperature, slowly refilling
the loop, and reheating the NaK and the loop to
1500°F. This process was repeated twice. The
data obtained in these tests are presented in
Fig. 1.2.6. The plug-indicator electromagnetic
flowmeter only occasionally showed a sharp de-
crease (break) in flow for a particular plugging
screen temperature during an oxide determination

run, and thus the break temperatures are shown as

ranges in Fig. 1.2.6 rather than as points. De-
creases in the break temperatures with time may
be seen at loop NaK temperatures of 1200 and
1500°F and for each dump-tank cooling process.
The initial operation at a NaK temperature of
1500°F showed an average break temperature of
1250°F (extrapolated to 1350 ppm O,) and the final
test run at 1500°F showed a break temperature of
900°F (575 ppm O,). The first break temperature
for operation at a NaK temperature of 1200°F was
approximately 1175°F (1275 ppm O,) and the last
break temperature for operation at 1200°F was
700°F (225 ppm O,). The process of draining the
hot NaK into the dump tank and cooling it there to
room temperature reduced the oxide concentration
in the system NoK. The oxide level decreased
with time during the initial operation at 1500°F
and increased with time during operation at 1500°F
ofter the 346th hr of operation. Since the com-
pletion of the oxide-level study, 500 hr of high-
temperature. operation has been accumulated with
this pump test loop. Head, flow, speed, power,
and cavitation data for the pump will be taken
during the next quarter.

Auxiliary NaK pump hot test loop No. 2 was
placed in hot operation during the quarter, and
calibration of the first set of 2-in.-IPS electro-
magnetic flowmeters for the ETU and the ART was
completed. Four of the calibrated flowmeters were
removed and replaced by four uncalibrated flow-
meters. During the shutdown a defective liquid-
level probe was removed for replacement and carbon
deposits were found on it. The pump was therefore
removed for inspection, and a lower seal was found
to be contaminated in o manner similar to that

38

found on the primary NaK pump; that is, the seal-
face loading springs were clogged with grease.
The pump was cleaned, reinstalied, and calibration
of the second set of 2-in.-IPS electromagnetic
flowmeters will begin early next quarter. At shut-
down for replacement of the flowmeters, 1070 hr
of hot operation had been logged on this test loop
that brought the total hot operating time for auxiliary
NaK pumps to 1370 hr.

REACTOR COMPONENT DEVELOPMENT TESTS
D. B. Trauger

Heat Exchanger and Radiator Develepment Tests

J. C. Amos
R. L. Senn D. R. Ward

A summary of heat exchanger and radiator test
operations during the quarter is presented in Table
1.2.1. The small heat exchanger stands were shut
down for installation of fest pieces.

The semicircular heat exchanger shown in Fig.
1.2.7, which was designed to simulate the tube
stresses of the ART main heat exchanger, is
currently being instalied in small heat exchanger
test stand B, ond test operation is to start early
next quarter. A description of this heat exchanger
and the test objectives were presented previously.”

A 25-tube heat exchanger, type SHE-7, 8 fabricated
by the Process Engineering Corp. and designated
No. 4, has been installed for testing with the fuel
mixture NaF-ZrF 4-UF 4 (56-39-5 mole %, fuel 70).
This test will be a repetition of the previous test
which was interrupted prematurely after 1438 hr of
operation by failure of York Corp. radiator No. 16
(for details see Chap. 3.6, this report). '

A Black, Sivalls & Bryson intermediate heat
exchanger No. 3 (type IHE-8), which was installed
in test stand B, had survived 168 thermal cycles
and 421 hr of high-temperature operation when the
test was terminated by failure of the furnace inlet
line. Daily spectroscopic analyses of fuel samples
for potassium from the NaK system, which was at
a higher pressure than that of the fuel system,
revealed no evidence of leakage. A first-approxi-
mation thermal-stress calculation had predicted a

 

7). C. Amos et al., ANP Quar. Prog. Rep. March 31,
1957, ORNL-2274, p 43.

8 . H.Devlin and J. G. Turner, ANP Quar. Prog. Rep
Sept. 10, 1956, ORNL-2157, p 50, Fig. 1.4.5.

4

o

 
 

)
O

 

UNCLASSIFIED

ORNL-LR-DWG 25802

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1800 # 4 e o * 4 l
PUMP STOPPED AND NaK
e . DUMPED AT 16:50, 7/30/57
. PUMP STOPPED AND NaK DUMPED PUMP STOPPED AT 22:05, 7/26/57, [
1600 . TO SUMP AT 16:30, T/17/57 BECAUSE OF FROZEN COLD TRAP VALVE ~ COLD TRAP DPENED AT
T o - \ - \ '/ / 16:45, 8/8/57
1400 . - ;
o PUMP STOPPED AT 45:25,
fi © 7/22/57 FOR CALROD _
f : , REPLACEMENT\ _ B
j ey _
1200 , * : o . = J
& REMOVED INSULATION FROM
COLD TRAP AND BEGAN COLD
- T T TRAPPING NaX AT 09:45, 8/9/57
- ‘ ‘
< 1000 : T I
: T !
&
=
-
« .
o L
W .
ui _ fl
& T
W 800 1
{ }BREAK TEMPERATURE RANGE
ON PLUG INDICATOR; POINT :
INDICATES MOST SIGNIFICANT
BREAK.
600
IA
400 -
LOOP FILLED SLOWLY (Bhr] AT NaK
. TEMPERATURE OF 200°F‘ ON 7/20/57 STARTED PUMP AT 09:2'5, _
. 7/30/57 (COLD TRAP OPEN) LOCP FILLED SLOWLY (9hr, 40 min)
: ‘ | N AT NaK TEMPERATURE OF 100°F ON
200 |————— 1 SUMP FILLED AT 09:00, 7/12/57 8/6/57 WITH USE OF VENT BUBBLER —
LOOP FILLED AT 15:00, 7/12/57 STARTED PUMP AT 08:30, :\J AFTER SUMP HAD COOLED 160 hr
S | - ‘ 7/25/57 (COLD TRAP CLOSED) |
: | ~w—5TA AT 08: 7
* —— PUMP STARTED AT {3:40, “W——STARTED PUMP AT 09:00, ?COESDT::;"ELJS%%‘;O' 8/8/5
7/16/57 (COLD TRAP OPEN) 7/22/57 (COLD TRAP CLOSED) l
o "l. . L A A l_ 14 A b -_4'A. /l':l.
120 130 240 250 320 “340 350 440 610 620 630

0 100 : Ho

Fig. 1.2,6. Results of a Study of the Oxide Level in the NaK Used in the Startup of a New Inconel Loop for High-Temperature Pump Tests with NaK.

Auxiliary pump test loop No. 1.

TIME FROM FILLING OF SUMP {hr)

640

LS61 ‘08 ¥IIWILHIS ONIAND aoly3d

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 1.2.1. Summary of Heat Exchanger and Radiator Operations (As of September 4, 1957)

 

Howrs of Operation

Howrs of

Number of

 

Test Unit Test at ART Design Nonisothermol Total Hours Thermol Status of Test
Stand Temperature or Above Operation of Operation Cycles

Black, Sivalls & Bryson IHE-B 671 800 2122 173 Terminated because of
heat exchanger No. 2 NaK furnace faojlure
(type HE-8)

Black, Sivalls & Bryson IHE-B 168 253 1131 168 Terminated because of
heat exchanger No. 3 NaK furnace failure
{type IHE-8)

York Corp. 500-kw IHE-B -800 2122 173 Terminated because of
roc'!iciors Nos. 11 NaK furnace failure
and 12

York Corp. ART test IHE-C 73 480 870 9 Terminated because of
radiator No. 1 radiator foiluwre

York Corp. ART test tHE-C 100 140 285 39 Test continuing
radiater No. 2 :

 

life of 35 to 45 cycles for this test, which was
phase |l of a program previcusly described.’ Each
thermal cycle consisted of the following con-
ditions, as achieved experimentally: 42 min of
steady-state operation at 1200°%, isothermal;
transition to power over a 50-min period; 40 min
of steady-state operation at power; and traensition
to 1200°F isothermal operation over en 18-min
pericd. Conditions during power operation were
the same as for phase |, and the traonsition to
power was at a constant rate of temperature change.
The transition from power operation to isothermal
operation was accomplished as fast as the thermal
inertia of the test stand would permit, and the
resultant rate of tempercture chonge ot the NaK
inlet header of heat exchanger No. 3 ranged from
71.5°F/min starting at 1700°F to a final rate of
12.5°F/min between 1300 and 1200°F. Black,
Sivalls & Bryson heat exchanger No. 2, which also
was operated during phase | of this test, had been
thermally cycled 173 times., Phase | included
503 hr of steady-state operation with & meximum
NaK temperature of 1700°F. At the time of failure,
the NaK furnace had operated for a total of 5329 hr,
including 671 hr at @ maximum tube temperature of

1750°F. Since the furnace was designed for 3000 hr

 

 

%). C. Amos et al., ANP Quar. Prog. Rep. March 31,
1957, ORNL-2274, p 46; ANP Quar. Prog. Rep. June
30, 1957, ORNL-2340, p 39.

40

at ¢ maximum NaK temperature of 1590°F, it must
be replaced before further testing can be done in
this stand. Because of the ensuing time delay,
the heat exchangers are being removed for ex-
amination and will not be tested to failure.

The test of York Corp. ART test radiator No. 1
was terminated after 870 hr because of a leak in
the tube matrix opproximately ¥ in. from the NaK
inlet header in a tube near the center of the fourth
row from the air outlet side of the radiator, as
shown in Fig. 1.2.8. The leak was noted immedi-
ately following the first power cycle, and rapid
shutdown was effected to limit damage to the unit.
Even so, the tube that lecked was damaged suf-
ficiently by fire to destroy evidence that would
indicate the nature of the failure. Metallurgical
examination of 16 tubes, four of which were in the
in the immediate vicinity of the tube that failed,
did not reveal a cause for the leak. There were no
incipient cracks and no defects in damaged areas,
and no impurities were evident {for details of the
examination see Chap. 3.6, this report). Inspection
of assembled ART prototype radiators Nos. 2 and 3
will be even more rigid that the quite thorough
inspection of unit No. 1. Heat transfer performance,
gir pressure drop, and the increase in the NaK
circuit pressure drop for radiator No. 1, as well
as for radiater No. 2, which is now being tested,
are in agreement with design values.

 
 

ly

 

 

UNCL ASSIFIED
PHOTO 29722

   
       

FLUID=-NAK HEAT EXCHANGER
20 TUBES
MATERIAL- INCONEL
MANUFACTURER - ORM

Fig. 1.2.7. Twenty-Tube Semicircular Heat Exchonger Type SHE.9.

LS61 ‘0E YIIWILJ3S ONIGNI dOI¥3d

 

 
 

ANP PROJECT PROGRESS REPORT

 

 

Fig. 1.2.8. ART Test Radiator No. 1. {(Serrfrdmniialaith-cupTion

UNCLASSIFIED
PHOTO 29280

 

 

*

 
 

 

"

 

During radiator removal and replacement, extreme
care was exercised to maintain an inert atmasphere
on the test loop, The ART-type circulating cold
trap was drained but not cleaned: In order to
prevent oxide remaining in the cold trap from re-
turning to the system, the cold trap was water-
cooled at the design NaK flow during the heatup
pericd, This procedure is contrary to the previous
startup procedure of allowmg the cold-trap temper-
ature to increase with the system temperature,

cooling the cold trap with air until water can be

introduced safely, and then bringing the cold-trap
temperature down.  There was no indication of the
oxide plugging formerly observed when the cold-

trap system was kept cold during preheatmg_of the

system, - When the system reached 1200°F, a plug-
indicator reading indicated a systemoxide satura-
tion level of approxurnately 600°F; 48 hr later this
level had been reduced to below the sensitivity

~of the plug indicator (300°F). - This startup pro-

cedure demonstrated that if ‘adequate care is

exercised to prevent OXIdGl’IOI'I of residual NaK -

when o system is cut open, the - cold-trap circuit

need not be replaced prior to restartlng system"

operation,
The excellent pertormance of the cold trap in

~ the HE-C test stand has led to elimination of plug
indicators from the ETU ond. ART. -During both

the initial startup and the restarting of the system,
the temperature llmltatlons imposed on.the ART by

the permissible berylhum temperatures were adhered

to closely. This included establishing o temper-
ature gradient in the radiater before the system
temperature necessary for oxide cleanup of the
new piping had been reached.

planning the steps in the operatlon. _ Elimination
of these. devsces ond their assocuated _equipment

“and instrumentation should effect o substantlal-
-saving for the reactor faclllttes. e

Valve Development Tests L
J A Canlm
l T Dudley

rvalve deslgnated ORNL-1, which .was described

prewously, were terminated after 1500 hr ot

1300°F and 910 hr at 1500°F. During the test at -

Pl ug-lndlcator ,
readings were taken to confirm that low oxlde_'))
levels had- been achleved but were not utnl:zed in-

A G Sm1th Jr.“.o_ ucceptably large. Although small scars have been

Fuel Dump Valve.'.---' Tests of the fuel dump

PERIOD ENDING SEPTEMBER 30, 1957

1500°F a steady increase was noted in the force
required to open the valve. The force required
increased from 865 |b immediately after the 500-hr
closure ‘period to 2100 ib following 400 additional
hours of cycling at 24-hr intervals. There was no
measurable leakage during the test at 1500°F.
Examination showed that galling had occurred

between the Stellite-coated valve stem and the

Inconel bonnet at the outer stem guide where the
surfaces were near 1500°F and were exposed to
the atmosphere. The valve was in excellent con-
dition in all other respects, including the Kentanium
151A seat and plug, and is considered to be entirely

~ satisfactory for service at 1300°F,

A second test valve (designated ORNL-2) which
also incorporates the guided-plug feature, has been
assembled, This valve, except for the valve plug
cermet configuration and the omission of the sodium
cooling jacket, is identical to the proposed ART
fuel dump valve. The plug and seat, respectively,

-are the cermets Kentanium KM and K-162B, the

materials now ordered for the ART valves; however,

‘the plug cermet does not incorporate the mechanical
‘retaining feature of the ART valve, since the new

pieces are not yet available. The valve includes
o double bellows to minimize the chance of fission-
gas release in the event of a bellows rupture in

‘the ART and o K-162B bonnet guide against
‘Stellite-coated stem to eliminate the galling which
‘occurred in the ORNL-1 valve. A test is scheduled

to start early next quarter. The pressure drop

characteristics of this type of valve when fully

open are given in Fig. 1.2.9; the data presented
were obtained from water tests. The ART valve

* pressure drop - characteristics should be the same

as those given in Fig. 1.2,9.
‘NaK- Dump Valve. = The first prototype NaK

* valve!! received from Black, Sivalls & Bryson

was tight when tested with water, but it leaked
~ badly when tested in NaK, The valve was dis-
assembled and stress relieved, but there was no

improvement - until it was rapped with a hammer

during a stress-relieving cycle. 1t was then tight
. with water aond  initially with NaK at-the fest

temperature . of - 1000°F. The leakage increased
rapidly with time, however, and was . soon un-

 

109, assignment from Pratt & Whitney Aircraft.
"y A. Conlin, 1. T. Dudley, and M. H. Cooper, ANP

Quar. Prog. Rep. June 30, 1957, ORNL-2340, p 41.

 
 

 

ANP PROJECT PROGRESS REPORT

SONTHEENTHre
ORNL—LR—DWG 25803
100 -

 

50

20

NoK VALVE

10

HEAD LOSS (ft)

FUEL VALVE

{0 20 50 100 200
FLOW (gpm)

Fig. 1.29. Pressure Drop Characteristics Observed
in Water Flow Tests of ART Prototype NaK and Fuel
Dump Valves. {(Serretwith-eeption)

observed on the seat and poppet, the erratic results
appear to be due to weld or thermal stress dis-
tortions. The fuel valve, which had not exhibited
this type of behavior, has a more massive end
connection and o separate seat insert which is
brazed to the body in a configuration which should
minimize strains, The NaK valve is being rede-
signed to provide similar relief as well as to
accommodate cermet seats if this proves to be
necessary. Water flow resistance was also measured
for this valve, as shown in Fig. 1.2.9.

“ln.Line’’ Vealve Actuater Test, — The NaK
valve *‘in-line’’ actuator!! operated satisfactorily
throughout a 3000-cycle test ofter revision to pro-
vide for locking of the spring load adjustment nuts
and correction of other minor defects.

NaK Cold Trap Throttle Valves. — The NaK

cold-trap throttle valve!! configuration has been

44

established, ond the required valves were delivered
for the ETU. Modification of the remaining valves
required for the ETU and ART is in progress.

~ Sodium Circuit Water Flow-Te:si's
J. A, Conlin S. Kress'2

Two tests run during this quarter brought to com-
pletion the water flow distribution and head loss
examinations of various regions of the ART sodium
circuit. A test was conducted in order to determine
the proper size for ‘an erifice in the control red
sodium cooling passage circuit to limit flow through
that circuit to 0.032 cfs with the island flow at its
design value of 0.697 cfs. The test piece simulated
the control rod passage below the thimble and
included the Inconel shoulder which screws into
a threaded recess at the bottom of the island. Four
s/1 -in.-thick erifices were tested which had inside
diameters of 0,250, 0.375, 0.500, and 0.625 in.
ond which were machined to fit snugly into a
1.422-in.-dia section at a transition to a 0.9«in.-ID
passage. The head loss vs flow relationship for
each orifice was correlated in terms of a loss
coefficient, K, defined from the relationship

AH = KV2/2g

where AH is the head loss (in ft) across the orifice,
V the velocity (in fps) through the 0.9-in.-ID
passage, and g is the proportionality constant
relating force to mass and acceleration. The loss
coefficients are shown in Fig. 1.2.10 as a function
of orifice diameter. |

The second water test was a study of the sodium
flow divider in the volute elbows located at each
ART sodium pump discharge nozzle. These dividers
split the sodium flow from the pumps, direct a
portion downward into the reflector, and direct the
remainder up to the island entrance region. It is
estimated that to obtain the design reflector-to-
island flow ratio of 2.2 at a sodium pump flow of
1.106 cfs, the head loss on the island leg of the
flow divider should be 8 ft greater than that on
the reflector leg. The purpose of this test was to
determine the proper position for the flow divider
tongue and to determine the magnitude of the head
loss through each leg. The test piece, shown in
Fig. 1.2.11, duplicated the ART component dimen-
sionally, except for the flow divider tongue adjust-
ment. A diagram of the test piece as installed in

 

129, assignment from Pratt & Whitney Aircraft.

"

 
 

 

 

 

UNCLASSIFIED
ORNL-LR-DWG 25804

1000

500

g 3

1]
Q

N
o

LOSS COEFFICIENT, & (DIMENSIONLESS)
S

 

0 0.4 0.2 03 04 05 06 07
‘ ORIFICE DIAMETER (in.) '

Fig. 1.210. Results of Tests for Determining Proper

Size of Orifice in ART Control Rod Coollng Passage.
& rigd Hom)

the test stond is shown in Fig. 1.2.12. Data were
obtained at each flow divider tongue position (A,
B, C, and D of Fig. 1.2.11) to determine the total

head loss through each leg of the test piece, the
flow ratio between the reflector and island, and the o

entrance velocity head. ~The head losses were
correlated - with the ‘entrance velocity by loss

coefficients defined os stated above. The loss - -
coefficients cobtained are plotted in Figs. 1.2,13°
ond 1.2.14 .as functions of the flow raho and fher

tongue pos:flon. CoT

PERIOD ENDING SEPTEMBER 30, 1957

WONRTENTI
ORNL-LR-DWG 25805

 

TO REFLECTCR

 

FLOW DIVIDER TONGUE

 

 

A B c D

Fig. 1.211. ART Sodium Pump Volute Test Section
Showing the Test Positions of the Flow Divider (A, B,

'C, D). Divider shown in pesition A.

required -head difference and also results in the
iowest net head loss. - :

_Outer Core Shell Thermal Stability Tesi
' J. C. Amos - R. L. Senn

The second test of a one-fourth-scule outer core

shell. mode! for determining dlmens:onol stability
“under thermal cycling conditions was terminated
ofter 339 cycles,  The test period was extended
e R J _ - beyond the scheduled 300 cycles since no changes
The head 'alffeféfic'e' between the reflector and

island legs of the flow divider ot the design flow -
and flow ratio is shown as @ function of the flow
divider tongue position in Fig. 1.2,15. Position A ~
of the flow divider tongue gives approxlmutely the

were noted in fluid circuit resistances. Loss of
lubricating oil flow to the cold sodium loop pump

"necessltated shutdown of the system, and the test
‘piece was removed for examination. Visudl inspec-

tion has disclosed no damage ‘to the core shell,

45

 
 

o b b it

 

 

 

ANP PROJECT PROGRESS REPORT

WATER INLET - E _
—— . -

 

 

 

TAPS MANIFOLDED TO OBTAIN HEAD LOSS

 

 

ISLAND CIRCUIT

R

NOZZLE AND GAGE FOR MEASURING
ISLAND CIRCUIT FLOW

 

 

WATER OUTLET

  

FLOW DIVIDER

VALVES TO VARY FLOW RATIO

GNP Tid
ORNL-LR-DWG 25806

TAPS MANIFOLDED TO OBTAIN HEAD LOSS

 

 

 

 

   

REFLECTOR CIRCUIT

NOZZLE AND GAGE FOR MEASURING
REFLECTOR CIRCUIT FLOW

\ 4

 

 

WATER OUTLET

Fig. 1.2.12. Apparctus for Water Flow Tests of Volute of ART Sodium Pump.

and it is currently being measured to determine
whether any dimensional changes occurred during
the test. The core shell model and the test con-
ditions were described previously.'3¢14 The shell
will next be subjected to an extended creep buckling
test at 1500°F by using an externul helium gas
pressure of 52 psi.

Liquid-Metal-Vapor Condensers
J. A. Conlin A. G. Smith, Jr.

The sodium vaepor condenser for the sodium-pum;;

‘purge-gas vent system has satisfactorily completed

a 1500-hr performance test in a system in which
1000 liters of helium per day was being bubbled

46

through sodium held ot 1200°F. The condenser
consists of 7 ft of 3/4--in. pipe inclined at an angle
of 5 deg to the sodium-pump helium vent line. The
condenser temperatures, which were lower than
those used previously in order to cbtain more com-
plete vapor removal, were 730°F at the inlet, 330°F

at the midpoint, and 90°F at the outlet end. In the

previous tests these temperatures were 810, 470,
and 90, respectively. Upon sectioning of the

 

~13g. p. Whitman, A. M. Smith, and R. Curry, ANP -

Quar. Prog. Rep. March 10, 1956, ORNL.-2061, p 58.

145, C. Amos and L. H. Devlin, ANP Quar Prog. Rep.
March 31, 1957, ORNL-2274 p 50.

o

X

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

OIS NTA L
ORNL-LR—DWG 25807
6
—— 1 POSITION A
5 a4 POSITION 8
~——-—9 POSITION C A .
===-=18 POSITION D e
[+ ‘,a"
g ’a’P”
a7
l&:“ 1 ,4’
14 ,’
2 /~ /.o—"'
; 3 "’ — -
i / L
2 J e
. " /
wi A /'
S > ,-' /
@ / /
3 |/
' ¢
q ;" ‘/ ) |l
i / /"—-_-
I
"l /' //
4 * 5 "-p-—v # -
il
0
0 { 2 3 4 5

REFLECTOR-TO- ISLAND FLOW RATIO

Fig. 1.213, Reflector Loss Coefficient vs Reflector-
to-1stand Flow Ratic and Flow Divider Tongue Position.

condenser, it was found that only small droplets
of sodium had reached the outlet end. The inlet
end contained some condensed sodium that had
been frozen and trapped in the condenser.

The NaK vapor condenser for the NaK pump

purge vent, a. vertical 2:ft section of 2-in. pipe
filled with Demister pocklng, was . sansfactoniy'-;r :
tested with NaK for 1500 hr under the same - flow: =

and temperature condmons as those used for -the

test of the sodium pump condenser, Upon. termi-
‘nation of the test, a very small quantity of NaK "
- was ‘observed in the condenser outlet gas line.

The . condenser pocklng was - covered - wnth very
small droplets ond fme strmgers of NaK

The prototype NuK dump tank vent condenser, :
"consisting of a vertical-12-in. section of 2-in, pipe "
at the inlet ‘or_lower end followed by .o 12-in. -
~section -of 6-m. pipe filled with Demister packing
at the outlet or upper end, which was tested under -

simulated NaK dumps ot 1200°F ond proved to be

- HEAD DIFFERENCE BETWEEN REFLECTOR AND
ISLAND LEGS AT TEE EXIT (ff)

PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

 

 

 

 

SONFIGLMNTITY
ORNL-LR—DWG 25808

12
"
10

\' \
s %

A\

\\G ~———¢ POSITION A

 

A POSITION B
=—-=—0 POSITION C
=====08 POSITION D

 
 
  

 

 

 

 

 

LOSS COEFFICIENT FOR ISLAND
o»

 

 

 

 

 

 

 

 

 

 

5
4
3
2
! S —] :
"'"-..,‘____' -""‘l-._
Ty ——
0
0 { 2 3 4 5

REFLECTOR-TO-ISLAND FLOW RATIO
Fig. 1.214. Island Loss Coefficient vs Reflector-to-

Istand Flow Ratio, and Flow Divider Tongue Position.

. BSNMTENT e
ORNL-LR-DWG 25809

 

20

N

 

 

v

N
AN

A B . ¢ D
o ~ FLOW DIVIDER TONGUE Posmou

 

1
o .

 

i "
N
o .

 

 

 

 

 

Flg. l 2.15. Head D|flerenee Befween VTee Exils vs
Flow - leder Tongue Position for a Reflector-to-Istand

Flow Ratio of 2.2 and an Entronce Velocity Head of
11.43 #.

47

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

99.9% efficient in removing NaK vapor,!® was
sectioned and found to have fine droplets of NaK
on the Demister packing similar to those found in

the NaK pump condenser. As a result of these -

tests the development work on the NaK and sodium
pump and NaK dump tank condensers is considered
to be complete. The NaK pump condenser is
presently in use in conjunction with NaK pump
tests to obtain further information on its performance
in actual service.

Zirconium Fluoride Vapor Traps
J. A. Conlin A. G. Smith, Jr,

Plugging of the zirconium fluoride vapor trap
inlet'® was alleviated by increasing the outer-
wall temperature of the inlet pipe to 1600°F, and
the test was continued to completion. The test
included 500 hr of continuous operation at a helium
flow rate of 5000 liters/day from o fuel sump
maintained at 1200°F and 50 cycles of 30 sec
duration at ¢ helium flow rate of 10 scfm from fuel
at 1300°F. There was no observable carryover
of ZrF, to the off-gas line downstream of the trap.
Light powdery ZrF, deposits were found on the
water-cooled coils in the condenser inlet section
and inside the tubes of the rear heat exchanger
section, as shown in Figs. 1.2.16 ond 1.2,17. In
the rear section of the trap, all the tubes were
coated with vapor deposits on the inner surface
down to and covering the surface of the Demister
packing. A section of the packing removed from
one of the 1%-in.-dia tubes is shown in Fig. 1.2.18,
and the depth of penetration of the ZrF ¢ deposit
into the packing may be seen. Although none of
the tubes were plugged, the !é-in. tubes offered
considerabie resistance to flow, and the pressure
drop across the large tubes was sufficient during
simulated dumps to move the packing te the rear
of the trap.

The quentity and physical properties of the
deposits were considerably different from those
of the deposits found previously. A total of 758 g
was measured as compared with the 3400 g expected
on the basis of previous tests. Densities ranged
from 0.55 to 0.88 g/cm® compared with previous
deposit densities of 4 g/em3; similarly, the thermal
conductivity decreased to 0.04 Btu/hr.°F-ft2 from

 

15M. H. Cooper and A. G. Smith, Jr., ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340, p 47.

165, A. Conlin and M. H. Cooper, ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340, p 48.

48

the previous 0.15 Btu/hr.F.ft2, The small quantity
of ZrF, collected indicates that the purge gas
was not fully saturated, and modifications are
being made in the sump tanks to ensure more com-
plete vapor saturation. The inlet to the trap is
also being changed to a conical section to reduce
thermal losses and to minimize the need for in-
creased heating of this region, since the tempera-
ture range is limited for the ART sodium system,
which provides heating and cooling, as required.
Additional thermocouples are included in the new
test trap to facilitate more precise measurement
of the thermal conductivity of the deposited ma-
terial. This is particularly important, since the
final design of the ART trap cannot be established
until the volume and thickness of the deposit are

established. When the reactor is operating at full
power the surfaces wiil be subject to considerable
beta heating from fission gas decay which will
limit the deposit thickness as a function of its
thermal conductivity. ' '

Other tests are in progress to simulate the effect
of beta heating by radiant heating of deposit
samples in order to determine the effect the beta
heating may have on the physical properties of the
deposited material.

Island Bellows Test
W. B. McDonald
W. H. Kelley, Jr. A. S, Olson

A test of the ART island bellows was undertaken
because of the difficulty of accurately predicting
by calculational methods the stresses to be ex-
pected in this compeonent under service conditions.
The test consisted of cycling the bellows under
conditions of strain, temperature, pressure, and
environment that simulated those of the reactor,
In the test the bellows was compressed from normal
to 0.090 in. in 5 min, held compressed for 25 min,
returned to normal in 5 min, and left in the normal
position for 25 min, The test temperature was
1250°F, and the pressure on the outside of the
bellows was 3 psig for 50 cycles and 20 psig for
the remainder of the test. The inside of the
bellows was at atmospheric pressure. The fuel
mixture NaF-ZrF -UF‘ (50-46-4 mole %, fuel 30)
contacted the bellows on the outer surface, and
the inner surface was in a helium atmosphere.
Sodium was not used inside the bellows because
it would have added little to the value of the test

”

(e

 
 

PERIOD ENDING SEPTEMBER 30, 1957

"UNCLASSIFIED
7 PHOTO 29050

B

 

 

 

 

|
\
| -
U Fig. 1.2.16. Inlet End of ZtF ;~Vapor Trap Tested Under Simulated Fuel Dump Conditions. {Semerat with.oaptiom

49

 

 
 

 

ANP PROJECT PROGRESS REPORT

o,
o
L&
“2
<P
ao
0

9T
Lo

L

 

il e

 

 

 

Rear Sectlon of ZrF,-Vapor Trap Tested Under Simulated Fuel Dump Conditions.

17'

Fig. 1-2
s

 

 

 

 

- [y N

PHOTO 29092

INCHES

 

d Fig. 1.2.18. Pucking Removed fr;:m ZrF‘-Vapor Trap Tested Under Simulated Fuel Dump Conditions, “{Secret-with-egptiat)-

LS6L ‘08 ¥3GWILdIS ONIANI Aoi¥3 d

 

 

 

 
 

 

Y TR A b e e e

ANP PROJECT PROGRESS REPORT

and would have greatly complicated the equipment,
The beliows was tested to failure.

~ The test assembly is shown in Fig. 1.2,19. The

bellows was housed in a covered pot which con-
tained the salt bath and a helium blanket over the
salt, The bottom of the beliows was welded to
the pot, and the top was welded to a thick plate

UNCLASSIFIED
ORNL~LR-DWG 25810

 

AIR CYLINDER

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1
SALT — | L

 

BELLOWS

 

 

 

 

 

 

DIAL INDICATOR

Fig. L219. Apparatus for Testing ART Islend
Bellows. §Secretwithcaptiom

52

that was integral with an actuating stem coupled
to an air cylinder. Helium was supplied to the
blonket area and to the inside of the beliows.
Calrod heaters and thermocouples were installed
outside the pot, and the pot was insulated. Instru-
mentation was provided to properly cycle the
bellows ond to indicate and to control the pressures
and temperature. _

A preliminary cold spring-rate test of the bellows,
performed to provide information for designing the
hot test rig, led to a calculated force of 4900 Ib for
compressing the bellows 0.090 in. at 1250°F. The
force actually required during the test was only
2500 b, which indicated that the bellows had been
stressed beyond its elastic limit and had perma-
nently deformed. The bellows leaked after 80
cycles, and subsequent visual inspection showed
numerous peripheral cracks in the center convolu-
tion at the small diameter of the bellows, These
cracks were evident on both the inner and outer
surface and obviously had propagated to form the
leak. The two-convolution bellows was thus proved
to be inadequate for the service conditions.

Fuel Fill-and-Drcin Tank Test
W. B. McDonald
W. H. Kelley, Jr. A. S, Olson

The scheduled operation of the ART may impose
stress and temperature conditions on the fuel fill-
and-drain tank that are greater than those for which
design properties are known with any certainty.
This uncertainty and the obvious consequences of
a tank failure justify the performance of a test to
verify the tank integrity under service conditions,

The test tank is being made as simple as possible
while retaining pertinent structural characteristics
of the ART tank. The principal change from the
ART design is the deletion of one of the two tube
sheets ond one-half the NaK tubes, NaK will be
circulated through the tubes and cocling annuli by
a centrifugal pump, and the NaK will be heated by

‘electrical surface heaters on the piping. A heated

fuel supply tank to simulate the reactor will be
located above the test tonk so that the fuel can
be displaced from the test tank and be heated. Thus
the test tank can be tested under all phases of
reactor operation, including filling, enriching, loss
of NaK pressure, emergency fuel dumping, and
endurance.

 
 

 

 

 

All major components for the test are on hand,
except the test tank. The test site has been
prepared, and the structural steel supporting frame-
work is in place. The pump bowl, fuel supply tank,
and NaK dump tank have been installed, and piping
subassemblies are complete. The test tank is
being fabricated, and bids for the tube-to-tube

PERIOD ENDING SEPTEMBER 30, 1957

‘sheet brazing have been requested. The tank is

scheduled for delivery in December 1957, and
efforts are being made to reduce this delivery time
by simplifying structural features and construction
methods. Control panels are being fabricated, and
the remaining instrumentation and electrical design
work is being completed.

53

 
 

 

 

ANP PROJECT PROGRESS REPORT

1.3. INSTRUMENT AND CONTROLS DEVELOPMENT'

E. R. Mann

ART CONTROL ROD DRIVE TEST
E. R. Mann C. S. Walker

The ART control rod drive test was terminated
after 3000 hr of successful operation. During the
first 1129 hr of operation, the rod was withdrawn
ond inserted 1161 times. The withdrawal wos at @
slow speed, that is, 370 sec for complete with-
drawal, and the insertion was at a fast speed, that
is, 31 sec for complete insertion, as described
previously.!

After the initial rod-cycling phase of the test
was completed, o series of tests was carried out
in which the rod was fixed at positions spaced
5 in. apart for a minimum period of 200 hr at each
position. The time periods frequently exceeded
200 hr because the 200-hr interval usually termi-
nated at night or on a weekend rather than during
normal working hours. After each nominal 200-hr
period, the time intervals required for insertion
and for withdrowal were measured in order to de-
termine whether there was any reduction in rod
velocity because of sodium having deposited and
frozen on the gears., No changes in the speed of
insertion or withdrawal of the rod were detected.
When the tests in which the rod was held fixed for
a 200-hr period were completed, the remainder of
the 3000-hr scheduled test was taken up by more
rod cycling in @ sequence identical to that used
during the first phase of the test.

The upper surface temperature of the sodium,
which was surrounded by the water jacket, re-
mained constant throughout the test at 230 + 10°F,
except in one instance when failure of a control
component allowed this temperature to drop to
150°F, which is below the freezing point of
sodium. This incident caused no apparent change
in performance of any part of the system.

For the first 1800 hr, the recorder for three of
the thermocouples attached to the wall of the
sodium container was in error by 400°F, and these
three temperatures were 1650°F instead of 1250°F.
This deviation apparently did no damage to the
system.

 

_ !s. C. Shuford and C. . Walker, ANP Quar. Prog.
‘Rep. June 30, 1957, ORNL.- 2340, p 58.

>4

C. S. Walker

R. G. Affel

The rod test system has been dismantled, and
samples of the test sodium have been tcken for
analysis. The Lindsay ‘‘mix'’ rare-earth-oxide
control-rod slugs are clso being examined, os are
the walls of the sodium contaginer. The water
side of the heat exchanger is being examined for
scale formation. [f the results of these examina-
tions are satisfactory, it may be concluded that
the ART regulating rod and its actuating equipment
have met all the specifications that can reasonably
be tested without the actual nuclear tests.

FUEL-EXPANSION-TANK LIQUID-LEVEL
INDICATOR

R. F. Hyland

Bubbler Plugging Tests

The test apparatus was completed that was
being constructed for determining the couse of the
plugging of the helium bubbler tubes of the fuel-
expansion-tank liquid-level indicator. The op-
paratus and operating conditions were described
previously.?2 The system successfully passed a
mass spectrometer leak check, and preliminary
tests of the oxygen- and moisture-removal systems
are under way.

The moisture content of the 200-psi building
helium supply was found to be surprisingly vari-
able. Heretofore, spot checks made on the header
with an Alnor model 7300 dew-point meter showed
an average H,O content of about 0.08 ppm by
volume. On the new apparatus, however, a Beck-
man mode! 179 electrolytic hygrometer is used,
and in one month of continuous monitoring the
moisture of the building helium was found to vary
between 4.0 and 28.0 ppm by volume. The Beck-
man instrument responded very well to liquid-
nitrogen cold trapping of the helium at ofl H,0
concentrations, while spot checks with the Alnor
unit showed no change during the entire monitering
period.

 

2r. F. Hyland, ANP Quar. Prog. Rep. March 31, 1957,

ORNL-2274, p 23.

n

 
 

 

 

The oxygen content of the 200-psi supply has
likewise been continuously monitored for a one-
month period with a Baker model $$-2 super sensi-
tive Deoxo indicator, The oxygen content has
been found to vary between 12.0 and 14.0 ppm by
volume. The internal calibration of this unit was
checked with an external electrolytic cell, and the
two units agreed to within 2 ppm.

Attempts to remove the small amount of oxygen
with hot metal getters have been unsuccessful
thus far. One type of getter that consists of copper
turnings maintained ot 1200°F appears to be
promising. A 50.0 ppm concentration of O, was
introduced into the system and was reduced to
14.0 ppm by this getter.

Aluminum North Head Expansion Tank Tests

Additional tests of the helium-bubbler type of
liquid-level indicator were run in the fuel expan-
sion tank of the aluminum mockup of the ART
north head in order to further investigate the pre-
viously noted? level changes that were associated
with changes in pump speed. Three bubblers were
installed as indicated in Fig. 1.3.1. Typical dota
obtained during ‘tests with these bubblers are
shown in Figs. 1.3.2 through 1.3.5. When the

 

3R. F. Hyland, ANP Quar. Prog. Rep. June 30, 1957,

ORNL-2340, p 50.

 

 

PERIOD ENDING SEPTEMBER 30, 1957

pump speed was increased, the level at bubbler
No. 1 increased and the levels at the bubblers
Nos. 2 and 3 decreased. Thus the previously
postulated *‘dishing’’ effect on the surface in the
expansion tank is further confirmed. There is an
actual change in level, as indicated by the meas-
uring system. A decrease in pump speed created
an opposite and approximately equal effect. These
tests were all conducted with a static level of
1.5 in. of water in the expansion tank and with the
throttle valve set for design flow ot design pump
speed,

Attempts to confirm these results by using high-
speed photography were inconclusive because of
the difficulty of photographing the level through a
small viewing port on top the expansion tank. The
general direction of the level change, as observed
photographically, agreed with that indicated by
the measuring system, but it was impossible to-
determine the magnitude of the change. Improved
photographic techniques are to be used in further
experiments,

Tests were also run to determine how accurately
the level measuring system would respond to a
change in level with both pumps at design speed
(2700 rpm) and with the system at design flow
(645 gpm). Water was accurately metered into and
out of the system under static conditions and the

SR
ORNL-LR-DWG 25822

 

 

 

o _%7.35in.—~——\
© BUBBLERNO. - ‘
7L

"~ {REACTOR)

Fig. 1.3.1. Diagram of Fuel Expansion Tank Showing Locations of Three Helium-Bubbler-Type Liquid-Level

Indicators.

55

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED UNCLASSIFIED
ORNL—-LR—DWG 25823 ORNL—LR—DWG 25825

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6 : I 6 | |
S00rpm 3000 rpm:g-—
5 5
{000 rpm 2500 rpm—f
4 _ 4 .
E K' 'E 2000 rpm%—
w 3 =500 rpm w 3 —
= S’ | g TEST NO.2 1|500 rpm
- BUBBLER NO.{ j
2 oo —| 2 1000 rpm
|
TEST NO.Y | 500 r
BUBBLER NO.! 2500 em
q pm____ | { : N
3000
rpm
0 . 0
O i 2 3 0 i 2 3

FLUID LEVEL (in. H,0) FLUID LEVEL (in. H,0)

Fig. 1.3.4. Fluid Level vs Time at Bubbler No. 1

Fig. 1.3.2 Fluid Level vs Time at Bubbler No. 1
During Second Test.

During First Test.

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED _
ORNL—-LR-DWG 25824 ORNL—LR-—-DWG 25826
6  amem
| | —500 rpm 6
TEST NO. 1 ' 3000 rpm
5 f==w- BUBBLER NO.2 {3 5 :
——BUBBLER NO. 3_1::L1000 rom {?’
3 _[i—‘2500 rpm
4 ':[ 4 -
E 1500 rpm E 2000 rpm
W 3 {4 | w 3 @
= f—E— 2000 rpm = !
- i l - ‘ 1500 rpm
2 T .,
§—2500 rpm 2 "' {000 rpm
¢ : l
!.' ] . 500 rpm
{ S 1 { 5 !
s 3000 rpm : TEST NO.2
< l { |~---BUBBLERNO.2
0 { 0 { |——BUBBLERNO. 3
0 ! 2 3 0 { 2 3
FLUID LEVEL (in. H,0) FLUID LEVEL (in. H,0)
Fig. 1.3.3. Fluid Level vs Time at Bubblers Nos. 2 Fig. 1.3,5. Fluid Leve! vs Time at Bubblers Nos. 2

and 3 During First Test. and 3 During Second Test.

56

 
 

 

quantity required to vary the level 1 in. was de-
termined. The same quantity was added or re-
moved with the system at design speed and flow,
and typical results are shown in Figs. 1.3.6 and
1.3.7. A peculiarity noted in these tests was' that
while bubbler No. 1 yielded consistent results,
bubbler Nos. 2 and 3 gave erratic and inconsistent
results. _ o -

it was felt that severe ingassing of the system
might seriously affect the accuracy of the bubbler

because of its fluid density dependence.. Con-

sequently, a test was run in which the system was
purposely ingassed, and high-speed photography
was used to check the level indicated by the
measuring system. The ingassing was accom-
plished by operating both pumps at 1000 rpm and
then shutting off pump No. 2. The results ob-
tained are shown in Figs. 1.3.8 and 1.3.9. The
indicated increase in level averaged approximately
1.2 in.,, whereas the increase determined photo-
graphically was approximately 0.8 in. No density
error is apparent here because the decrease in
density caused by the ingassing should have
produced a low level indication. Instead, the in-
dicated level was 0.4 in. higher than the photo-

UNCLASSIFIED .
ORNL—-LR-DWG 25827

6 - >

T~

 

 

 

 

 

 

 

4 Y—1in. OF H,0
| - REMOVED

N

 

 

 

 

START TO REMOVE H
L |

DESIGN SPEED—2700 rpm
DESIGN FLOW—645 gpm

 

TIME (min)
n o
n
LY

 

 

 

 

 

 

i J
| TESTNO. 6 _ L (I
fBUB_BL‘ERrIN();q\ Y
iy

 

~ FLUID LEVEL (in. Hy0)

Fig. 1.3.6. Fluid Level vs Time ot Bubbler No. 1

When Water Was Being Removed Durlng Sixth Test.

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL—LR—DWG 25828

 

6
. | 1in.OF H,0 ADDED—'%
° p

 

4

§

 

 

 

 

 

 

4
= | TESTNo.6
£ BUBBLER NO. 1 /
w3 ' i . i J
Z | DESIGN SPEED— /

5 2700 rpm

~ DESIGN FLOW— -
START TO
645 gpm ? ADD H,0

0 }

0 1 2 3
FLUID LEVEL (in. H,0)

 

 

 

 

 

 

 

 

 

Fig. 1.3.7. Fluid Level vs Time at Bubbler No. 1
When Water Was Being Added During Sixth Test.

UNCLASSIFIED
ORNL~-LR—DWG 25829

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6 t
)
(‘
5 ‘
TEST NO.3
----BUBBLERNO.2 ¥
4 | ——BUBBLERNO.3 <&z
- | BOTH PUMPS | %
£ TO 2700 _lrpm, DEGASSIING 1 .
E_ N A
w 3 SYSTEM INGASSING —F
2 i =T
| PUMP NO. 2 OFF —37]
: 1000 rpm —j
1 ' 3
4
0 4 2 3

FLUID LEVEL (in. H,0)

Fig. 1.3.8. Fluid Level vs Time at Bubblers Nos. 2
and 3 During Third Test.

57

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL—LR—DWG 25830

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

© 1
BOTH PUMPS
i TO 2700 rpm, DEGASSING
4
- SYSTEM INGASSING —
=
W > [ PUMP NO.2 OFF —
= 1
C 2 1000 rpm
TEST NO.3
BUBBLER NO.1 \
1
o 3
0 ' > 3

FLUID LEVEL (in. Hy0)

Fig. 1.3.9. Fluid Level vs Time ot Bubbler No. 1
During Third Test.

graphed level. The reason for this error is un-
known at this time.

ON-OFF LEYEL PROBES
G. H. Burger R. E. Pidgeon, Jr.4

In an attempt to eliminate failures® of the on-off
probes used in NeK pump bowls, the intemal
copper wire was aluminized to prevent oxidation.
However, these level probes continued to fail at
high NaK temperatures. Further examination of
the defective units showed that the probes failed
because of internal oxidation of the copper wires
and that such oxidation occurred when the copper-
to-Inconel welds were made. To eliminate this
type of failure, the probe was redesigned so that
the copper and Inconel junction could be brazed
instead of being welded. No probes of this design
have yet been tested in service.

There are sixteen *‘on-off’' level probes now in
use in the gas pots of engineering test loops.

 

40n loan from Radio Corp. of America.

S5r. E. Pidgeon, Jr., and G. H. Burger, ANP Quar.
Prog. Rep. June 30, 1957, ORNL-2340, p 51.

58

The probes in one auxiliary NaK pump test loop
have operated 1640 hr and those in a primary NaK
pump test loop have operated 2162 hr without
failure. These probes are installed where the
NaK temperature is usually below BOO°F. Six
probes are in use in the NaK pump bowls of two
of the pumps being tested. As mentioned above, a
number of these probes have failed and have been
replaced. It is planned to replace all these probes
with the redesigned units as soon as they become
available following complete testing.

MAGNETIC FLOWMETERS

G. H. Burger C. L. Pearce, Jr.*

Calibration and testing of the 2- and 3!5‘5"-
magnetic flowmeters for use in the ART and ETU
were continued according to the program described

‘previously.®  Six 3’4-in. units in lot 1 were op-

erated for 770 hr and six 2-in. units in lot 1 for
1078 hr in NaK pump test loops PKP-2 and PKA-2,
After termination of the calibration test runs on
lot 1, four flowmeters were removed from each loop
and replaced by four uncalibrated flowmeters
(lot 2). Two 3!(‘,- and 2-in. units from lot 1 were
left in the loops for continued testing and to
provide reference to lot 1. Thus far the 3‘4-in.
vnits of lot 2 have operated for 1408 hr ond the
2-in. units for 431 hr. The two 3!5-in. reference
units have operated 2178 hr ond the two 2-in.
reference units for 1469 hr. The units which were
removed from the loops were cleaned, rechamfered,
and inspected in preparation for installation in the
ETU and the ART. Eight magnetic flowmeters
required for the ETU have been delivered and five
units have been installed.

Analyses of the experimental data from the tests
of lot 1 are being made. In general, the results of
these tests with respect to accuracy and reliability
of the units are in agreement with the results
presented previously.® In tests of the lot 2 units,
which were instclled ofter the loops were opened
and rewelded, the oxide level of the NaK appeared
to be higher, as evidenced by erratic loop be-
havior at temperatures below 800°F, and a drop in
sensitivity of the flowmeters at low flows was
observed which had not been evident to such an
extent in the tests of lot 1. The earlier calibration

 

60. H. Burger and C. L. Pearce, Jr., ANP Quar. Prog,
Rep. June 30, 1957, ORNL-2340, p 51.

 
 

 

 

©

showed the 3'/2-in. units to be linear to better than
1% over the flow range from 800 to 1600 gpm.
From 600 to 800 gpm at fluid temperatures below
800°F, the sensitivity appeared to drop as much
as 2% in some of the runs. For the 2-in. units of
lot 1, no such decreases in sensitivity were
evident. The effect was inveshgated in the 2-in.
lot 2 units by running the loop at a high oxide
level without the cold trap. The high oxide level
in the loop did not affect the readings at the higher
flow rates (450 to 600 gpm), but the oxide level or
some other factor produced an apparent change in
sensitivity of as much as 4% at lower flow rates
(250 to 450 gpm). The sensitivity decreased with
a flow decrease. The use of the cold trap did not
correct this condition. Effort is being directed
toward determining the exact mechanism by which
the change in sensitivity occurs.

All the :’%-in. mognetic flowmeters required for
the ETU have been delivered end several have
been installed. The units were delivered without

the magnets attached to permit easier installation

and handling. The temperature tests of the units
have been completed and the results were satis-
factory., The tests indicated that the units will
perform in a satisfactory manner without continuous
temperature and magnet flux monitoring. A curve
of flowmeter output signal vs flow will be supplied
with each individual flowmeter.

LIQUID-METAL-LEVEL TRANSDUCERS
G. H. Burger R. E. Pidgeon, Jr.

Construction work was continued on the ORNL
resistance-type level elements for use in the NaK

‘pump bowls. Twenty of these units have been
completed. The first four units completed con-

tained a sand-type (coarse-grained MgO) - of in-

sulating material for the Inconel wires within the
Inconel tube. Additional units were constructed'
with’ fme-gromed MgO powder, -however, in an at-
~ tempt to get more dense packing of the material.

Construction: of the level element containing the

| fine-grained MgO was complicated by the tendency
~ of the packing tool-and the Inconel wires to gall -

and ‘cause improper spacing of the Inconel wires.
After experimenting with packing techniques, it
was decided to return to- the use of the sand-type
MgO for insulation. Some investigation is con-
tinving of different insulating materials for ap-
plications similar to this resmtunce -type level
probe.

PERIOD ENDING SEPTEMBER 30, 1957

Two of the ORNL level probes are in operation
in the NaK pump bowls in the PKA-1 and PKA-2
test loops. These probes have operated a total of
835 and 1590 hr (plus 500 hr each in the test rig
prior to installation in the pump bowls). Two level
probes manufactured by the General Electric
Company are in operation in the PKP-1 and PKP-2
test loops, and they have accumulated 4740 and
2162 hr of operation, respectively. Eight addi-

tional units have been tested in the NaK-level

test rig for the total operating times given below:

Nc. of Units Operating Time (hr)
2 3060
2 2340
1 875
1 700
1 480
1 320

Thus far all failures of these resistance-type level

probes while in operation have been caused by
weld failures, and no failures have occurred during
this quarter,

Recorder readings of the NaK level taken at
NaK temperatures of 1000, 1200, and 1400°F are
shown in Fig. 1.3.10. All the values given lie

UNCLASSIFIED
ORNL-LR-DWG 25834

 

 

 

 

 

 

4 NaK
* {O00°F
10 —eo 4200°F
9 & {400°F
] 8
27
o6
o
s
-4
3
2
1 K
O - -
0O 20 40 60 80 {00
RECORDER READING
Fig. 1.3.10, NaK-Level Calibration Test of @

Resistance-Type Level Transducer.

59

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

within 0.1in. (1.4% of full scale) of the line drawn
for the data taken at 1200°F. The variation in
the zero-level reading with NaK temperature is
shown in Fig. 1.3.11. These data were taken
while the level was being cycled from above to
below the level probe. As may be seen the zero-
level reading corresponds to the maximum output
voltage of the level transducer. Data taken
during this type of test indicate a zero-level shift
of less than 3% over the temperature range of from

below 300 to above 1200°F.

UNCLASSIFIED
ORNL—LR—DWG 25832

LY CYCLING

!
wmloowm

ZERO-LEVEL SHIFT {FROM CALI TESTS)

|

{~in-LEVEL SHIFT {FROM CALIBRATION TESTS)

LEVEL CHANGE (% OF FULL SCALE)

|
rnlolw

 

300 400 500 600 700 800 900 {000 110C 1200 {1300 4400
NaK TEMPERATURE (°F)

Fig. ‘.3"‘.

Transducer as a Function of NaK Temperature.

Variation in Zerc-Level Reading of

The shift in the maximum level reading is es-
sentially independent of NaK temperature, as
would be expected, since the output voltage of the
level transducer for the maximum level is only 1 to
2% of the full output voltage. During the 3000-hr
test, the maximum level reading remained essenti-
ally the same after the probe was wetted.

Four level probes were operated for 54 days
consecutively with the NaK temperature being
held at 1200°F and the NaK level being cycled
from below the level probe to near the top of the
probe. The recorded zero-level reading was in-
spected for drift which would indicate a change in
the calibration of the transducer. This reading

60

drifted, but the drift was masked by a drift in the
recorder calibration. A transformer used in the
modification of the recorder was sensitive to tem-
perdture, and, as a result, the voltage span of the
recorder changed by an average of about 10.5%
daily because of ambient temperature changes.
Most of the drift in the zero-level reading was
therefore caused by drift in the recorder calibra-
tion. The recorder modification has been cor-
rected so that the recorder calibration does not
change.’

In order to observe the wetting effect, a new
level transducer was installed and the NaK was
heated to 300°F with the NaK level below the
level transducer. The probe was then immersed
in the NaK, a series of step changes in level was
made, and finally the usual output vs level cali-
bration run was made. This procedure was re-
peated for NaK temperatures of 400, 500, and
600°F. In the 300°F test, the indicated level
continued to rise after the step changes in level
had been made. The indicated level reached to
within 5% of its final value after 10 min and to
within 2% of its final value after 2 hr, The re-
sponse to the level changes at NaK temperatures
of 400°F and up was good. The indicated output
was about 1% of its final value immediately after
the leve! changes.

More tests under more rigidly controlled condi-
tions will be required for a final evaluation of the
wetting phenomenon. It is planned to continue the
tests to determine the exact wetting effects on the
measured level accuracy at various fill tempera-
tures and to find, if possible, a solution to the
wetting problem by using other probe material or
by applying a special coating or plating to the
level probe.

Further testing of the NaK level probes has been
delayed by repairs and modifications of the test
rig. The level probes are installed in the test rig
by clamping them to the NaK pots and sealing
them with O-rings. The seals used leaked vapor,
and, as a result, the test rig had to be rebuilt. .

Construction was begun on ORNL resistance-
type level elements for use in the sodium expan-
sion tank. None of these units have been com-
pleted. Construction was also started on the rlg
for testing these level elements.

The design was completed of the level element
for continuous level measurements in the furnace-
circuit drain tank of the ETU. This probe will be

 
 

 

 

mounted in the bottom of the tank and will have a
30‘{,—in. range. The sensing element will be made
from a single 1J‘,--in_. sched-40 Inconel pipe.

The design of the probe for the auxiliary sodium
expansion tank was completed, and fabrication
work was started. This probe will be mounted in
the tank bottom and will have a linear range of
25/8 in. The unit will be of the single-tube type,
rather than the J configuration. A total of nine

units will be required for tests, operational units,

and spares.

ART THERMOCOUPLES
-~ J. T. De Lorenzo
Sheathed Thgrmocouple‘s |

Inconel-sheathed thermocouples (0.250-in.-0D
sheath, MgO insulation) with hot-junction closure
welds made by the Heliarc welding process are
still being tested. Approximately 200 hr of addi-

tional exposure time has been accumulated since

the previous report. = Eighteen Chromel-Alumel
thermocouples with closure welds thot passed

dye-penetrant and x-ray inspection have now com-

pleted over 6000 hr of operation at 1500°F in
sodium.” The operating temperature has been
reduced weekly to 1300 and to 1100°F for readings

at these temperatures.  No major shifts in drift

{1100°F

MANUFACTURER'S

1300°*F

1500°F

_DEVIATION IN' TEMPERATURE (°F)
o ®

TURER’'S LOWER TOLERANCE

!
o

-6 11000 hr 2000 hr 3000 hr

=16 ) - - LOWER TOLERANCE

 

PERIOD ENDING SEPTEMBER 30, 1957

were observed dfiring the first 2800 hr of operation,
but after that time four of the sheathed assemblies

showed sharp downward trends and finally ex-

ceeded the vendor's lower tolerance limit of minus
3‘4% at approximately 3800 hr. Since that time all
four thermocouples have reversed their downward
trend and are drifting back into tolerance, as
shown in Fig. 1.3.12. A fifth couple that showed
a sharp downward drift that stayed within the
manufacturer’s - tolerance has leveled off and
started o slight vpward drift. The remaining
thirteen couples have, almost equally, shown
either a slight upward drift or a slight downward
“drift. Typical plots showing these slight upward
and downward drifts are presented in Figs. 1.3.13
end 1.3.14. The total drifts for these thirteen
couples have ranged from 2 to 8 degrees over the

~ entire test period, and it seems that the drift is

independent of the test temperature. The data for
all 18 thermocouples are summarized in Table
].3.].

Drift dota for similar thermocouples tested in
air revealed upward drifts of all couples tested.

 

7). T. De Lorenzo, ANP Quar. Prog. Rep. June 30,
1957, ORNL.-2340, p 56. The earlier report cited data
on 19 units; but one unit, which was suspected of
having o leak, was pulled out of the apparatus after
about 5000 hr of operation.

UNCLASSIFIED
ORNL~-LR-DWG 25833

46— — 4 MANUFACTURER'S LOWER TOLERANCE

4000 5000 hr

Wog V243 V2ng M Yoy 2 2% M3 s M3 Yes U3 e Y2 %27 e Ter B Bne

C O TIME

Fig. 1.3.12, Drift Curves of 0.250-In.-OD Inconel-Sheathed Chromel-Alumel Thermocouple Showing Downward

Drift Starting at About 3000 hr.

61

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL~LR-DWG 25834

OO°F MANUFACTURER’S UPPER TOLERANCE

 

12 1300°F - MANUFACTURER'S UPPER TQLEF\;ANCE

2 : MANUFACTURER’S UPPER TOLERANCE

DEVIATION IN TEMPERATURE (°F)

{500°F

. 4
0 1000 hr 2000 hr 3000 hr 4000 hr 5000 hr

Yo 243 Y2 Y2 Yor M Phe Uz Y Y3 Yes Uz % %2 G2 T r By %%

TIME

Fig. 1.3.13. Drift Curves Typical of Those Obtained for 0.250-in.-OD Inconel-Shecthed Chromel-Alumel Thermo-
couples That Showed Slight Upward Drifts. '

UNCLASSIFIED
QRNL-LR-DWG 25835

1100°F MANUFACTURER’S UPPER TOLERANCE

12 1300°F MANUFACTURER’S UPPER TOLERANCE

1500°F MANUFACTURER’S UPPER T R

DEVIATION IN TEMPERATURE (°F)

a _

0 1000 hr 2000 hr 3000 hr 4000 hr 5000 hr

 

" 2/, 1 4 "4y - '
g ‘243 2he Yo Yor P Phe U3 b Y3 Yhe 3 e %% S T Tor By
TIME '

Fig. 1.3.14. Drift Curves Typlcal of Those Obtained for 0.250-in.-0D Inconel-Sheathed Thermocouples That
Showed Slight Downward Drifts. '

62

fe:

 
 

 

W

 

 

 

L

 

It was believed earlier that leaks in sheaths of
the thermocouples tested in fuel and in sodium
could account for the downward drifts, and there-
fore one sheathed couple which had exhibited a
sharp downward drift was removed from the sodium

Table 1.3.1. Results of Drift Tests of Inconel-Sheathed
Chromel-Alume! Thermocouples in Sodium

 

 

 

Test Thermocouple Rea‘din‘g (CF)¢
Period — =
(hr) Deviation Spread®
Test Temperature: 1500°F
0 52 3.5
1000 5.1 4.5
2000 5.4 6.5
3000 4.3 5.5
4000 4.4 : 7.3
5000 24.0(4. o 18.5(8.5)%
6000 1.9(13.5)¢ 30.5(13,5)¢
Test Temperature: 1300°F
0 7.3 X
1000 . 8.1 . 50
2000 57 . 60 -
3000 - 47(5.6° 12.3(5.0)¢
4000 s 19373y
5000 - 3.8(5.5¢  15.7(9.7)¢
6000 . 2.7(4.6¢ . 19.5(9.0)¢
Test Temperature: 1100°F
0 68 26
1000 68 .34
2000 62 "'4*6"
0 6y . 10
4000 4.26.2)° 23,5(7.7)91;
5000 4963 “14.0(9.9¢
14; 5(9 5,4, ‘

- gsooo . ,_-'-_:4.6(6-.2)“’

 

tained two fhermocouples cged 24 hr ot 135
pnor to testing. -

-®Deviation of the average of the ihermocouple feodings :
_from the test temperofure. e

“Maximum spread of the fhlfty-elght readings obtamed.“fl“-

14 25,31, 87, and 91,

eYalues in pufenthesee exclude two assembhes, Nos

14 and 31.

fValues in parentheses exclude three assembltes, Nos

14, 31, and 91.

PERIOD ENDING SEPTEMBER 30, 1957

for -examination, Metallographic examination re-
vealed no evidence of a leak; in fact, the unit
appeared to be remarkably sound even after the
5000 hr of operation in sodium at 1500°F. |t is
now believed that the temperature cycling of the
fuel and the sodium could have produced ‘‘breath-
ing'’ of the magnesium oxide on the unsealed
connector end and thereby induced conditions
which might possibly cause the downward drifts.
The temperatures were not cycled during the drift
tests in air. Additional drift tests with sealed
samples are being proposed in an effort to de-
finitely establish the cause of the downward drifts.

Thirteen similar specimens and three platinum,
platinum=10% rhodium units have completed over
4000 hr of operation at 1500°F in NaF-ZrF ,-UF,
(50-46-4 mole %, fuel 30). Weekly readlngs were
also taken at 1300 and 1100°F. After the first
1400 hr of operation, three Chromel-Alumel units
started showing sharp downward drifts similar to
those shown in the first 4000 hr in Fig. 1.3.12,

‘and one unit finally exceeded the vendor's lower

tolerance limit at 2000 hr. The drifts of the other
two units hove leveled off, and slight upward
trends have started. The remaining ten Chromel-

- Alumel assemblies showed drift performances

similar to those shown in Figs. 1.3.13 and 1.3.14
over the first 4000 hr. For comparison, the typical
performance of the platinum, platinum—10% rhodium
units as they aged for over 4000 hr in fuel 30 is
shown in Fig. 1.3.15. The data on these 16
thermocouples are summarized in Table 1.3.2.

A cross section of the hot-junction end of a

- sheathed Chromel-Alumel thermocouple is shown
. in Fig. 1.3.16. - The closure consists of a four-
"+ hole Inconel plug which is welded to the sheath
" and wires with the Heliorc welding technique. "A
- new, improved closure technique is being investi-
gated, in which Coast Metals brazing alloy No. 52
- will be used to. form a brazed joint that will leave
. a minimum void below the tip and will permit
Nmteen assemblies were tested eoch a%;T:Ihye;:'c:’r;_x ' :

accurate rpositioning'_df_‘_fhe‘ thermocouple junction.
This new type of closure.is itlustrated in Figs.

3j:f.— 1.3.17 .and 1.3.18." The test samples for the addi-
- tional - tests will 'be""_'fa!:ric_p?ed- with the new
ch!ues in purentheses exclude flve nssemblles, N°"."7 brazmg technlque. | 7

A 50-kw Megatherm mducflon ‘heating unit is

being used to rapidly heat samples of these

Inconel-sheathed thermocouples in order to deter-
mine the effect of heating rate on the integrity of

63

 
 

 

 

ANP PROJECT PROGRESS REPORT

" UNCLASSIFIED
ORNL-LR-DWG 25836

 

 

1100°F
0
-8
MANUFACTURER'S LOWER TOLERANCE
-6
&
& ® |
o
E' 1300°F
c o0
o
=
i
- -8
Z —
z MANUFACTURER'S LOWER TOLERANCE
= ) :
g
>
ur
O
8
)
-8 ACTURER'S LOWER TOLERANCE
-16 1000 hr 2000 hr 3000hr 4000 hr
oy g Y3 Y Y3 g Y3 O %> 17 o B 8y
TIME

Fig. 1.3.15. Drift Curves Typical of Those Obtained for 0.250-in.-0D Inconel-Sheathed Platinum, Platinum-10%

Rhodium Thermocouples.

UNCLASSIFIED

 

Fig. 1.3.16. Cross Section of Helicrc-Welded Junction of a 0.250-in.-OD Inconel-Sheathed Chromel-Alumel
Thermocouple . Showing Inconel Filler Tip, Weld to Sheath, Thermocouple Wire, Void, and Magnesium Oxide In-

sulation,

64

i

 
 

 

 

®

the couple. The test sample is mserted in a NaK-
filled Inconel charge container made of /,6-m.-0D
0.030-in.-wall tubing. One sample failed after
240 heating cycles between 800 and 1200°F at a
heating rate of 100°F/sec. In this failure, both
Chromel wires were broken, but the Alumel wires
remained intact. In addition to supplying informa-
tion regarding the physical soundness of these
sheathed couples, the induction heating unit
provides a simple means for obtaining aging data

PERIOD ENDING SEPTEMBER 30, 1957

The study of drift as a function of time and tem-
perature of similar sheathed thermocouples ond
conventional beaded thermocouples operating in
air has continved. In these tests, all the thermo-
couples were opercted continuously at one tem-
perature. Three temperatures were investigated,
1800, 1600, and 1300°F. The 1800°F test was
terminated after 5000 hr; the test at 1300°F at
6000 hr; and the test at 1600°F is still under way,

with data having been accumulated for over

6000 hr.

All thermocouples that were tested at

under various degrees of cyclic thermal shock.

Table 1.3.2, Results of Drift Tests of Sheathed Thermocouples in Fuel

 

Chromel-Alumel Chromel-Alumel Pt, P+—10% Rh

 

 

 

 

PTe?fd Assemblies? Aged 24 hr Assemblies® Aged 200 hr Assemblies®
erio
(ht) Deviation? Spread® Deviation? Spread® Deviation? Spread®
(°F) °F) - P (°F) (°F) (°F)
Test Temperature: 1500°F
0 1.9 45 2.9 0.2 -1.0 5.5
1000 3.5 3.8 - 5.1 0.7 0.4 1.0
2000 2.4 7.0 7.3 0.5 ~0.6 2.0
3000 0.2(1.35) 23.5(1y 8.4 0.8 ~1.4 5.2
4000 0.7(2.2)f 22.0(8.5) 8.0 1.0 0.7 3.0
Test Temperature: 1300°F
0 7.0 | 3.0 6.9 0.1 -0.1 2.7
1000 4.3 8.1 6.9 0.9 ~1.2 2.1
2000 2.5(3.2). 14.2(9.0Y 7.8 , 0.4 -0.6 4.2
3000 1.9(3.3) 20.5(10.5) 8.8 0.7 -0.8 4.2
4000 3.8(4.8) 13.7(6.4) 8.8 1.2 -0.6 5.3
' Tei!_ T_eniperofure: "I'IOOOF
o . 42 24 4.6 0.3 ~2.1 4.0
1000 52 43 b4 0.8 -0.3 1.3
) 2.8(4.8) 1320240 77 1.0 ~0.5" 27
3000 2.4(3.6Y 18. _5(10,0)f 83 06 -0.3 3.2

4000 2.7(3.9')’ 17.3(_7.4)[ ;3.1‘_ 1.1 0o 38

 

aElghf assemblies were tested' each ussembly comained two thermocouples oged 24 hr at 1350°F in helium prior
to testing.

PTwo assemblies were tested each essembly contamed two fhermocouples uged 200 hr at 1350°F in helium prior
to testmg

CThree essemb!:es were tested each assembly contained two thermocouples.
dDeviohon of the average of the ihermocouple readmgs from the 1est temperature.

' €Maximum spfeud of the thermocouple reodmgs. . ' '
fvaives in parentheses exclude assembly No. 143.

65

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
Yo

 

Fig. 1.3.17. Parts for Completing Brazed Closure
at Hot Junction of a 0.250-in.-0OD [Inconel-Sheathed
Thermocouple. In this view the sheath has been cut

" to show the wires. The four-hole Inconel plug, the

Inconel end plug, and the Coast Metals No. 52 brazing
alloy ring are also shown.

1800 and 1300°F are being prepared for metallo-
graphic examination. The dato token during these
tests are summarized in Table 1.3.3.

Well Thermocouples

The design of a well thermocouple for tempera-
ture measurement of high-velocity, high-tempera-
ture (1400 gpm, 1100 to 1500°F) liquid metals in
Inconel piping has been completed, and several
units were constructed. The thermocouples (two
per well) employed in these units are fabricated
from 20 AWG, solid, Chromel-Alumel wire (ac-
curacy, 13/8 of 1% over range 530 to 2300°F) and
are welded with Heliarc welding techniques to the
bottom of the well. All the ceramic beads used
contain a minimum of 96% aluminum oxide. A
typical unit is shown before and after assembly in

Fig. 1.3.19.

66

 

Fig. 1.3.18, Cross Section of Brazed Hot Junction.
The light gray areas are brazing alloy; the dark gray

areas are MgO insulation. Note pores in braze material
above the thermocouple wire and the lack of voids

below the closure.

The well immersion depth was arbitrarily chosen
as one-third the inside diameter of the pipe line.
The well-body material selected for use with 2'/2-in.
and larger pipes was 3/B-in. sched-40 Inconel pipe,
and ‘/4-in. sched-40 pipe was selected for wells in
2-in. pipes. The immersed portion of the well

body was turned down to a wall thickness of

0.050 in. This well-body design was tested for
pressure drop effects on a water loop at velocities
up to 1200 gpm. The permanent pressure drop was
less than 2.0 psi ot 1200 gpm. The lag length of
the well body did not extend beyond the 4 in. of
pipe insulation in an attempt to minimize heat
losses and the selective oxidation effect that is
common in long, narrow thermocouple wells used
at elevated temperatures. The thickness of the
well bottom was determined on the basis of
stresses at operating pressures and temperatures.
A value of 0.109 in. was selected on the basis of a
stress analysis.

o

i

»

 
 

o

 

f

 

 

 

In order to ensure a reliable and easily re-
producible attachment of the thermocouple junction
to the well bottom, a technique was developed
which permitted the coating of the thermocouple
junction with Inconel without undue burning of the
wires. This ““wetting'’ of the junction simplitied

PERIOD ENDING SEPTEMBER 30, 1957

the subsequent Heliarc welding to the well bottom.
The diameter of the spherical junction produced
can be controlled to within £0.005 in. with very
little difficulty. It was discovered that careful
cleaning of oxide from the thermocouple wires
facilitates the formation of a reliable junction.

Table 1.3.3. Results of Drift Tests of Sheathed and Beaded Thermoeouples in Air ot 1300, 1600, and 1800°F

 

Cl;ll;omel-Alumel

Specicl Begded
b

Chromel-Alumel
Sheathed
Assemblies?

Test

Period Assemblies

Chromel-Alumel
Normal Beaded
Assemblies®

Pt, Pt~10% Rh
Sheathed
Assemblies®

Chromel-Alumel
Sheathed

Assemblies?

 

(hr) Deviation! Spread® Deviation! Spread® Deviation! Spread® Deviation/ Spread® Deviation/ Spread®

CH R CH CF

CF)

CH  CAH A CH R

 

Test Temperature: 1300°F

0 5.0 3.3 -7.0 2.1 5.3
1000 5.5 3.7 -4.0 3.1 -4.4
2000 5.3 2.4 -3.8 3.5 -4.1
3000 60 25 30 - 4.5 -3.0
4000 6.2 2.0 -2.0 4.5 -2.0
5000 7.8 2.0 0.6 5.0 0

1.1 Measurement error
2.2 0.2 2.0
6.5 -0.5 2.0
5.0 ' ‘ =3.5 5.0
6.0 -3.0 4.0
5.0

Test Temperature: 1600°F

0 3.8 2.8 6.0 2,0 -7
1000 6.4 2.6 3.9 2.5 -2.2
2000 7.0 2.7 6.2 5.5 =0.7
3000 6.0 3.0 120 . 6.0 5.0
4000 13.0 4.0 16.0 5.0 7.0
5000 12.0 7.0 17.3 4.0 0.6
6000 9.9 4.0 15.5 4.0 0.3

3.5 4.7 4.3

6.1 5.6 3.1
10.7 14.0 4.0
14.0 17.0 4.0
18.0 12.0 5.0
22,0 12.0 6.0
24.0

Test Temperature: 1800°F

c 10 2.6 ~7.5 4.3
1000 7.7 15.1 5.2 19.2
2000 © 80 . 16.2 26.2 9.5
3000 120 200 30.0 4.0
4000 140 1800 . 300 150

15000 - 161 20 330 150 7'

103
8.7

nz
18.0

21.0

"‘34'.‘_0' )

2.1 5.5 4.2 Measurement error
14.3 - 10.2 10.5 0.2 4.0
19.2 140 - 140 -9.0 6.5
29.0 16.0 12.0
46.0 140 12,0

68.0  20.0 15.0

 

“Sn: assemblies were tested cnch ussembly coniuined two thermocouples nged 24 hr ot 1350°F in helium pruor to

teshng.

bsix assemblles were tesfed ecch nssembly contalned fwo thermocouples fcbricoted wlth especmlly cleaned wire
(GCCUI’GCY. +3 é of 1% over range 530 to 2300°F) in ccreful Iy cleaned wells.,
€Same as speciol ussemblies except that they received no speclul cleaning. -

dScme as nssembl:es descrlhed in footnote @ except fhaf they were aged for 200 hr ot 1350°F in helium,

- €Five ossemblies were tested; eoch asumbly contained two thermocoupies.

fDevmhog of the average of the readings of the fhermocouples tested from the standard test temperature of 1300,

1600, or 1800

EMaximum spreod of the readings obtained.

67

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

   

UNCLASSIFIED
859

 

   

    

 

T T T T UL |
EE Jd T

" e e 8 SO

et S

B

 

 

 

 

Fig. 1.3.19. Well Thermocouple Before and After Assembly. The unassembled unit shows the copper chill

block for junction formation, the Inconel plate used in making the weld junction at the well bottom, and the plastic

tubing used to protect the unit during hendling.

The cross sections of junctions shown in Figs.
1.3.20 and 1.3.21 reveal the effect of careful
cleaning of the wires.

Extreme care was taken in the assembly of the
well thermocouples. All ceramic beads were care-
fully washed with distilled water and methyl
alcohol and baked at 1500°F for 1 hr. All wires
were washed with methyl alcohol and air dried.
All well-body pieces were degreased, washed with
methyl -alcohol, and baked at 400°F for 1 hr. All
subsequent handling after the cleaning was done
with clean white gloves. After installation o
clean, coil-spring type of flexible extension will
be provided for each well thermocouple to eliminate

68

any possibility of bending near the region of high-
temperature gradients.

Drift tests on five sample units mstolled
flowing NaK have been started. The test proce-
dure will be similar to that used in tests of
sheafhed couples.

Surface Thermocouples

Thermocouples for surface attachment to Inconel

piping and components will be fubrlcated in a

manner identical to that used for the fabrication of
the well couples. A reliable and simple Heliarc
method of welding these surface-type couples to

"

b

 
 

 

PERIOD ENDING SEPTEMBER 30, 1957

¥ UNCL ASSIFIED
] T-10307

 

 

 

 

 

 

 

 

 

 

 

 

i

-Alumel Thermocouple Wires.

Coated Chromel

sn of Inconel-

Welded Juncii

Cross Section of Heiicn;

20.
Chromium oxide deposits may be seen in the Chromel wire. -

1.3

Fig.

on ©

 

69

 
 

 

ANP PROJECT PROGRESS REPORT

 

BEY UNCLASSIFIED
T-10731

aum,
1
) fivwwwfl.ré

.hu:

PN gt

 

 

 

&
A

“.u..
e,
w35

A

AR

o

%

mw

e

 

-Coated Chromel-Alumel Thermocouple Wires

Cross Section of Hellarc-Welded Junction of Inconel

That Were Carefully Cleaned Before Welding.

Fig. 1.3.21.

Note absence of oxide deposits.

70

 
 

 

 

the various Inconel surfaces has been devised by
the Metallurgy Division.
Apparatus for Checking Thermocouples

A melting-point apparatus filled with NaF-CaF,
(68-32 mole %) has been completed that is being

PERIOD ENDING SEPTEMBER 30, 1957

used for weekly spot checks of the three Pt,
Pt-10% Rh couples that have been used as the
standards for all the thermocouple tests. The
melting point of the NaF-CaF, mixture as meas-
ured with these three couples appears to be
1487.5°F and is reproducible to within £1°F.

71

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT .

1.4. ENGINEERING DESIGN STUDIES
A. P. Fraas

APPLIED MECHAN!CS AND STRESS ANALYSIS
'R, V. Meghreblian

Thermal Stress Cycling Tests of Beryllium

D. L. Platus V. H. Kelley
J. R. Tallackson

The beryllium reflector in the ART will ex-
perience thermal strain cycling whenever the
power level of the reactor is changed. The most
severe strains in the beryllium will occur in the
region beneath the load ring. These strains will
be produced by the gross radial temperature
gradient across the reflector combined with the

30 deg TYPICAL

  

15 deg TYPICAL

 

 

‘9/32—“‘1. R
23%5-in. R

 

 

 

 

 

29/32—in. R
1945 -in. R

 

 

 

 

 

129%,-in. R

 

/
v
1.996 in.

 

 

 

local gradients in the vicinity of the coolant
holes. Superimposed on this strain field will be
the strain concentrations arising from the presence
of the holes themselves. Over any extended
period of operation these strains will be cyclic
in character and can lead to a fatigue crack.

A small-scale experiment has been designed to
test the resistance of beryllium to fatigue cracking
under the tempercture conditions anticipated in
the ART. The test specimen is a short, thick-
walled cylinder with an axial pattern of small
holes (Fig. 1.4.1). The size and locations of
the ‘/u-in.-dic holes simulate the cooling-hole
pattern in the reactor, and the radial temperature

UNCLASSIFIED
ORNL-LR~-DWG 25889

NaK
INLET

 
 
  

FILLER PLUG TO INCREASE
FLOW VELOCITY AND IMPROVE
HEAT TRANSFER

  

NaK
INLET

Fig. 1.4.1. Beryllium Thermal-Cycling Test Specimen.

72

 
 

 

 

 

 

gradient across the cylinder has been calculated
to produce a gross strain variation similar to that
in the reflector.

A forced-circulation loop (Fig. 1.4.2) is used
to produce the thermal cycles. The beryllium
specimen is housed (Figs. 1.4,2 and 1.4.3) so
that both the inner and the outer cylindrical
surfaces are exposed to flowing sodium. Thermal
deformations are produced by the radial temper-
ature gradient across the cylinder. Sodium temper-

atures are controlled so that the test specimen

is cycled between the isothermal and temperature-
gradient conditions at Y-hr intervals, as shown
in Fig. 1.4.4, The first test will be for a period
of 500 hr and will produce 500 complete strain
cycles.

A detailed study of the stress (elastic) distri-
bution in the specimen is under way., Relaxation
methods are being used for this study at present,
but analytical techniques are being considered.

Thermal-Cycling Test of a Welded Core
Shell Model

B. L. Creenstreet

Shells | and Il (the inner and outer fuel annulus
shells) for the ART and ETU are to be made from
rolled plate weldments which will be machined
to the final contour. This method of fabrication
replaces that of spinning, The weld joint lo-
cations are shown in Figs. 1.4.5 and 1.4.6. As
may be seen, each half of shell | has both circum-
ferential and longitudinal welds, whereas fhe
halves of shell Il have two longlfudmul ones.

The matenul in the heat-affected. zones of the_'
welds will- exhtblt re!atlvely large grain structure,
This material will permit preferential deformation,
since it will be weaker under steady load. than
“the rest.of the shell.. This strain mtens:flcahon,
coupled with a lower resistance to strain cycling "
damage, ‘may. result in premature fcnlure. From .
this- sfandpomf the most critical regions are the
locations " at which the c:rcumferenhcl and longl- -

tudmol welds mtersect.r, R

cycling condlhons. ,

the weld pattern. This model is to be tested in

PERIOD ENDING SEPTEMBER 30, 1957

~ the existing core-shell test rig and will be sub-

jected to simulated service condmons identical
to those applied to the two /4-sca|e specimens
tested previously.! This test will establish the
the adequacy of the welded reactor shells.

Stress Anclfsis of Island Bellows
S. E. Moore

~ The geometric configuration of the island ex-
pansion bellows of the ART consists of a fourfold
repetition of the unit shown in Fig. 1.4.7. The
unit is composed of three elements, a concave
flange, a very shallow cone, and a convex flange.
Adequate methods are presently available for
analyzing these shapes,?3 and IBM 650 and 704
digital-computor subroutines have been written
for these and other types of rotational shell
elements. These routines have been applied to
investigate the stress distribution and load-
deflection curve for the complete configuration,

The design condition for the bellows is 300
cycles of 90-mil axial deflection at a temperature
of 1250°F. The stress analysis of the present
design revealed that under these conditions the
bellows would be subjected to considerable
plastic deformation at the outer bend of the
convolutions. Correlation with strain-cycling data
for Inconel indicated that the fatigue life of the
design was less than the proposed 300 cycles,
This was verified in a recent test when the

‘bellows failed after 80 cycles.4

The genera! subroutines mentioned. above have

" been applied to obtain designs of several bellows
 which would withstand the deflection and external
-pressure (20 psi) onticipated within the reactor.

A design has been selected from this group which

- consists of four convolutions of 0,062-in.-thick

(,fmctérial . The outer diameter of the new bellows
will be: % f in, greafer thun that of the bellows of

i;,v_an. 1 4

 

o W “Bell,” ANP Quar. Prog 'Rep. Sept. 10, 195,

A Yescale model of shell Il made - with thei{;’ .°R"L 547, p51-52. "

shell l weld pattern wull be tested under sframj.,:, ORNL CF-56-12.37 (Aug. 22, 1957).
It is ‘expected that the
difference in contour between the two shells will

have little effect on the distortions arising from.

M. Essl inger, Stauc Calc'ulauons of Bo:ler Bottoms,

3See, for example, S. Timoshenko, Theory of Plates
and Shells, McGraw-Hill, New York, f940.

4W‘ H. Kelley, ART Island Bellows Test, ORNL
CF-57-8-123 (Aug. 29, 1957).

73

 
 

 

ANP PROJECT PROGRESS REPORT

o
w
L
vy
vy
<
od
Q
z
3

-

PHOTO 29724

@

 

 

 

«Cycling Experiment Test Rig.

Beryllium Thermal

4.2,

F'g-

74

 
 

 

 

 

 

9

 

PERIOD ENDING SEPTEMBER 30, 1957

= ¢ UNCLASSIFIED
T.29723

Figs 1.4.3. Beryllium Specimen in Housing.

UNCLASSIFIED .
ORNL-LR-DWG 25820

   

 

 

 

 

 

 

 

1CYCLE ———=1_ BERYLLIUM
R _-———4/2hr e iyghy ] " TEST PIECE
=~ 1250 — P
£ 1200 | — N
wl
i
=
2
Wt
F 900

 

 

Fig. 1.4.4. Cycle of Temperature vs Time Used for
Beryllium Thermal-Cycling Tests.

73

Stress Analysls of Shield Support Structure
J L Leggett

The two prmcnpal structural components of the

"ART ' shield are the lead support structure and

the water bag. ‘A detailed stress analysm of the
lead suppcrt has been concluded, and “an ‘analysis

of the water bag has been initiated.

The mu|or structurol member - for SUpportmg the
lead is the equatorial shell which fits around the

girth of the reactor and carries the equatorial lead

(13,937 Ib), The shell is to be fabricated from
a ;’lé-in. carbon steel plate to which the lead

will be bonded. In addition to providing support

75

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

. "BECRET
ORNL~-LR-DWG 25894

 

EQUATOR

l

10in. (£ Y in)

 

 

 

 

" Fig. 1+4.5. Shell | Weld Pattern. |

SEERET
ORNL—LR~DWG 26453

 

EQUATOR

 

Fig. 1.4.6. Shell 1l Weld Pattern.

for the lead, the shell will also transmit the
weight loads from the north head lead shield
(4678 Ib), the south head lead (5475 Ib), and the
down-drain lead (6210 Ib).
examined in the stress analysis of this structure:
(1) the bolts that will transfer the load of the
equatorial lead shield to the overhead reactor
_support structure, (2) the lugs that will transfer
loads to these bolts, (3) the steel backup shell

76

Five creas were

UNCLASSIFIED
ORNL-LR -DWG 25892

0.501 in.—

 

 

 

(D CONCAVE FLANGE
(® CIRCULAR FLAT PLATE
(@ CONVEX FLANGE

Fig. 1.4.7. Elementary Unit of Island Expansion
Bellows. (Secret with caption) '

to which the lead will be bonded, (4) the north
ring of the equatorial shielding shell, and (5) the
stresses in the steel and lead at the bond surface
as a result of dissimilar coefficients of thermal
expansion,

Eight bolts unsymmetrically placed around the
north ring of the shell will transfer a load of
approximately 30,300 Ib to the reactor support.
The stress in the most heavily loaded bolt was
checked on the basis of the net cross-sectional
area at the root of the threads. Stresses in the
welds connecting the lugs to the steel ring and
the ribs of the steel backup shell were also
examined.

The investigation of the steel backup shell was
based on the primary assumption that the combi-
nation of the steel shell and ribs will carry all
the weight of the lead and the steel. The lead
was not considered to be carrying any load. The
critical point on the shell was assumed to be

located 10'/l in. above the equator of the reactors
where accommodation of one of the coolant tubes

would require that the rib be notched (see Fig.

124

 
 

 

 

 

 

1 )

1.4.8). The analysns indicated that o %-in. rib
combined with a /l -in. shell would suffice,

The north ring of the equatorial shielding shell
was investigated from the standpoint of bending
and torsion. Here it was assumed that the weight
of lead and steel would be applied along one edge
of the right-angle cross section as a uniformly
distributed load. Stresses in the steel and lead
as a result of changes in shield temperature were
investigated at the bond surface. Two conditions
were considered: a slow increase in temperature
of 180°F (that would result in ¢ bond temperature
of 250°F) to simulate the approach of the reactor
to power operation and @ sudden decrease of
120°F (a bond temperature of 130°F) to simulate
shutdown of the reactor. An elastic analysis
based on these operating conditions revealed that
the stresses in both the steel and the lead ot
the bond would remain within acceptable limits.

An analysis of the aluminum water bag is
presently under way. Since the north head is
made of 10 or 12 segments and the equatorial
section of 5 segments, it will be necessary to
examine the integrity of the individual pieces as
well as the whole unit, Preliminary analyses

UNCLASSIFIED
ORNL-LR-DWG 25893

 

 

 

 

 

 

 

 

 

. 287 in, 47in,
J y _
} ¥ ¥
i NOTCH
£
@
= _
Q | -
| ~EQUATOR Yo"\l 5d
- - 29in. - —
Ny 329 - o /
: 30.3in. :
- N - 43-inLEAD\ [fis70]
E-) . \,b /16 in. STEEL: y o
. ' . 7
€ REACTOR - 0/03 1-inRIB
A |
Vv r
bt 24,38 in.——————

 

 

e———————— 28.75 in, —————( 188010

Fige 1.4.8. Steel Backup Shell for Equutfiricl Lead

Shield. {(Eonfidenticlwith-eapticr)

PERIOD ENDING SEPTEMBER 30, 1957

indicate that it may be necessary to fit the
segments together carefully in order for the
assembly to act as o single structure, This will
also require shear attachments along the joints
between segments on both the inner and the outer
surfaces of the bag. '

Stress Anclysis of a Pressure-Measurmg
Instrument

B. Y. Cotton

An analytical study was made to determine the
elastic stresses in a pressure-measuring device
manufactured by the Taylor Instrument Company
(Fig. 1.4.9), The instrument consists of a riser,
two flat circular plates, a cylindrical portion
containing the weldment, and a pressure-trans-
mitting capillary. A sandwich-type diaphragm of
five layers of 0.005-in,-thick Inconel sheet sepa-
rates the fluid being measured from the trans-
mitting fluid,

Under the specified operating temperatures,
creep is on important consideration, and the
pressure forces on the circular flat plates will
tend to distort them into spherical shapes. If
the pressure forces are applied for extended
periods of time, the creep deformation will cause
an appreciable increase in the internal volume
of the instrument and will result in instability
or drift.

The anticipated operating conditions for the
instrument vary widely, since the instruments
are to be employed in various test loops in
different fluid environments, pressures, and tem-
peratures. The results of the stress analysis are
summarized in Teble 1.4,1 in order to provide a

basis for evaluating particular applications. The

_indicated estimated life is the time required to

- Table 1.4,1. Life Span of a Pressure-Measuring
- Device Based on 0.2% Creep Deformation

 

Internal Pressure Maximum Stress Estimated Life

 

. . at 1300°F
(psi) : (psi) (hr)

25 | 3,546 2000
50 S 7,092 . 50
100 14,185 0
200 28,371 0

 

77

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

 
  
    
 
 

i il ks S A A— —
——— — ————— ———

UNCLASSIFIED
ORNL—-LR-DWG 25894

SANDWICH-TYPE DIAPHRAGM

TRANSMITTING FLUID

\ HIGH-STRESS
REGION

s~ FLEXIBLE TUBING

CAPILLARY

Fig. 1:4.9. Pressure-Measuring Device.

exceed 0,2% creep elongation. The analysis
indicated that this device should not be used at
1500°F and above under the stress conditions
indicated in Table 1.4.1. In applications where
the dimensional stability of the unit is not the
most important criterion, testing would be de-
sirable in order to demonstrate the extent of the
useful life.

APPARATUS FOR ETR IN-PILE TESTS
OF MODERATOR MATERIALS
G. Samuels

A preliminary layout of an apparatus for in-pile
testing of moderator material in the temperature

range of 1500 to 2000°F has been completed.

The apparatus is designed for testing six moderator

78

specimens at one time. The specimens will be

3 in. long ond will vary from ‘/2 to 1 in. in

diameter. The initial moderator materials to be
tested are beryllium oxide, yttrium hydride, and
beryllium metal.

The purpose of the tests of yttrium hydride and
beryllium oxide is to determine their maximum
allowable thermal stress under various reactor
operating conditions. In the case of the beryllium
specimens, the thermal stress consideration is
secondary to the study of compatibility with
various canning materials.

The moderator materials will be canned in
Incone! jockets, and, where necessary, a third
material will be used between the moderator and
jacket.  The beryllium will be separated from

>

 
 

 

 

L}

the Inconel by either chromium, molybdenum, or
an oxide coating on the beryllium. For yttrium
hydride, the intermediate material will be mo-
lybdenum.  Beryllium oxide will not need the
buffer material.

The jacketed moderator specimens will be
centered in a water-cooled nicke! sleeve. The
clearance between the specimen and sleeve will
vary from 0.015 to 0,060 in. depending on the
moderator material, the diameter of the specimen,
the location in the reactor core, and the desired
minimum surface of the specimen. The heat from
the specimen will be removed by radiation and
conduction to the nickel sleeve. Helium or o
mixture of helium ond orgon will be bled slowly
through the gas space. For normal operating
temperatures {about 1500°F), 80% or more of the

PERIOD ENDING SEPTEMBER 30, 1957

heat flux is carried by conduction so that the
surface temperature will be very sensitive to
the argon content of the gas, which will thus
provide a convenient means of controlling the
temperature,

Two thermocouples will be spot-welded to the
jacket of each specimen at equal distances from
each end and 180 deg apart. In addition each
BeO specimen will have one thermocouple located
in the center of the specimen at a distance of
3’8 to 3, in. from the end. Only one specimen each
of yttrium hydride and beryllium will contain a

" center thermocouple during any test run, Knowl-

edge of the tempercture difference between the
center and the surface of the specimens will
facilitate the determination of the internal hect
generation rate and thermal stress,

W
o 56
i

79

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

1.5. DESIGN PHYSICS
CAM. Perry

ART FILL-AND-DRAIN TANK SHIELDING
H. W. Bertini

The general method of attack on the problem of
shielding the ART fuel fill-and-drain tank was to
choose a field point on the surface of the shield
and calculate the dose at this point from radiation
emitted from the surface of the tank. '

The sources at the surface of the tank were
considered to be point sources of magnitude

Sv ' - "~ Mev
— cos 0 x,A(———-——) ,
4p sec-steradian

S = Mev/sec.cm? of fuel volume,

g = total linear gamma-ray cross section of
the fuel (cm="),

6 = angle between the normal to the dump tank
surface and the line between the field
point and the point source,

A = area of the dump tank surface represented
by a point source at its center.

The (S, /4mu) cos 6 term is the surface source
strength at the interface between a thick slab
source and a thick slab shield that will give,
essentially, the correct dose rate at the surface
of the shield.

To the extent to which it was completed, the
lead shield was designed so that the dose rate
at any point on its surface was less than 0.2 r/hr
after 100 hr of operation at 60 Mw, 9 days after
shutdown.

Under these conditions, S was calculated for
a fuel volume of 2,5 x 10° em®, and the curves
of decay energy after shutdown given by Moteff!
were used. The 1,65 and 2,55-Mev groups were
the only ones used in these calculations, and the
p volues for the fuel were taken from ORNL-2113
(ref 2) for these energies. In all cases, cos O
was set equal to 1.

 

J -Moteff, M:sce!laneous Data for Shielding Calcu-
latzons, APEX-176 (Dec. 1, 1954).

2H. W. Bertini et al., Basic Gamma-R Data {or ART
Heat Defosz'tion Calculations, ORNL-2 3, p 61 {Sept.
17, 1956).

80

Penetrations

‘The problem of penetrations in the shield is

aggrovated by the need to allow for relative motion

of the tank, all pipes connected to'it, and the
tank shield. All penetrations through the shield
are therefore larger than would ordmurlly be
anticipated. '

In the region where the lead shield for the drain
line meshes with that of the tank the dose rates
at 16 field points from 6 point sources were
calculated by using the technique described above.
In addition, the dose rate ot a point at the upper-
most surface of the tank shield from single
scattering in the lead around the drain line was
calculated and found to be within design limits.
At another point in this general region, where
only 3 in. of lead separates it from about :?4 in.
of unshielded drain line, the dose rate was cal-
culated and found to be acceptable, The valve
actuator penetration through the lead allowed the
dose rate in this vicinity to be rather high. A
2-in.-thick, 8-in.-dia disk of lead fitted around
the actuator arms in this region would reduce the
dose rate to acceptable limits,

The pipes for carrying the NaK to heat and cool
the fuel cannot be allowed to penetrate straight
through the shield, because the dose rate at such
penetrations would be prohibitive. A feeling for
the magnitude of the problem can be obtairi¢d by
noting that the dose rate at the surface of the
tank is of the order of 2 x 10° r/hr 9 days after
shutdown from operation for 100 hr at 60 Mw.

By repeated calculations of the direct and
reflected radiation and the radiation incident at
slant angles it was found that the geometry
sketched in Fig. 1.5.1 would meet the shielding
requirements for penetrations in the tank. The
direct radiation was calculated by using the point
source concept outlined above. The reflected
radiation was calculated by using the data given
by Rockwell® to estimate the angular distribution

 

37, Rockwell, NI (ed.), Reactor Shielding Design
Manual, T|D-7004, P 330 (March 1956).

™

i

 
 

 

 

 

 

 

4.

~ dose rate was ca_lcul_ated to be about 2 x 108 t/he.

SESREF
ORNL —LR—DWG 25835

NoK LINES LEAD SHIELD

——\[
=

 

 

 

   

 

 

 

 

 

 

 

 

 

——TANK

 

 

MANN

Fig. 1.5.1. Top of ART Fill-and-Drain Taenk and
Shiefd Showing Recommended Configuration for NaK
Lines Penetration,

of reflected radiation and the data given by Zerby*
to estimate the fraction of the incident radiation

that is reflected. The data of Rockwell do not
correspond to ART fuel fill-and-drain tank ganima--
ray energies or material, and the data of Zerby

do not correspond to ART fuel fill-and-drain tank
gamma-ray energies. This calculation appears to
be about all that can be done at this time, how-
ever, and the calculated dose rate is probably in
error by a factor of the order of 2. The dose rate
for radiation incident at slont angles was col-
culated from the work of Peebles.® v

At the bottom of the reactor, the fill-and-drain
tank support penetrates the lead shield.  About

2 in. below this penetration the dose rate was

calculated to be ‘about 2.5 x 104 r/hr._ In the_

vicinity of the top_of the air cylmder for the tank
support, about. 29 in. below the penetration, the

Overlcp m Shleld at Removcble Top Section B

In order ‘o facnhtate assembly ‘and dlsassembly,

it was necessary to deSIgn the tank shleld '$0

 

4C.D.Zerb Transm:sszon o[ Obl:quely lnczdent

Gamma-Rad:‘atz‘on Tbrousb Stratified - Slab’ Bamers,-

ORNL-2224 vol 1 (Nov.
1957).

56. H. Peebles, Gamma-Ray 'I'ransm:sszon Through
Finite Slabs, R-240 (Dec. 1, 1952).

9, 1956) and vol 2 (Jan. 2

PERIOD ENDING SEPTEMBER 30, 1957

that the top portions of the lead would be re-
movable in sections. The problem was to de-
termine whether any overlap of these sections
would  be required when they were in place on
the tank.

- By using the method descrlbed above, the dose
rate at a point opposite a /l -in. crack in the
lead which would expose the tank surface was
calculated to be about 120 r/hr. Therefore a
clearance of this size. separating the sections
of the lead would be much too large. |

If the sections of the lead were brought togefher
with no c!earance, but in such a manner that the
%fi-m. steel cladding layers around each lead
section touched and formed a / -in. slab of iren
between the adjoining lead sechons, the dose
rate at the outer surface of the slab of iron would
be about 2 r/hr, which is also not acceptable,
it was therefore concluded that all adjoining
pieces should be overlapped ot about their mid-
thickness.

Incompleted Work

Additional calculations are required for the
shielding of the small lines which connect with
the upper region of the foce of the tank and
penetrate the shield.  Calculations are also
required to determine the dose rate near the
bottom of the shield as a result of scattered and
direct radiation through the penetration in the
lead required for the tank support.

DOSE RATE IN REGION OF ART PUMPS
H. W. Bertini- C. M. Copenhaver
In order to ascertain the feas:blhty of having

‘a man_enter the ART ‘reactor cell to unbolt an
‘inoperative fuel or sodium pump for removal, the

_dose .rate was calculated at a point, P, 62.5 in.
' _above the reactor midpoint and 22.5 in. from the

axis, such point being in the vicinity of the bottom

 of the pump motor. The condition chosen was

9 days cfter shutdown followmg 100 hr of reactor
operation at 60 Mw..
‘The  sodium activation contnbutlon was found

~:-to ‘be -0.23 r/hr, .of which 0.05 ¢/hr. would be
~ through lead, 0.13 r/hr would be through NaK

pPipe - penetrations to the sqdium-to-NaK_' heat ex-
changers, and 0.05 r/he would be through other
penetrations. The corresponding sodium dose rate

at shutdown is 5000 r/hr.

81

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

The fuel contribution would be principally from
residual fuel pockets whose probable volumes and
locations are presently not well known. The fuel
pumps, the fuel onnulus around the control rod
in the sodium expansion tank, etc., would con-
tribute about 3 r/hr, with about 2 r/hr being
through the fuel pump . penetrations. On the
assumption that 2% of the total fuel in the reactor
system would remain uniformly distributed in
the main fuel-to-NaK heat exchanger upon draining
of the reactor, the dose through the NaK pipe
penetrations would be about 500 r/hr.

The off-gas line will pass very close to one
of the fuel pumps and part of it will be visible
through the pump penetration. A total of
3 x 1015 Ba'4® nuclides per second® was assumed
to pass into the off-gas line and deposit uniformly
on its surfaces. If it is assumed that a 3-cm
length of off-gas line will be directly visible
through the penetration, the dose rate was cal-
culated to be 260 r/hr from this source. A
summary of this preliminary calculated dose rate
at P is given below:

Source Dose Rate (¢/hr)
Sodium 0.2
Fuel pockets 3
Residual fuel in heat exchanger 500
Off-gas line 260

DECAY OF GAMMA ACTIVITY IN U235
CONTAINING ART FUEL FOR SHORT
TIMES AFTER SHUTDOWN

J. Foster

The rates of decay-gamma energy emission from
U235 fuel for the first 180 sec after reactor
shutdown following 500 hr of operation were
calculated for six gamma-ray energy groups. The
results (for the six energy groups) were plotted
(Fig. 1.5.2) as Mev/sec for an operating level of
one fission per second as a function of time after
shutdown.  Although data have been available
for the rate of decay of gamma activity for times
of a few minutes or more after shutdown, no data
have been available up to this time for the period

 

66, Samuels, private communication to H. W. Bertini.

82

Py
-LR-DWG 25896

   

DECAY ENERGY [(Mev/sec)/(fission/sec)]

2 5 2 5 ¢ 2 5 10 2 5 2x10
TIME AFTER SHUTDOWN (sect

Fig. 1.5.2. Decay-Gamma Energy vs Time After Shut-
down for Uzss-c‘\miulning Fuel Operated for 500 hr at
60 Mw in the ART. Decay energy calculated for the
six gamma-ray energy groups indicated.

immediately ofter shutdown. These short-time
decay rates are important in calculating heating
rates and cooling requirements for such system
components as the fuel drain valves. Results
of this study yield a total decay-gamma energy
of 5.83 Mev fission at 10=3 hr after shutdown.”

DECAY-GAMMA HEATING IN ART FUEL
DRAIN YALVES

J. Foster

Calculations were made of the heat deposition
rates in the bellows and stem of the ART fuel
drain valves by decay-gamma energy originating
in fuel draining through the valve immediately
after shutdown following 500 hr of 60-Mw oper-
ation. ' :

The total decay-gamma energy deposition rate
at the instant of shutdown was calculated to be
2.5 w/cm® of Inconel, which corresponds to a
rate of temperature rise of about 1°F/sec. |f it
is assumed that fuel remains in the valve for
3 min, which is the maximum allowable duration
of the fuel draining operation, and the decrease
in gamma activity during this period is aliowed
for, a total temperature rise of 106°F is obtained.
No allowance was made in the calculation for heat
loss from the Inconel piping during the 3-min
pericd. ' :

 

7The basic experimental data were taken from interim
unpublished work of W. Zobel oand T. A. Love cf ORNL
and were modified by C. Copenhaver of ORNL, '

y

 
 

 

 

 

 

W

PERIOD ENDING SEPTEMBER 30, 1957

1.6. MATERIALS AND COMPONENTS INSPECTION
A. Taboada

MATERIAL INSPECTION

A. Taboada G. M. Tolson
Tubing

Approximately 3 miles of tubing, 15,827 ft, was
inspected during the quarter, and, of this omount,
4126 ft (26%) was rejected. The rejected tubing
included 1109 ft that can be salvaged by pickling;
1093 ft of the rejected tubing was returned to
the vendor for credit.
material  was downgraded for Iess critical ap-
plications.

included in the. tublng mspecfed was 128 ft-

of INOR-8 tubing of which 5 ft (4%) was rejected.

Typical defects found on the INOR-8 tubing are-

shown in Figs. 1.6.1 and 162 ond a defect not
previously observed is shown in Fig. 1.6.3.

As mentioned previously,! a phosphoric acid
pickling method has been developed for removing

 

‘G M. Tolson, ANP Quar. Prog. Rep. June 30, 1957,
ORNL-2340, p 75.

0.25in.

 

L
¥ 1

‘The remaining rejected .

magnetic particles stuck to the inside of CX-900
subing. In order to determine whether there
would be any harmful effects from the pickling
solution, a sample of Inconel tubing was immersed
in phosphoric acid for 160 hr and then examined
metallographically. No attack was found, and it
‘therefore appears that the pickling method may
be used satisfactorily for the removal of magnetic
particles,

A large number. of oufer surfuce scratches are
found during the inspection process, and, although
the scratches are usually superficial, they interfere
with subsequent ultrasonic tests. Investigations
have revedled that intermittent polishing and re-

“polishing can reduce the rejection rate substan-

tially. Tubing rejected previously by ultrasonic
inspection is being re-evaluated.

Pipe, Plate, Sheet, Rod, and Fittings

A total of 532 ft of Inconel pipe was inspected,
ond 73 ft (14%) was rejected. Also, 1027 f? of

UNCLASSIFIED
T-12859

 

Fig. 1.6.1. Photomacrograph of INOR-8 Tubing Showing Two Cracks Which Penetrate Completely Through the
Wall of the Tube. The cracks were found by radiographic inspection, 5X.

83

 
 

 

 

ANP PROJECT PROGRESS REPORT

0.25 in.

 

y ~ UNCLASSIFIED
! T-12860

 

Flg. 1.6.2 Ph'dfbmacrogruph of VGouge Found on Inside Surface of INOR-8 Tubing by Radiographic Inspection.

5X.

plate and sheet was examined, of which 47 2
(4.6%) was rejected. The inspection of 81 ft
of rod resulted in the rejection of 18 ft (22%).
There were 41 f{ittings inspected, and 5 were
rejected.

Embrittlement of Inconel by Penetrants

An experiment, described previc:msly,2 was con-
ducted to determine whether the solutions used
in penetrant inspection would couse embrittlement
of Inconel at high temperctures. Metallographic
examinations of samples which had been heated
at 1500°F for ¢ period of 50 hr showed no embrit-
tlement or attack as a result of exposure to the
penetrants. ' '

Chemical Analyses of Materials

Chemical anclyses of materials received for
inspection are now being obtained by spectro-
graphic techniques. A comparison of spectrographic
results with the standards established by wet
chemical analysis indicates an accuracy of +10%.

21bid., p 80.

84

Electrical Calrods

Radiegraphic inspection was made on 1964 ft
of electrical Calrods in order to determine the
positioning of the conduit. Those units which
showed improper positioning or faulty conduit
were either rejected or downgraded.

WELD INSPECTION
A. Taboada R. L. Heestand

Welds fabricated on ART and ETU components
and on experimental test loops by the inert-gas
shielded-arc method were inspected. The defective
conditions found by inspection were porosity,
cracks, tungsten inclusions, lack of fusion, and
iack of penetration. Several welds were examined
which had been fabricated on core shells for test
purposes. In general, these welds were found to
be acceptable for the intended use; however, in
some instances defects were noted that would have

been cause for rejection or repair if they had been

found on ETU or ART parts.

 
 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

0.25 in.

UNCLASSIFIED
T-12868

 

 

 

"

 

 

 

 

 

 

 

 

 

 

 

Fig. 1.6.3. Photomacrograph (a) and Photomicrograph (5) of. Defective INOR-8 Tubing Located by Radiographic
lnspeeti!fiou. (a) 8X. (&) S0X. Etchant: giyg:e;egid."””- 7

85

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

INSPECTIONS OF REACTOR SHELLS
A. Taboada G. M. Tolson

Inner Core Shells (Shel_l 1)

Radiographs of inner core shell sections made
by forming, welding, and machining were found
to have *‘white lines’’ in the weldments. Metal-
lographic examinations showed the lines to be
associated with large dendritic grains that ex-
tended through the entire weldment.

Outer Core Shells (Shell 1)

Inspections of spun outer core shells described
previously as “‘in process’’ were completed. The
results of all the inspections are presented in

Table 1.6.1.

It is to be noted that none of these shells are
to be used in the ETU and ART. Complete in-
spection was performed in many cases for checking
manufacturing techniques and quality of the finished
parts and for obtaining data for use in other tests.

Shell 1il

The bottom section of core shell Il was in-
spected by radiography following an intermediate
machining operation. One weldment was found to
contain pores and areas that were contaminated.
No repairs were made on the weldment because
the poor areas may be removed during subsequent
machining. The shell will be inspected again
in the final form. A

Shells IV and V

Shell material formed by deep-drawing in the
Y-12 shops was examined to investigate the
method that is to be used by Kaiser Metal Products
in the fabrication of shells IV and V. Specimens
consisting both of weld deposits and of the parent
material were found to be acceptable.

Additional specimens of as-received material and
of annealed material (1800°F for 10 min) which
haed ruptured during fabrication ‘were received
from Kaiser Metal Products. Rockwell hardness

tests were performed (Table 1.6.2), and the

ductility of the material in various stages of the

Table 1.6.1. Results of Examinations of Spun Outer Core Shells

 

Metallographic

Nonde structive

 

Lot No. Piece No. Examination of Sample Inspection of Shell

1 1 Not required Rejected for surface defects

1 2 Not required Rejected for defects

1 5 Rejected because of Not required

overpickling*
4 12 Accepted* Not reguired
3 | 1 Rejected Thin wall areas
4 Accepted Rejected by visual inspection

because of overpickled con-
dition

4 8 Accepted Not required

4 n Accepted Not required

4 10 Rejected Thin wall areas

4 4 Accepted* Not required

4 7 Rejected Rejected on visual and dye-

penetrant inspections; thin

wall areas

 

 

*Metallographic specimens obtained from cut-up shell.

86

 
 

 

 

 

Table 1.6.2 Hardness and Ductility of
Inconel Specimens Recelved from
Kaiser Metal Products Company = .

 

Hardness, Estimated Elongation

 

Specimen Rg (% in 2 in,)
As received* 80 41
As drawn* = 93 . 22
Annealed | . 73 ‘ . 43

 

*A control sample of this material has been taken for

subsequent comparisons.

shell-forming process was extrapolated from the -

hardness data. Although data of this type are
reasonably accurate, tensile and elongation tests

are also being performed, since ‘large variations.
in mechanical properties are usudlly found in
different heats of the material and occasionally"

in the same heat. It may be noted from Table
1.6.2 that the hardness of the annealed material
is not so low as that sometimes found in deep-
drawn good-quality Inconel. Therefore it has been

recommended that the material be annealed at

1800°F for 20 min in the future.

INSPECTION OF COMPONENTS RECEIVED
FROM YENDORS

R. L. Heestand G. M. Tolson

York Corp. Radiators und Radiator Matenuls

‘Radiegraphs  of - the tube-to-header ‘welds of -
finned-section units 9, 11, 12, and 14 of the main -
ETU NoK-to-gir radiators bemg supphed by the
York Corp. were reviewed to assure ‘that they'
had been: properly interpreted. Rejection of units
9 ond 11 by the ORNL inspector stationed at the
York .Corp. was affirmed; unit 12 was found o
be acceptable; and recommendations - were made
- for. the repair of two porous.welds on unit 14,
Radiator - units 18 and 19 were inspected and
" found to.be- ucceptable, and unit 20 is bemg ;
.mspected _ =

- Two " l-Mw test rad:ufors, units 2 und 3, were
recelved for inspection,  Unit 2 was found to be

acceptuble, but it was found that the return bends
of five of the tubes of unit 3 had been domaged
during installation of a sawing fixture for the

PERIOD ENDING SEPTEMBER 30, 1957

slitting of the fins. The deepest defect was
repaired by welding.
A header piece for the south head of the main

- fuel-to-NaK heat exchanger that had been fabricated

for stress onalysis tests was also inspected.
Several pinhole leaks found in the brazed tube-
to-header joints were repaired by the ORNL
Welding and Brazing Laboratory by flowing Coast
Metals brazing alloy No. 52 over the pinholes
with on acetylene torch. The necessary inlet and
outlet tubes were also installed, and the unit
was submitted for testing.

inspections of samples of stainless-steel-clad
copper fins on which the cladding appeared to
be blistered revealed a thin layer of oxide on

‘the stainless steel surface that was of sufficient

thickness to prevent bonding. It was recommended

“that all fin material thot showed defects of this
. type be rejected. '

Control samples, which duplicated units 5, 6,

7, 8, and 12, were examined metallographically

before and after exposure to air at 1500°F for

100 hr. Control samples 5, 6, 7, and 8 were

acceptable, but sample 12 showed lack of flow
of the brazing alloy, and 60% of the fin lips
were found to be unprotected after oxidation.

‘Recommendations were made to the York Corp.
for the use of jigs and fixtures in making the

-welds of tube-to-header joints in order to prevent

the braze cracks found on ETU radiator unit 1.

. The eracks resulted from excessive weld shrinkage.
‘A procedure for the repair of the cracked unit
~was also recommended. Spacers for use during

brazing were prepared and sent to York Corp.

These Inconel X spacers were fabricated by rolling

and processing in a hydrogen atmosphere for 30

~min at 1950°F; 200 spacers were prepared.

.Griscom'-Rus'séll Co. Heat Exchangers

Sodlum-to-NoK heat exchanger vnits 1 and 2

'._‘Zfabrlcated by Griscom-Russell Co. were submitted
~ for .inspection prior to .installation in the ETU
- north. head.. Root passes of the welds on each

head of both heat exchangers showed lack of

_ penetration because of shrinkage and inadequate

- repair. -work. Cleanup passes will be made on
future “units to eliminate defects of this type.
‘Since no defects serious enough to cause failure

of the units were found, they were accepted for
use in the ETU system.

87

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT.

A K-fin radiator job sample was examined and

~ rejected because of a manufacturing defect imposed

on the tube by the winding of the fin. It was re-
quested that the tube-sheet design for the K-fin
radiator be revised because there are areas in
the present design which would be stressed
beyond the yield strength of the Inconel.

The foliowing procedures and samples were
examined and opproved: (1) o tube-to-header weld
and braze sample with eight tubes; (2) three tube-
to-header welder-qualification samples; (3) a fuel-
to-NaK heat-exchanger-header assembly sequence;
(4) a dummy sodium-to-NaK heat-exchanger-header
assembly; (5) a joint change of a sodium-to-NaK
heat-exchanger nozzle-to-header weld; and () a
sodium-to-NaK heat-exchanger header for stress-
analysis testing.

Black, Sivalls & Brysoh Heat Exchengers

A resistance technique for bending the fuel-to-
NaK heat exchanger tubes was approved after
destructive and nondestructive exomingtion of ten
tube samples submitted by Black, Sivalls & Bryson.
North and south header stress analysis samples
were received, and after repairs of several cracks,
the north head sample was accepted for testing.
The south head stress analysis sample is being
inspected.

The following reports, procedures, and samples
were examined and approved: (1) radiographic
techniques for examination of tube-to-header welds;
(2) a report covering the modification of the
channel-to-web weld joint; (3) samples of and a
report on the bulkhead weld; (4) a report on tube-
to-header brazing covering the use of rings con-
toured to fit the header; (5) braze samples 1 and
2 that illustrated vertical and horizontal brazing;
(6) proposed welding techniques for the channel,
flute filling, inlet pipes, outlet pipes, and the
bulkhead-to-tube-sheet joints.

Midwest Piping Company Forgings end Weldments

Nondestructive and destructive examinations were
completed on samples of hot-forged Inconel re-
ducers which were fabricated at temperatures of
1800 to 1850°F by the Midwest Piping Company.
Metallographic examination revealed the grain size
to be no greater than No. 5, and therefore the
fabrication method was approved.

An Inconel pipe saddle weldment was examined
and approved for design and weld integrity. Also

88

a 90-deg bend in a piece of 3‘/2-in. Inconel pipe
was checked for wall thinning, flattening, and
other defective conditions and was found to be
satisfactory. ' o ’ :

Process Engineering, Inc., Tanks and
Bellows Expansion Joints

Inspection was completed on standard NaK tank
No. 2 received from Process Engineering, Inc.,
and it was found to be acceptable for the intended
use. Four other tanks have been received and
are being inspected. Twelve bellows expansion
joints for the ART were also received for in-
spection. '

Fulton-Sylphon Company Bel lows

A sample ETU sodium expansion bellows was
received from the Fulton-Sylphon Company for
evaluation of fabrication techniques, Examination

- revealed that the seam welds on each end of the

unit were porous. Metallographic examination re-
vealed dirt imbedded between each of the three
plys of the bellows. In the weld metal, pores up
to 0.005 in. in diameter were found. It was recom-
mended that more care be taken in the fabrication
of the bellows to eliminate sources of dirt and
contamination.

Hoke, Inc., Inconel and Stainless Steel Valves

Fifteen Inconel and fifteen type 316 stainless
steel valves, which have two welds each, were
received from Hoke, Inc., for acceptance of the
welds. Inspection resulted in the detection of
five defective welds., These welds were repaired
and were found to be acceptable upon re-examination.

METAL IDENTIFICATION METER
A. Taboada  G. M. Tolson

The examination with the metal identification
meter of an ART-type fuel pump after test operation
indicated that a material other than Inconel had
been used. Further tests with the metal identifi-
cation meter and a Spotchek chemical kit revealed
that the labyrinth seal was fabricated of a 300-
series stainless steel. Since material other than
Inconel has been detected in other components
by this equipment,® two additional meters have

 

3G, M. Tolson and R, L. Heestand, ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340, p 78.

i

 
 

 

 

]

been ordered for adequate inspection to ensure
that only Inconel is used in the ETU and ART.

The meter now available is being used to check
all components being received from vendors.  The
second meter will be used in the Y-12 Plont
General Shops in order to check all incoming
materials and to check all components that are
fabricated there. The third meter will be used
to check all material that is welded in the welding
shop established for the assembly of the ETU
and ART.

INSTRUMENTATION AND CONT ROLS
INSPECTIONS

Nlckel-Copper Welds in quuld-Level Probes

Several ‘failures- have occurred in the nickel<

copper welds in liquid-level probes. A microscopic -
examination of one of the failed welds revealed -
stringers of copper oxide adjacent to the ‘failure

zone. Radiographs of several other probe welds
showed large pores, whlch indicate high. oxygen
content of the copper.
failed are being examined to obtain comparative
information that will be useful in further investi-
gation of the cause of failure.

A liquid-level indicator which cracked and Ieuked

in five areas was found to have failed because

of ratcheting of the ceramic insulators. The tube
failures were typical of tube-burst test failures
under biaxial loading. A second liquid-level in-
dicator failed because of leakage through a pin-
hole. The initial radiographs taken of this area

were re-examined, but no indication of the defect

was found. Since the unit had also passed dye-
penetrant inspection, the pinhole probably origi-

nated in the inner lead ~weld which- ‘was’ not

accessible for dye'Penetrcmt inspection.  ~

Thermocouple Welds

A test fhermocouple which failed durlng testmg -

in NaK at 1500F was examined mefollographucall;
The failure was found to have occurred in the

closure weld of the thermocouple sheath and was
due to lack of penetration of the weld. - At ‘the
point of fallure ‘the magnesium- oxide insulating

material had been leached to a 'depth of approxi-
mately Y% in. during the 5000 hr of exposure to
NaK ot 1500°F, and NaK had diffused through the

insulating material approximately 10 in. No

Probes which have not =

PERIOD ENDING SEPTEMBER 30, 1957

deterioration of the metal parts of the unit could
be seen. The failure zone is shown in Fig. 1.6.4,
and the thermocouple assembly is shown in Fig.

1.6.5.

AR LUNCLASSIFIED
T-12895

 

Fig. 1.6.4, Sectioned Thermocouple Assemblies After
Service in NaK at 1500°F.
operated satisfactorily for 5000 hr, The unit shown

The unit shown in (a)

in (b) failed because of a crack in the closure weld
on the sheath., Note removal of insulation by reaction
with NaK. The dark insulation visible in (b) is
satyrated with NaK.

UNCLASSIFIED
T-13094

   

\‘<\\ FAILURE AREA SHOWN

IN Fig.1.6.4 2 ' &
WNEA W\
kN %‘\

‘P:\\O‘A‘ g\\b\}\

6 N

?E‘A \\
o S

o

Fig. 1.6.5. Thermocouple Assembly Which Failed
During Exposure to NaK for 5000 hr at 1500°F,

89

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

PROCEDURES FOR RADIOGRAPHIC
INSPECTION OF WELDMENTS ON ART ISLAND
AND REFLECTOR

A. Taboada G. M. Tolson

There are seven critical welds on the ART and
ETU which can be radiographically inspected only
through sections of beryllium. Normal radiographic
techniques .cannot be used because beryllium
scatters the .incident x-’rqy beam and causes
general fogging of the film. Experiments are being

90

run in order to determine techniques for obtaining
maximum sensitivity. o

Several radiographic fechniques ‘were used in
the inspection of the equatorial weld on the ETU
island shell assembly, but it was impossible to

estimate the radiographic sensitivity obtainable

by these methods. A mockup of the assembly with
known discontinuities was therefore obtained,
and the sensitivities of the inspection methods are
being investigated. '

 
 

PERIOD ENDING SEPTEMBER 30, 1957

1.7. HEAT TRANSFER STUDIES
H. W. Hoffman

THERMAL.CYCLING RESEARCH
Pressurized System
H. W. Hoffman D. P. Gregory!

The experimental investigation of the effect of
thermal-stress cycling on Inconel tubes filled with
flowing NaF-ZrF -UF, (50-46-4 mole %) was
continued with the use of the pressurized-flow
system. Data were obtained at temperatures near
1600°F for tube bends, tube welds, and straight

 

10n assignment from Pratt & Whitney Aircraft.

tubes cycled at frequencies of 0.4 and 1.0 cps.
The results of the experiments performed during
the quarter are presented in Tables 1.7.1 ond 1.7.2
and are summarized in Fig. 1.7.1.

Straight tubes were tested at a cyclic frequency
of 1 cps for comparison with the data of experi-
ments made at frequencies of 0.01 and 0.4 cps.
As previously reported,? the tests ot 0.01 and
0.4 cps frequencies indicated no significant effect
of cycling frequency on depth of attack. However,

 

24, W. Hoffman and D. P. Gregory, ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340, p 82.

Table 1.7.1. Results of Thermul-CycIIng Test of Inconel Tubing Containing
Flowing NaF-ZrF -UF, (50-46-4 mole %)

Cycling tate: 0«4 cps
Tubing size: 1{‘«tn. 0D, 0.035+in. wall

Heater length: 8 in.

 

 

Test Conditions ET-27% _ ET-29® ET-30 ET-31 ET-32 ET-33 ET-34 ET-35 ET-36 ET-37
Duration of run, hr 4 n 65 30 25 54 19 45 11.5 15
Heater section

Outer wall temperature, °F
Average 1597 1795 1725 1769 1767 1763 1716 1700 1664 1648
Fluetuation g5 116 197 197 196 tg9 177 180 170
Maximum depth of attack, mils

General subsurface veids 2 2 4 6

intergranular attack 3 4 4 5 105 & 8
Test section

Outer wa!l temperature, °F : B . :
~ Average 1597 1623 1604 1590 1613 1620 1598 1626 1590 1584

Fluctuation - 3 - #1314 16 113 t15 15 e e
 Maximum dqfith of attack, mils e

Generel subsurface voids - 2 5 4 4.5 7 3 35

. Intergranular attack 3 . 5. 4 35 7 6 - 35
inlet mixed=mean fused~salt 1597 ‘1573 1578 1573 | 1574 1573 1542 1598 1555 1562
temperature, °F ' ' _
Cuu§§ of test termihatlon Voluntary®

 

%) sothermals
bHigh-frequency cycling rate: 1 cpse

€All other runs were terminated by instrument failure.

9

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 1.7.2. Results of Thermul-CicHnfi Tests of Bent Inconel Tubes Containing Flowing
NuF-ZrF(UF‘ (50-46-4 mole %) Cycled at a Rate of 0.4 cps

 

Attack in Straight
Pertion

Attack in Bends

 

Comments |

 

Run Description _— s ‘
Depth Type of De::th TAY pe :f
{mils) Attack (mils) Hac
ET-30 Prior creep, fine grained, 3.5 General 4 General in all ‘Equ'él.atfocl_: on both tension and
four 90-deg bends bends compression sides of each bend
ET-31 No prior stress, fine 3.5 lnfergrcnulér 4 Intergranular Deepest attack found ne&r center -
grained, four 90-deg in all bends of each bend with no difference
bends " between tension and come
pression sides of bends
ET-32 No prior stress, fine 4.5 Intergranular Tendency toward slightly deeper
grained : ' intergranular attack on tension
Two 30-deg bends 6 intergranular sides of bends
Two 150-deg bends 5 Heavy inter-
& granular
ET-33 No prior stress, coarse 7 General 8 General in ail Equal ottack on both tension and
grained, four 90=deg bends compression sides of bends
bends
ET-34 No prior stress, coarse 3 General 3.5 General in all Egqual attack on both tension and

grained, two 30-deg
bends, two 150-deg
bends '

bends compression sides of bends

 

UNCLASSIFIED
ORNL~-LR-DWG 25897

90-deg 8?7 o 0O.4cps
//‘)c'
-
30-deq /’0
BEND-~® o

Ve o
pEN
o .go-deg /<
&
i q'(ao—deg BEND
HO; = +
;{ PRIOR CREEP

 

 

 

I

 

~1SOTHERMAL

 

 

{ 1.0cps~COARSE GRAINED
O O.4¢ps-COARSE GRAINED —
® 0O.4cps- FINE GRAINED

O I1SOTHERMAL -COARSE GRAINED
& ISOTHERMAL- FINE GRAINED

[ |
40 60 80 100 120 140
TEST DURATION (hr)

 

MAXIMUM DEPTH OF ATTACK (mils)

 

 

 

 

 

o
n
o

Fig. 1.7.1. Summary of Data Obtained in Thermal-
Cycling Experiments with Incone! Tubing Filled with
Flowing NnF-ZrF‘-UF‘ (50-46-4 mole %), (Seeretauith
“apsicn)

92

preliminary results of @ run (ET-29, Table 1.7.1)
at 1 cps showed test-section attack that was
somewhat greater than that observed at the lower
frequencies. The low temperature amplitude (13°F)
at the outside wall of the test section resulted from
the increased cycle frequency. The observed
intergranular attack of 5 mils with a temperature
tluctuation of only +3°F indicates that the extent
of the attack on Inconel is frequency dependent,
with a critical frequency in the vicinity of ‘é to
1 cps. Additional data are currently being ob-
tained at the 1-cps frequency in order to verify this
conclusion.

The effect of prior strain on the extent of the
corrosive attack on Inconel under thermal cycling
conditions was investigated in the series of runs
ET-30 through ET-34 (Table 1.7.2). For run ET-30,
a fine-grained tube was prestressed to give o fixed
strain, allowed to relax for 24 hr at 1300°F, and
then bent, as shown in Fig. 1.7.22. Because of

 
 

 

 

 

o

UNCLASSIFIED
ORNL-LR-OWG 25898

 

 

 

 

t Nl .

i: H e 2 i1, —

Il oy ‘

ELECTR i

“,//— ODESNI & T -
| , e '
| e

ili ' i!’ S *. -9

il HEATER SECTION it ‘é‘ . l .

 

 

LF— 8in. : =

TEST SECTION
{a)

HEATER SECTION

 

 

 

 

P

 

 

(5)

TEST SECTION

Fig. 1.7.2. Test Section Configurations Used in the
Determination of the Effect of Thermal Cycling on Tube
Bends. {2) 90-deg bend. (b) 30-deg bend.

the crudeness of the creep apparatus, the exact
strain imparted is not known. However, it is felt
that the results give at least a qualitative estimate

of the effect of stress. Runs ET-31 through ET-34-

were performed with annealed (as-received) tubing.
Two of these tubes, ET-31 (fine grained) and
ET-33 (coarse grained), were bent as shown in
Fige 1.7.2a; while the other two tubes, ET-32

(fine grained) and ET-34 (coarse grained), were
- bent as shown in Fig. 1.7.2b. It may be noted from
Table 1.7.2 that in all cases the attack was greater .
in the bends than in the straight sections of the .

tube. However, no difference was detecied between

the attack on the tension and compression sides of
the bends. On the basns of data obtamed by the'_

Experimental Engineering Group in tests of tube

“bundles -in ‘a flucride fuel environment, this resuh!

is somewhat unexpecfed.‘, A" duplicate series of -

| ‘thermal-cycling  tests -with bent tubes-. will be -
‘initiated shorfly thaf wnll be scheduled for 100 hr_-,:'
rof operahon. o '

lt may also be noted in Table ].7.2 that, as.in

__'prewous ‘tests, the fine-grained tubes showed -
intergranular attack and the coarse-gramed tuhes_'i_;,i
showed genera! attack. Photomicrographs showmg .

the attack in two tube bends are presented in
Figo 1.7030

- short . exposures.

PERIOD ENDING SEPTEMBER 30, 1957

Since it is planned to use welded and machined
shells in contact with the fluoride salt in the ART,
a study of the effect of thermal cycling on machined
welds was undertaken. The test sections were
formed by butt-welding two pieces of Inconel tubing
- {0.25-in. OD, 0.035-in. wall thickness). The inside
of the weld was reamed to the original inside tube
diameter, and the outside was machined back to the
original outside tube diameter. The weld was
located approximately 1 in. downstream from the

-heater.  The results of these tests, runs ET-35
- through ET-37, are given in Table 1.7.1. Photo-

micrographs of the weld areas of the specimens
are not yet available. Run ET-35 was made with
a fine-grained tube, while coarse-grained tubes
were used for runs ET-36 and ET-37. Further,
the specimen used in run ET-37 was constructed
by building up ‘4 in. of weld metal on the ends of

~the tubes, machining the weld metal to the original

tube dimensions, and welding the weld-metal

- pieces together. The specimen used in run ET-35

showed the same depth of attack in the weld region
as in the unwelded portion of the tube, with perhaps
somewhat denser subsurface void formations in the
weld. The specimens used in runs ET-36 and
ET-37 both failed in the weld areas after relatively
In all cases the welds were
fabricated by a qualified Inconel welder. However,
since the welds were not critically inspected, the
possibility of initial weld porosity exists. Ad-
ditional tests of this type are to be run with welds
which have satisfactorily passed inspection.

Pulse-Pump System

J. J. Keyes, Jr. A. . Krakoviak
A. G, Smith, Jr.1

The high-frequency pulse-pump thermal-cyclmg
loop has been successfully operated with the fuel
“mixture NaF-ZrF -UF, (56-39-5 mole -%) as- the
fluid medium. The initial run extended over ap-
“proximately 72 hr of intermittent isothermal and
~eyelic temperature-dlfference operation. The mean
fluid temperature was in the range 1300 to 1350°F,
and hot-_to cold- leg temperature differences of as

. ‘much as I75°F were achieved. Much difficulty was
;}‘A}:experlenced however, -in controlling the levels of
“the  six free liquid surfaces incorporated in the
':i;lédp.a At the end 6f the' 'ihifiai run, the loop was

 

 

35, E. Mott and As G Smith, Jr., ANP Quar. Prog.
Re .Dec. 31, 1936, 0RNL-222 p 54~58. For details
of oop, see esp Figs. 1.4.27 anJ 1+4.28.

93

 
v6

   

 

 

 

 

UNCLASSIFIED

T.13299 oo e o1
00T ey | ooz
.00 !
-5- -‘i‘-n -II:I:‘-
2 z 2
.008 008 | oon,
2oL uoer. .00t
008 008 | 008
LRAe | 00® oon
01 Lou N

by T | A A S Y G

{g) STRAIGHT PORTION (&) 90-deg BEND~TENSION SIDE (¢) 90-deg BEND—COMPRESSION SID

RUN ET=3{: NO PRIOR STRESS, FINE GRAINED, FOUR 90-deg BENDS

B UNCLASSIFIED

EREB
EEE

Y

|NC|.'|ES
T

INC';IES

       

 

      

 

 

008 .ooe.

L.oo? oot

008 008

Looe oo

et i 91 , X . o1,
: . / el ,w.: [ 3 AL e, AR . R ’
(d)} STRAIGHT PORTIO () 90-deg BEND-TENSION SIDE (#) 90-deg BEND—-COMPRESSION SIDE

.RUN ET-33: NO PRIOR STRESS, COARSE GRAINED, FOUR 90~deg BENDS

Fig. 1.7.3. Specimens of Inconel Tubing Exposed to NaF-ZeF,-UF, (50-46-4 mole %) ot 1600°F in Thermol Cycling Tests of Bends ot 0.4 cps.
Etchant: modified aqua regia. 250X. (Secret-with-eaption)

‘I . . L]
¥ . ® e + - W . v

1¥0d3Y $S3A90Ud 1I370d¥d INV

 

 
 

 

 

L1

 

cooled to allow for some modifications of the
liquid-level control system.

For the second run there was a goal of 100 hr
of operation at a fluid flow rate of 8 gpm with a
frequency of 1 cps and a temperature differential
of 200°F. Difficulties again occurred with liquid-
level control, and, following another 72 hr of
intermittent operation, the system was shut down
for general repairs and further modifications of the
level controls.

»After installation of a new test section, the
third run completed the 100-hr test peried at a
fluid flow rate of 8 gpm, and the test was dis-

 

PERIOD ENDING SEPTEMBER 30, 1957

continued without incident. The pulse frequency
was 1 cps, the hot-leg temperature was 1500°F,

and the cold-leg temperature was 1220°F. The

test section (Fig. 1.7.4) was fabricated from an
11ein. length of 5/s-in.-()D Inconel tubing with a
&5-milthick wall. The central 9 in. of the test
section -was machined to a wall thickness of
29 mils (11 mil). The two ends of the test section
were left at the original wall thickness to facilitate
welding of the unit into the system. Chromel-
Alume! thermocouples (5-mil wires) were attached
to the outer surface of the tube by resistance-
welding. The total amplitude of the outer wall

-UNCLASSIFIED
PHOTO-29668

Fig. 1.7.4, Typical Test Section After a Run in the High-Frequency Pulse-Pump Thermal-Cycling System.

95

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

temperature oscillation as indicated by these
thermocouples was 127°F at the test section inlet
and 103°F at the outlet. The inner surface temper-
ature oscillation was estimated to be between
+67 and 195°F. This variation exists because of
uncertainties in establishing the boundary con-
dition at the outer surface in the equation describing
the decay of the radial temperature wave. The
metallurgical examination of this test section has
not yet been completed; however, standard dye-stain
and radicgraphic examinations show no superficial
damage. Future tests will again be made at a
cyclic frequency of 1 cps, but the amplitude of the
temperature fluctuation will be varied.

Thérmocouple Development
Jo E- MO"'

An improved technique was used for further
measurements of the surface temperature of a thick-
walled pipe in contact with a fluid whose temper-
ature is a periodic function of time. An Inconel-
nickel ‘‘gunbarrel’’ thermocouple® was shrunk-fit
into the pipe wall, the inside of the pipe was
reamed so that the junction would be flush with the
wall, and a thin (approximately 1-ux) layer of nickel
was deposited by vacuum evaporation. This nickel
layer introduced negligible resistance to heat flow,
and thus enabled instantaneous measurement of
the interface temperature. The thinness of the
nickel layer ensured negligible perturbation of the
fluid flow in the vicinity of the thermocouple.

Typical resuvits for a test section fabricated from
a 0.47-in.-ID, 0.25-in..wall Inconel pipe exposed
to sinusoidal temperature oscillations in water are
shown in Fig. 1.7.5 for a film heat transfer coef-
ficient of 3700 Btu/hr-ft2.°F, As may be seen,
the results are in reasonable agreement with the
Jakob equation# for an infinitely thick plane wall.

An Inconel-nickel ‘‘gunbarrei’’ thermocouple
(not attached to a tube) was tested in static
NaF-ZrF -UF, (50-46-4 mole %) at 1200°F for
300 hr. During this time the thermocouple indicated
a steady temperature of about 1230°F. Metallurgical
examination of the thermocouple, of which o
longitudinal secticn is shown in Fig. 1.7.6, re-
vealed, however, that the metallic junction between
the central nickel wire and the Inconel had been

 

4M, Jakob, Heat Transfer, vol 1, p 298, Wiley, New
York, 1949,

96

UNCLASSIFIED
ORNL-LR-DWG 25899

 

07

 

 

0.5

 

e
THEORETICAL CURVE o~
BY JAKOS h\.\_‘
03 P

 

 

0.2

 

 

 

 

 

 

 

oA
) 2 -4 6 8 {0

FREQUENCY (cps)

Fig. 1.7.5. Ratic of Wall Temperature Amplitude to
Fluid Temperature Amplitude, 7, as a Function of the
Temperature Cycling Frequency for a Thick-Walled
Inconel Tube., Film coefficient: 3700 Bfu/hflfiz-oF.
removed by the fluoride salt.? This is shown more
clearly in the photomicrograph of Fig. 1.7.7. lt is
concluded that the conductivity of the melt was
sufficient to provide electrical continuity. Since
the fluoride salt was found to have penetrated some
distance along the nickel-oxide insulation, it is
evident that this type of thermocouple cannot be
relied upon for long-term accurate measurement of
surface temperatures in a fluoride salt environment.

ART HYDRODYNAMICS
FullsScale Core Studies
W. J. Stelzman

The investigation of auxiliary flow guidance for
improving hydredynamic stability in the ART core
was continved, and two new configurctions were

 

SRe Jo Gray and J. He DeVan, Metallograpbic Ex-
amination of a Gunbarrel Thermocouple, Metallograpby
Specimen No. 16203~Metallography Report No. 294,
ORNL CF-57-6-71 (June 17, 1957).

 
 

 

o

 

 

PERIOD ENDING SEPTEMBER 30, 1957

B UNCLASSIFIED:
" ¥.22959

 

Fig. 1.7.6. Longitudinal Section Through “Gunbarre!’’ Tfiermocouple After 300 hr of Exposure to Static NaF-
ZcF ,-UF, (50-46-4 mole %) ot 1200°F. Oblique lighting causes the nickel wire to appear dark and the insulating gap,
white. (Secret-witireeption) . :

el UNCLASSIFIED
¥-22960.

 

 

 

Fig. 1.7.7. -Photomlcrogmph of End ;f "Gunfiérté-.l"' 'Th-e'l"mldcouplé 'Af;er 300 hr of Exposure to Static NaF-ZrF .
UF, (50-46-4 mole %) at 1200°F. Note the absence of the nickel _pla_te between the end of the central nickel wire
and the Inconel on either side. {Serretwith-eaption)

97

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

tested. One configuration consisted of a pair of
inserts and ramps mounted at each pump discharge
port. The insert, one side of which maintained the
circular contour of the core entrance passage,
while the opposite side provided a gredual area
transition for flow from the pump discharge, was
installed at the junction of the center volute and
pump-discharge passage and occupied a 38-deg
circumferential sector. It extended the full height

of the volute. The ramp was located on the roof -

of the center volute and occupied a 50-deg radial

sector adjacent to the insert. The shapes and the

positions of the inserts and ramps are shown in
Fig- ]07-8'-

For the second configuration, a pair of thin
(approximately l/a-in.) Plexiglas baffles, similar in
contour to the actual core shells, were suspended
in the annular passage between the island and
reflector shells, as described by Platus,$ to form
narrow (approximately 7/a-in.--wide) annuli adjacent
to the island and reflector walls. The baffles
extended approximately three-fourths of the length
of the core.

The first configuration (inserts and ramps) was
designed to increase the tangential component of
pump-discharge flow into the center volute. At
the same time it was intended to provide more
uniform flow distribution circumferentially into
the core annulus by increasing the velocity of
pump discharge ond by providing a downward
velocity component to the swirling fluid mass.
Observation of the flow pattern revealed no major
improvement in stability even when used in con-
junction with the GS-2 guide vanes. However,
additional tests are needed to properly evaluate
this design.

The second configuration (baffles) was proposed
primarily to improve the flow stability adjacent to
the core shells, to absorb some of the thermal
oscillations in the main stream, and to decrease
the peak fluid temperature near the inner and outer
wails. Preliminary work has been concerned with
obtaining proper flow in the three parallel channels
formed by the baffles. Qualitative observations
indicated that satisfactory flow could be achieved
by controlling the area of the inlet passages.
Many more experiments will be required to optimize

the design and to ascertain the effectiveness of
these baffles.

p, H. Platus, ANP Quar. Prog. Rep. March 31, 1957,
ORNL-2274, p 10-12.

 

- 98

Quarter-Scale Core Studies
G. L. Muller?

Velocity and pressure loss data were obtained in
water flow tests of the "’(“-scule model of the
ART 21-in. core with an ART-type entrance header,
a core inlet collimator, and five core screens.
The core and header system are shown in Fig. 1.7.9.
The collimator was fabricated from a 3/8-in.-thick
perforated plate of 0.42 solidity. The screens
were cut from 20 x 20 mesh, commercial, woven-
wire screening. The first four screens had a 0.385
solidity, while the final screen was of 0.510
solidity. , :

The velocity distributions at six axial positions
clong the core are shown in Fig. 1.7.10 for a
mid-plane Reynolds modulus of 6020. These profiles
were cbtained from analysis of photographs of the

. velocity profiles visualized at these positions by

the phosphorescent-particle technique (see Fig.
1.7.13 of the next section). Because of the
entrance collimator, the core flow was nontotational.
The results are in good agreement with previously
reported qualitative data on this system.? The
profiles at positions C, D, and E indicate that
comparatively low velocities exist near the
outer wall in the mid-plane region. A prelimi-
nary examination of velocity data obtained at
NRe,mia = 25,000 for the mid-plane region showed
no major deviations from the profiles of Fig.1.7.10.

The experimental variation of average fluid
velocity with axial position is given by the data
points of Fig. 1.7.11. The positions A°and D’
are located in a vertical plane rotated 90 deg from
the plane containing positions A through F. The
curves shown were calculated on the basis of flow
continuity by using the data at positions A and A’
for base points. The reliability of the experimental
data is indicated by the excellent agreement be-
tween the measured average velocity at position
F and that predicted by the calculated curves. The
average fluid velocities are equal at positions A

and A’ however, a 10% deviation exists at

positions D and D’. This difference lies within
the experimental error at this position.

Thus, the data indicate a core flow which is
peripherally symmetrical despite a large pressure
variation in the core header. The extent of this

 

7G. L. Muller and F. E. Lynch, Effects of Screen
Packing in the ART 2l-<lnch Core and in an RMR Core
Designed to Concentrate the Packing in the Core
Entrance Region, ORNL CF-56-12-5 (Dec. 20, 1956).

 
66

   

 

 

»

 

 

 

 

1S61 ‘0 ¥3GWILJIS ONIANI A0ly3d

 

 
 

 

 

ANP PROJECT PROGRESS REPORT

header pressure variation is presented in Fig. fluid kinetic energy evaluated at the mid-plane,

1.7.12, in which the experimentally measured
relative pressure difference at a number of positions
through the header is plotted as a function of the
mid-plane Reynolds modulus. The relative pressure

difference is the ratio of the pressure difference A schematic drawing of the header and the approxi-
between the test position and the core exit to the mate pressure tap locations are also given. From

WATER IN

_ 284y

2
pvmld )

K

.

 

SLoRTT
ORNL~-LR-DWG 25900

WATER IN

 

 

 

N

 

 

7

 

 

D

 

 

INLET "COLLIMATOR" /i
PERFORATED ~PLATE SCREEN,
SOLIDITY RATIO = 0.420
THICKNESS = 3/g in.

i/
W Ul

 

 

 

 

 

 

NN

/4/ MODEL OF ART-CORE
7 ENTRANCE HEADER AND
) , . PUMP VOLUTES

 

 

 

 

L e

 

 

 

HOLE DIAMETER = 0.0465 in.

 

 

WOVEN-WIRE SCREENS,
SOLIDITY RATIO = 0.385,
20 x 20 MESH —é

WOVEN-WIRE SCREEN,
SOLIDIDITY RATIO = 0.510,
20x 20 MESH — |

11N

 

 

 

 

 

1.500 in

 

 

 

 

&
0.74'}6 in.

 

 

f.491in.

 

2.238n.

 

 

2.982 in.

 

} MODEL OF ART
2{-in. CORE

Fig. 1.7.9. Cross Section of Experimental Mode! of ART Core and Entrance System with Inlet Collimator Plate

and Screen Packing.

100

*

AL}

 
 

 

 

 

‘i )

 

 

| MIDPLANE

 

SCREEN POSITION
(TYPICAL)

 

 

A

NR, mia = 6000 FOR ALL
DATA SHOWN
y= DISTANCE FROM INNER
WALL
b= WIDTH OF CHANNEL

PERIOD ENDING SEPTEMBER 30, 1957

PESRs
ORNL—-LR—DWG 259014

VELOCITY DISTRIBUTIONS AT POINTS SHOWN ON

Ymax

UMAX-

Umax

1.2

0.8

0.4

o

12 —

08

04 —

0

1.2 —

0.8

04 —

1.2 —

0.8

0.4

CORE CROSS-SECTION

A POINT A
Uyax = {4.0 cm /sec
Upy = 13,9 cm /sec

POINT A
Unax=14.9 cm/sec
Upy=13.9 cm/sec

 

0 0.25 050 0.75 1.00
Y/ b

POINT B
Unax=9.91 cm/sec

0.25 050 Q.75 1.00

POINT C
UMAX= 8.19 c¢cm /sec

0.25 0.50 0.75 1.00
_ Y/

..._____[f\

POINT D
Uyax =T7.34 cm/sec
Upy = 5.49 cm/sec

POINT D’
Upax = 7.39 cm/sec
Upy = 6.16 cm /sec

    

0 025 050 075 1.00

POINT E
Upax=7.7T5 cm /sec

- 0.25 050 075 1.00

Q

0.25 0.50 0.75 1.00
/b

POINT F
Upyax=13.0 cm/sec
Upy=12.3 cm/sec

 

Fig. 1.7.10. Radial Velocity Distributions in o Screen-Filled ART Core Model.

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

Fig. 1.7.12, the moximum relative pressure differ-

_ence in the header at the ART design flow rate

(Npy mig = 97,500) is approximately 125. This
corresponds to a pressure drop of 6.06 psi. In

UNCLASSIFIED
ORNL-LR-DWG 25902

w
z
<
-
o
a
=

VELOCITY (cm/sec)

CALCULATED
FROM CONTINUITY
BY USING A AND

 

5 4 3 2 % 0 { 2 3 4 5
DISTANCE FROM MIDPLANE (in.)

Fig. 1.7.11. Axial Variation of Average Velocity in
21+in. ART Core Model, (Setretwitirecption)

comparison, the relative pressure difference across
this core is 160 (7.76 psi).

Data have also been obtained for simulated
single-pump operation in the screen-filled core
model. A preliminary evaluation of the data shows
a relative pressure loss across the core of 220 at
one-half the design flow rate. At this same flow
rate the header relafive pressure difference was
200. No apparent change from the data of Fig. 1.7.10
was noted in either the radial or peripheral velocity
distributions. .These data indicate that, on the
basis of core flow distribution, single-pump oper-
ation with a screen<filled core is feasible.

Further investigations are planned with a 0.33
solidity entrance collimator to determine whether
the core pressure loss can be lowered without
adversely affecting the core velocity distributions.
From the present results, it appears that screens
of higher solidity are needed in at least the two
downstream positions to flatten the mid-plane
velocity distributions. This will be checked in an
additional study.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SEGREw
ORNL-LR-DWG 25903
200 :
STATIC PRESSURE
TAP NO.
RN
180 —
2N
160 _.12.\\\..; 1
1\ ~
8 \ ’
Z 140 2 TOP VIEW
o 1 OF HEADER SHOWING
J 8 APPROXIMATE PRESSURE TAP LOCATIONS
N
L 420 }—
o ; \
w
100 S_
:.l 80 _lg\‘\. 6
> 3
'—
J 60 —g 10
W \
40 k—-..._ 4
\'fil__
20
0
2 x 10? 5 10’
NRe, mid

Fig. 1.7.12. Relative Pressure Difference in Mode! of ART Inlet Header.

102

 
 

 

 

 

C

 

p) .

Instantaneous Velocity Profile Measurements
' F. E. Lynch
Velocity profile photographs were obtained for

flow through the 19/ 14-Scale model of the ART with
an ART-type entrance header, a core inlet colli-

mator, and five core screens. The experimental

system is as described in the previous section
(see Fig. 1.7.9). Photographs were taken at the
six axial positions A through F and at positions A’

~and D’ for twin-pump operation and ot positions A

and B for single-pump operation. Typical velocity
profile photographs are shown in Fig. 1.7.13 at
positions C and E. The lack of sharpness in the
photographs results partially from the fluid mixing
brought about by the screens and partially from
the slow shutter speed (% . sec) of the camera.
This study will be repeated with a 35-mm comera
(faster shutter speed) and a Reynolds modulus
higher than the 6000 to 9000 used for the present
test in order to obtain more sharply defined profile
photographs. Screens of various solidities will
also be investigated. :

In order to obtain quantitative velocity infor-
mation, it is necessary to obtain a photograph of

FERIOD ENDING SEPTEMBER 30, 1957

a grid placed in the plane of the visualized proefile,
as shown in Fig. 1.7.14. The distortion of the
grid lines by the curvature of the Plexiglas shell
can be clearly seen.

"FUSED SALT HE‘AT'TRANSFER
H. W. Hoffman ~ S. L. Cohen
D. P, Gregory

Forced-convection heat transfer. studies with
NaNO,-NaNO,-KNO, (40-7-53 wt %) flowing through
heated tubes I::ave been completed. The data cover
the Reynolds modulus range of 5000 te 25,000 and
show a heat transfer coefficient variation from
800 to 2900 Btu/hr+ft2.°F, The final results are
presented in Fig. 1.7.15 in terms of the heat
transfer parameter, N /N%;‘. For comparison,
earlier data obtained with this nitrate-nitrite
mixture, both in this laboratory® and by other
investigators,? are presented. All the data of

 

- 8H. W, Hoffman, Physical Properties and Heat Transfer
Characteristics of an Alkali Nitrate~Nitrite Salt Mixture,
ORNL CF-55-7-52 (July 21, 1955).

9W. E. Kirst, Wo M. Nagle, and J. B. Castner, Trans.
Am. Inst. Chem. Engrs. 36, 371 (1940).

SECRET
PHOTO 42241

 

 

 

(a) POSITION C

(b) POSITION E

Fig, 1.7.13. Photographs of Visualized Velocity Frofiles in Screen-Filled 10/44-Scale Model of 21«in. ART Core.

103

 
 

 

ANP PROJECT PROGRESS REPORT

S PHO TO 42242

 

Fig. 1.7.14. Photograph of Coordinate Grid Placed in
Plane of Visualized Velocity Profile.

Fig. 1.7.15 have been adjusted by using the value
for the thermal conductivity of this salt that was
recently determined by Powers (see subsequent
section of this chapter on *‘Physical Properties®’).
The experimental results are in good agreement
with the empirical equation, Ny /N34 = 0.023 NQ'S,
which describes forced-convecnon heat transfer
in ducts containing ordinary fluids.!0

Studies have been initiated of forced-convection
heat transfer with KCI-LiCl (41.2-58.8 mole %)
flowing through a type 347 stainless steel tube.
The experimental system is nearly identical to
that used for the NaNO -NaNO,-KNO, (40-7-53 wt %)
investigation. Prehmmary results fall below the
standard correlations, but this discrepancy is
probably due to uncertainty in the data on the
thermal conductivity of this chloride mixture.

 

l‘)Ol'c:llm:n'y fluids are defined as those fluids whose
Prandtl moduli lie in range 0.5 < Np_ < 100.

104

UNCLASSIFIED
ORNL-LR-DWG 25904

04

r

{00
© KIRST, NAGLE, AND CASTNER
® HOFFMAN AND
4 PRESENT RESULTS 5

50

20

N,
N - 0.8
o = 0.023 Ny,
NPr

 

10
5 10* 2 5 10°

REYNOLDS MODULUS, NRe

™
0

10

HEAT TRANSFER PARAMETER, Ay, /W,

Fig. 1.7.15. Hecat Transfer with NaNOz-NcNO -KNO3

_ (40-7-53 wt %),

The possibility of an interfacial film is also being
investigated by using both chemical and x-ray
diffraction techniques. The existence of a film

~ will be further checked in experiments with an
~Inconel tube. |

HEAT TRANSFER EXPERIMENTS
N. D. Green W. R. Gambill
ART-Type Core with Screens

Initial experiments have been completed on the
half-scale volume-heat-source ART core model
containing four 0.342 solidity screens and an inlet
collimator.  Both steady-state and fluctuating
temperatures were measured at the inner and outer
core walls. The transient data are compared in
Fig. 1.7.16 with datc obtained in previous measure-
ments1! of a vaned entrance system. While the
magnitude of the outer wall temperature fluctuations
peak is appreciably above the corresponding value
for the vaned core, the inner core fluctuations

have been reduced by an order of magnitude. In-

Fig. 1.7.17 the corresponding steady-state temper-
ature distributions are presented. The inner wall
temperatures lie consistently below the fluid
mixed-mean temperatures and indicate uniform
high-velocity flow along the inner wall. These
results are in good agreement with predictions made
on the basis of the velocity profiles experimentally
determined by Muller and Lynch? (see also,
Fig. 1.7.10, this chapter). Thus, some hydredynamic
instability in the mid-plane region on the outer wall

 

Ny, F. Poppendiek et ali, Analytical and Experi-
mental Studies of the Temperature Siructure Within the
ART Core, ORNL-2198 (Jan. 31, 1957). ,

fw

 
 

 

 

 UNCLASSIFIED
ORNL-LR -DWG 25905

NRE" 82,000
AT =175°C
W = UNIFORM
P = 0425

OF
~EDDYING JN ¥;-SCALE

VANED

OUTER WALL
INNER

 

o] 2 4 6 8 10 12 14 16 18
AXIAL DISTANCE FROM INLET (in.}

Fig. 1.7.16. Tmnslehf 'Tem.percture Fluctuations in
the Screen-Filled One-Half-Scale Volume-Heat-Source

Model of the ART Core. (‘S'e'eoe*wi-rh‘m‘p'rl'on’)

UNCLASSIFIED
ORKL-LR-DWG 25906

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2.0
SCREEN POSITIONS
€
[+ E =z 2| z
1.8 F-). ] hj el &l /='—-
S8 5 = & TN
S2 48 3 |
D
B g 8 g &
B g F 8
] 0, w |
14 1%
] Nge™ 82,000 (AXIAL)
AT =175°C
1.2 P =0.425mw
i / W = UNIFORM
e /
ST 10 ‘ : :
o
= hE /(~oun-:n WALL 1
0.8 | 1T~
. ‘ vd >
FLUID MIXED-MEAN;/ /
06 / W
/ INNER WALL.
0.4 / /"/
/ ~
0

 

 

 

 

 

 

 

AXIAL DISTANCE FROM INLET (In) = -

Fig. 17.17. Steudy;Sffife Wuli Teini:ieratfires’ in the .
Screen-Filled One-Half-Scule Volume-Heuf-Source Model o

of Ihe ART Core. (Gmfl-wfirtvphm}-

*is'réflected in thek observed peaking of the transient .
' temperature fluctuotlons on _the outer wall of the
~ volume-heat-source “system. - Slmxlurly, the flat.
: -,_velocuty prof:le (approxlmatlng the shape of the
" power profile) near the -inner. wall indicates a low
- surface temperature. - Conversely, a hsgh outer wall
" temperature, as shown in Fig. 1.7.17, was to be

expected.
A possible method of altering the core flow to
increase the flow along the outer wall without

0 2 4 6 B - 10 12 14 16 - 18 -

" MODIFIED COLBURN FACTOR, /
o

PERIOD ENDING SEPTEMBER 30, 1957

materially affecting the inner wall flow lies in
introducing ‘screens of variable radial solidities.
Screens of this type would have high pressure
drops (high solidity) through the central portions
that would force greater peripheral flow. Several
such experimenta! screens have been designed and
are now being fabricated.

Vortex Tube

Preliminary experimental determinations have
been made of the heat transfer coefficients for air
in source vortex flow. The measurements were
made by using metal Hilsch-Ranque tubes having
length-to-diameter ratios of 20. The tubes were
heated by the axial passage of a high-amperage
electrical current through the tube wall. The data
are shown in Fig. 1.7.18, in which a meodified
Colburn factor is plotted as o function of the
energy required per square foot of heat transfer
surfoce. For tubes of identical geometry and the
same over-all pressure drop, heat transfer in
straight turbulent flow and in source vortex flow
can be directly compared. On this basis, the
experimental data of Fig. 1.7.18 show vortex flow
to be a factor of 5 better than straight flow. The
correction to the vortex flow data to account for
the pressure drop associated with the momentum
increase of the air in passing from the tube inlet
to exit has not been included in this analysis.
The magnitude of this correction will be experi-
mentally determined in the 3-in. vortex currently
under construction. This tube will allow radial
and axial traverses of fluid velocity, temperature,

UNCLASSIFIED . -
© ORNL-LR-DWG 25907

100

w
o

N
o

.

™

 

04 02 05 4 2 5 40 20 50 400
ENERGY REQUIRED PER UNIT AREA (hp/f%)

Fig. 1.7.18. Experimental Heat Transfer to Air in
Source-Yortex Flow ond in Straight, Turbulent Flow
Through Pipes.

105

 
 

 

 

ANP PROJECT PROGRESS REPORT

and pressure. Optimization of the heat transfer
efficiency as related to such factors as tube
length-to-diameter ratio and entrance region design
is being investigated.

LIQUID-METAL VOLUME-HEAT.SOURCE
' EXPERIMENT
G. L. Muller

An investigation of heat transfer in a liquid

metal (mercury)system with internal heat generation

is in progress. This study is aimed at a better
understanding of the heat transfer mechanism so
that more accurate calculation of the thermal
structures in circulating-fuel reactors can be made.
Discrepancies between results from experimental
fiquid-metal forced-circulation heat transfer studies
and theoretical calculations have raised questions
as to the validity of corresponding analyses for
the volume-heat-source case. In particular, the
“*interface thermal resistance’’ theory postulated
to explain the reduced values of the experimentally
measured Nusselt modulus when heat is transferred
through the tube wall will be tested in this study.

A diagram of the experimental system now being
constructed is shown in Fig. 1.7.19. Mercury wil!
flow from the sump into the Moyno pump, through a
flow=measuring orifice, through the test section
and ccoler, and then return to the sump. In an
additional flow path, the mercury will bypass the
test section and cooler.

The test section is a l(“-in. glass tube interrupted
by three electrodes for volume electrical heating.
The arrangement is such that no heat is generated
outside the test section. The wall temperatures
of the test section will be measured by copper-
constantan thermocouples attached to the outside

of the glass tube. To prevent heat loss through
the test section wall, external insulation and guard
heating will be provided. Fluid mixed-mean inlet
and exit temperatures will also be measured. In
this apparatus Reynolds moduli of 500,000 can be
attained. A second test section of 1%-in.-ID glass
tubing will be available with which to make radial
temperature and velocity traverses.

MASS TRANSFER
J. J. Keyes, Jr.

A comparison with experimental results12.13
of the amount of mass transfer to be expected in
alkali metal-alloy systems was made for two
assumed diffusion mechanisms.'4  The results
are presented in Table 1.7.3. In the first mecha-
nism, the boundary layer was assumed to be satu-
rated and the rate of transfer was limited by the
rate of diffusion of solute into the liquid. A general
expression was derived that included effects of
variation in temperature and liquid composition
around a closed loop and which simplified under
certain conditions to

W(O) = k, Sty — t.)0,

where W(@) is the amount of mass transfer in time
0, k; is the conventional mass transfer coefficient,

 

125, H, DeVan and J. B. West, A Brie[ Review of
Thermal Gradient Mass Transfer in Sodium and NaK
Systems, ORNL CF+57-2-146 (Feb. 11, 1957).

13), H. DeVan and R. S. Crouse, ANP Quar. Prog.
Rep. Dec. 31, 1956, ORNL2221, p 175; ANP Quar.
Prog. Rep. March 31, 1957, ORNL-2274, p 154.

145, 3, Keyes, Jr., Some Calculations of Diffusion
Controlled Thermal Gradient Mass Transfer, ORNL.
CF-57-7-115 (July 22, 1957).

Table 1.7.3. Summary of Calculations and Comparison with Data for Two Postulated Mechanisms

of Mass Transfer in Sodium=inconel Systems

Exposure time: 1000 he
Temperature differential: 300°F

 

Mass Transfer Estimate, W (g)

 

Hot-Zone Temperature

Experimental Mass Transfer

 

(°F) LiquideDiffusion Limiting Wall-Diffusion Limiting (g) from Forced=Circulation Lo'ops- '
Method - Methed R :
1500 400-1000 2.-25 13
1-8 <0.5

1200 300-~700

 

106

{ »y

o

 
 

 

4

 

     
     

FLOW CONTROL VALVE

 

 

BYPASS -

MERCURY
FLOW DIRECTION

zone = temperatures, respectively.
forced-crrculahon loop conditions (tH ="1500°F,

ture dependent, in contradiction to observation, .

In the second mechanism, the rate of mass

transfer was assumed to be limited by the rate of

     
  

SECTION

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR—DWG 25908

MIXING CHAMBER AND
THERMOCOUPLE WELL

i

POWER j
TRANSFORMER

 

 

 

GLASS
TEST

\MQMQJ‘—I

 

 

 

 

 

 

\\SATURABLE-CO RE
REACTOR

\THIN-PLATE ORIFICE

N
o |

PUMP DRIVE MOTOR AND
SPEED SELECTOR

POSITIVE DISPLACEMENT MQYNO PUMP

Flg. 1.7. 19. . Schematic Diugrom of Voluma-Heaf-Source Experimenful Appcrurus in Which Mercury will be Circu-
iufed. L : :

S is the solublltty of the soiute (Nl) in the !:quld'i?'
(No or NaK), and -2, and tC are the hot- and ‘cold- -
For- typical

diffusion of a component (for example, Ni) of the

-solid alloy (for -example, Inconel) to the solid-
. liquid interface.
derwed for W(G) was
. =1200°F, 0 = 1000 hr), this mechanism. pred:cts ,
to be between 400 ond 1000 g, whereas: the‘-__-,
observed value is about 13 g.- Furthermore, this
mechanism predicts W to be only slightly tempera-

- The approximate expression

1/2 '
o i D6 x(ty = te)
: W(@) = 2pw ——77 ' —-———-—t ’

where D is the diffusivity of nicke! in the solid
wall, x is the bulk solid concentration, and p , is
the solid density. -

107

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Based on extrapolation of some data for the
diffusivity of nickel, as reported in the literature,
W(@) was estimated to be between 2 and 25 g for
ty = 1500°F, ¢, = 1200°F, 6 = 1000 hr. As may
be seen, this estimate is in better agreement with
the observed mass transfer accumulation than was
the estimate based on liquid diffusion. On the other
hand, the wall diffusion mechanism does not
correctly account for the dependence of the mass
transfer (W) on flow rate and time and does not
predict a sharp break in mass transfer at cbout
1350°F. It is concluded that ¢ more general
hypothesis is needed for combining the mechanism
of hot-zone attack with o nucleation-deposition
mechanism in the cold zone.

PHYSICAL PROPERTIES
W.D. Powers
Enthalpy and Heat Capacity

Studies were made to determine the enthalpies
and heat capacities of the eutectic mixtures LiCl-
SrCl, (22.5-77.5 wt %) and LiCl-BaCl, (32.2-67.8
wt %). Preliminary analyses of the dzata yielded
values for the heat capacities of 0.21 Btu/1b.°F
for the barium-containing mixture and 0.24 Btu/1b.°F
for the strontium-lithium salt. The reliability of
these data is in some doubt, since wide dis-
crepancies in enthalpy were observed between
duplicate test samples. This may be due to the
large density differences between the salts.
Further studies with this pair of chloride eutectics
will be made as soon as new samples can be
prepared.

Thermal Conductivity

Modification of the variable-gap thermal conduc-
tivity device to include a large guard heating ring
surrounding the heat meter has been completed.
This guard heater ensures lower heat losses from
the heat meter and at the same time provides a
heat flow path of cross section more nearly equal
to the cross-sectional area of the sample heater
ond its guard.

The results obtained with the salt mixture
NaNO,-NaNO,-KNO, (40-7-53 wt %) show increased
consistency ioth within runs and between runs.
The results of a number of typical experimental
runs are shown in Fig. 1.7.20. In this plot, the
slope of the data line is the reciprocal of the
thermal conductivity of the liquid sample. The

108

thermal conductivity values obtained in this study
of the nitrate-nitrite salt are given below:

Temperature Thermal Conductivity
(°F) (Btu/he-#1°F)
689 0.35
667 0.34
487 0.34
468 0.33

All temperature effects were within experimental
error.  The results compare favorably with data
obtsined by Deem!'5 (0.33 Btu/hr-ft-°F between
400 and 900°F) at Battelle Memorial Institute.

UNCLASSIFIED
ORNL-LR-DWG 25909
06

as A~
x |

 

 

 

 

0.4
L]
S
a
203
< e 353°%
o 255°C
02 o 242°C

 

 

7
o /
F

Q 0050 0.400 0.150
GAP THICKNESS (in.)

 

 

 

 

 

 

0.200

Fig. 1.7,20, Experimental Data Obtained in Measure-
ments of the Thermal Cenductivity of NcNOz-NcNO3-
KNO 4 (40-7-53 wt %)

The thermal conductivity of the fuel mixture
NaF+Z¢F UF, (50-46-4 mole %) has been de-
termined by several laboratories. The results are
compared in Table 1.7.4. No trend with tempera-
ature was noted. This salt mixture will be re-
investigated with the use of the modified variable-
gap device. The variation between the Mound
Laboratory and the ORNL and BM! data has not

yet been resolved.

 

154, w. Deem, unpublished data.

 
 

 

Properties of Zirconium Fluoride Vapor Deposits

The density of a sample of zirconium fluoride
powder collected in a "*snow’’ trap was measured
and found to vary between 0.55 and 0.8 g/cm3.
The samples were collected by pushing a thine
walled, sharp-edged tube through the ZrF, and
into contact with the metal surface to which the
powder was adherings. The thickness of the
sample, whose area was defined by the inside
cross-sectional area of the tube, was determined

by averaging the depth ‘indicated by a fine-wire

probe inserted at four positions immediately

adjacent to the sample. The sample thicknesses.

varied from 0.05 to 0.4 in. It was found, for sample
thicknesses greater than 0.1 in., that the density
varied inversely with the sample thickness, as
shown in Fig. 1.7.21. Tests have also been
initiaoted to determine the effect of subsequent
heating on the apparent density of the ZrF , deposit.

An attempt was also made to measure the thermal
conductivity of the ZrF , deposit. A sample of the
material was ploced in o cylindrical annulus
surrounding a central heated tube. The conductivity
of the material around the *‘hot wire’’ is determined
from the temperature rise of the electrically heated
wire. A value of 0.45 Btu/hr-ft-°F was obtained;

PERIOD ENDING SEPTEMBER 30, 1957

however, this measurement was made on a
‘*disturbed’’ sample, and it is planned to mount @
tube in @ snow trap and allow the ZrF , to collect
on the tube.

UNCLASSIFIED
CRNL—-LR-DWG 25910

 

 

 

 

 

 

 

 

 

 

 

1.0
P
'h-.--..__---. e
— ® ., @
" ® '--.__. ®
£ . .--—.'-.
'gx ® ™ -
= e
E 0.5
w
Z
w
O
0
0 0.1 0.2 0.3 0.4 0.5

SAMPLE THICKNESS (in.}

Fig. 1.7.21. Experimental Measurements of Apparent
Density of Zirconium Fluoride Vapor Deposits. (Gom
frckerrtirtwrith sien)

Table 1.7.4. Thermal Conductivity of Melten NuF-ZrF4-UF 4 (50-46-4 mole %)

 

Investigator Experimental System

Thermal Conductivity

 

(Btu/ hr+ft°F)
Powers, ORNL Variable gap 142=1.5
Fixed gap 0:8~146
Deem at BMI* Variable gap 1.2
- Jordan at Mqunfl Laboratory** Calériméfr'ic device | 049

 

*H. W, 'Deém, unpublished datas _

** Aircraft Prbpulsz'bn Reaétéf,s_, ~‘Mound "Lc‘b'o'ra'féry' Memorandum Report No. 57-7-40, p 5 (Aug. 5, 1957).

109

 
 

 

 

 

 
 

 

 

 

»

Part 2
CHEMISTRY
W. R. Grimes

 
 

 

 

 

 

#
A

 

 

 

2.1. PHASE EQUILIBRIUM STUDIES'

C. J. Barton

THE SYSTEM KF-UF
H. A. Friedman

Further equilibrium studies of the system KF-UF
have confirmed the liquidus curve presented
previously.? Results of thermal-gradient quenching
experiments have shown, however, several differ-
ences from the phase behavior proposed earlier on
the basis of thermal analysis and optical and x-
ray-diffraction examinations of slowly cooled melts.
Of the eight KF-UF , phases reported by Zacharia-
sen,® two phases, B-2KF-UF, and KF.2UF,,

have been shown to be equilibrium phases in the

system. |t has not been definitely determined
whether 2KF.UF, displays distinct temperature
ranges for the ordered (B) and disordered (8°)
forms. Further, no evidence has been found that
the phases designated by Zachariasen as
o-3KF.UF,, B-3KF.UF,, o-2KF.UF,, KF.UF,
(which occurs, rather, as 7KF «6UF ), KF- 3UF4,
and KF.6UF, occur in slowly cooled melts or in
quenched samples purified by hydrofluorination and
subsequently protected against exposure to the
atmosphere. 1f, however, such samples are pre-
pared exactly as described by Zachariasen,* that
is, exposed to air during heating and cooling, the
phases designated an3KF-UF4, B-3KF'-UF4, and
KF.6UF, can be formed in large amounts. No
phase that fits the description of the compound
KF.:3UF, has been found under either type of
exper:mental condition. In thermal-gradient
quenching experiments the phuses a and
B-3KF.UF, are apparently stable if initially
present in 1he samples. Protected purified melts of
the composition 3KF.UF, show a biaxial-negative
blue-green phase upon solld:ftcahon, and thermal-

gradient  quenching experiments and - “thermal

analysis indicate that this is the only equilibrium

phase for the compound 3KF UF4. No snmllar

 

]The pefrogmphlc ‘examinations _re orted here were
performed by T. N. McYay, Consultant, and R. A.

Strehlow, Chemistry Division. The x-ray examinations
were performed by Ri W, Thoma, Chemistry Division.

H. A. Frledmcn, ANP Quar. Prog. Rep. Dec. 10, 1955,
ORNL-2012, p 78 and Fig. 4.4, p 82.

3w. H. Zachoriasen, J. Am. Chem, Soc. 70, 2147-2151
(1948).

4W. H. Zachariasen, Crystal Structure Studies at the
Systems KF-UF4, KF-TbF4. and KFe-LaF,, CC-3426
(Feb. 9, 1946).

R. E. Moore

R. E. Thoma

phase has been reported by Zachariasen. The
phases a- and B-3KF.UF, and KF-6UF, may be
either metastable phases in the system KF-UF,
that are stable against inversion to equrhbnum
forms or oxygen-containing phases.

Zachariasen has reported a face-centered cubic
form (a) and two hexagonal forms (8, end B,) for
the compound 2KF-UF ,. If the alpha form were to
display phase behavior similor to that of the
isomorphous phase® reported to be a-2NaF-UF 4 it
would be observed as a primary phase at fhe
35 to 40 mole % UF, region, and its formula would
be 5KF-3UF,. No such primary phase for a cubic
material has been observed. The possibility of a
subsolidus existence for the cubic phase is not
precluded, although the phase has not been ob-
served within the 50°C temperature range below the
solidus. No observations of 2KF-UF , samples have
yet permitted a distinction of the ordered and dis-
ordered hexagonal forms. No well-crystallized
samples of 2KF:UF , have become available. The
compound 2KF.UF, has a lower limit of stability
of 620°C, at which temperature it decomposes into
3KF«UF, and 7KF-6UF,. Some unexplained low-
temperature thermal effects in the 2KF-UF, com-
position region require that this decomposition
temperature be considered to be tentative,

The equilibrium phases in the system KF-UF, and
their melting choracteristics are given in Table

2.1.1.

 

5This phase has been identified as 5KF'3UF4.

Table 2,1.1. Equilibrium Phases in the System KF-UF‘

 

 

Melting Melti
Formula Point ch e*tu-\g'.
©0) aracteristics

3KF.UF, 955 Melts congruently
2I(F'_UF4 _ 760 Melts incongruently to

3KF-UF and liquid;

decomposes at 620°C
7|‘(F'6UF4 790 Melts congruemly
KF-ZUF4 765 Melts incongruently to

UF ; and liquid

 

“()Ay 13

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

THE SYSTEM NuF-HfF‘
H. A. Friedman

Samples of pure, sublimed HfF , became available
during the quarter that were pure enough to per-
mit refined phase equilibrium studies of the system
NaF-HfF . Thermal analysis data obtained with
the pure materwl are in substantial agreement with
data obtained in thermal-gradient quenching experi~
ments, and thus some of the phase equilibrium
results based on previously reported® . thermal
analysis data are presently in doubt. Precision in
‘the determination of liquidus and solidus temper-
atures and temperatures of polymorphic tronsitions
has been improved to +3°C, in several cases, by
using sublimed HfF,. A tentative phase equi-
librium diagrom of the system NaF-HfF , is shown
in Fig. 2.1.1. Areas outlined by dashed curves are
not yet well-established.

UNCLASSIFIED
ORNL-LR-DWG 25916

TEMPERATURE (°C)

3NaF -HfF,
5NaF - 2HfF;
2NaF - 6HfF,

-HIF,
7NaF - 6HF,

 

NaF 10 20 30 40 50 60 70 80 90 Hfl';',
HfF, (mole %)

Fig. 21.1. The System NaF-HfF (Tentative).

Seven NaF-HfF, phases are isostructural with
NaF-ZrF, phases of corresponding stoichiometric
formulas. The analogous series of compounds in
the two systems are not continuous, however, s is
shown in Table 2.1,2; there are fewer phases in
the NaF-HfF , system then in the NaF-ZrF system.
The appeorcmce of the phase NaF- HfF4 as a
congruently melting stable compound is the only

 

‘8B, S. Landau, H. A. Friedman, and R. E. Thoma,

_ANP Quar, Prog. Rep. March 31, 1957, ORNL-2274, p 93.

114

marked deviation of phase béhavior foran NaF-HfF
analog of an NaF-ZrF , phase.

THE SYSTEM KBF ,-NaBF,
R. E. Moore

A quenching study of the system KBF (NaBF
was undertaken during the past quarter as part of cl
search for low-melting salt mixtures for use as
reactor coolants, Thermal data obtained pre-
viously?¢® defined a liquidus curve with a mini-
mum at about 90 mole % NaBF, and 360°C, but
there were no thermal effects on the cooling
curves that corresponded to solidus temperatures,
Thermal effects that presymably represented solid
transitions were found. in the range 180 to 280°C.

~Petrographic examinations of quenched samples
gave liquidus values that were in good agreement
with the thermal data. The liquidus temperatures
at 50, 70, and 75 mole % NaBF , were 441, 392,
and 387°C, respectively, with KBF, as the pri-
mary phase; the solidus temperatures were 356,
360, aond 355°C, respectively. The secondary
phase was identified as NaBF in each case. At
80 mole % NaBF , the primary phuse is KBF,, and
the solidus temperoture is 358°C. The composition
containing 90 mole % NaBF, is very near the
eutectic.  The primary phase is NaBF,; the
liquidus temperature is 355°C; and the solidus
temperature is 351°C.

X-ray diffraction and petrographic examinations
of previously equilibrated quenched samples
provided no indication of the existence of either
compounds or solid solutions in the system. |t is
a simple binary system with a eutectic near 90
mole % NaBF , and about 375°C.

YTTRIUM FLUORIDE SYSTEMS
R. J. Sheil

Petrographic and x-ray-diffraction examinations
of slowly cooled samples of LiF-MgF,-YF, and
LiF-ZnF o~YF, mixtures, which are of mferest
in the investigation of the production of oxygen-
free yttrium metal,? have revealed the same un-
identified compound in both systems, This sug-
gests the possibility of an LiF-YF; compound,

 

7). G. Surak, R. E. Moore, and C. J. Barton, ANP
Quar. Prog. Rep. Sept. 10, 1952, ORNL-1375, p 82.

J. G. Surak, ANP Quar. Prog. Rep. Dec. 10, 1952,
ORNL-1439, p i10.

9
Rc Jo Shel'l ANP 2are Pfogn Re . ure 30' 1957]
ORNL-2340, p 128. : e pe

"

 
Gl

 

 

 

 

 

 

observed

( % s v » v * " 0 ! ®
Table 2,1,2, Properties of Analogous Phases in the SYstef'ns NuF-Z_rF4 and Nc::F-HfF4
NGF-ZI'F4 ‘ NoF-HlfF“
‘ Phase Change | Phase Change
Phus_e'.. - Temperature Typ °. of : Phusel " Temperature TTVPQ_ ?f
Formula (°C_)- Transition Formula (°C) ransition
3N¢:F-ZrF4 o 850 | Congruent melting point 3NaF-HfF4' _ 855 Congruent melting point
SNuFQZrF4 : 639 . |nc§ng‘ruent melting point SNuF-ZIHl‘F4 606 Incongruent melting point
SNOF'Z.Zl;F4‘ : 525 'Iriver;ipn of a-SNoF-ZZrF"‘ 5N¢:F-2HI"F4 533 Inversion of G-SNoF-ZHfF4
R to 5‘5““’:'22":4 | to fiSNc:iF'ZHI‘F4
2NuF-IZrF.4' 544 ' lncdngruent melting point 2NaF - HfF, 593 Incongruent melting point
- of 3,-2NaF ZrF , of 3,-2NaF -HfF
2Na F'ZrF4 533 Inversion of B2-2NaF'ZrF4 ' No B2-2N0F-HfF4
. ‘ to fi3-2N-aF-ZrF 4 | observed
2NaFZrF 505 Inversion of B3-2NaF ZeF 2 NoF *HfF 520 + 15 Inversion of fi3-2NoF-HfF4
: to [3“-2Nch-ZrF4 to 54'2N0F'H”:4
3Nch-22rF4 487 Upper stability limit of No 3NoF2HfF
‘ ' 3NaF«2Zr F4 observed '
7NflF°GZrF4 525 Congruent melting point 7NaF'6HfF4 515 10 Incongruent melting point
Nch-ZrF4 Metastable phase formation NoF-HfF4 545 Congruent melting point
3N<IF-4ZI'F4 ‘ . 537 Incongruent melting point No 3NaF«4HfF ,

 

£561 ‘08 439W3Ld3S ONIANT Qoly3d

 

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

although published daota'® indicated that no com-
pounds are formed in this system. Two LiF-YF,
mixtures were prepared in the thermal apparatus
that permits visual cbservations. The mixtures con-
tained 20 and 25 mole % YF,, respectively, and
both were found to be composed of the above-
mentioned unidentified compound and LiF. Both
mixtures showed a solidus temperature of about
690°C and liquidus temperatures approximately
10 to 14°C above the solidus temperature; these
liquidus temperatures are much lower thon the
values reported by Dergunov.'®  Further study
will be required to identify the LiF-YF; compound.

An equimolar mixture of YF, and MgF,, which

116

had a liquidus temperature of 1025°C and a solidus

temperature of 980°C, was found to contain only

the two crystalline starting materials. Another
mixture in this system that contained 75 mole %
YF; did not appear to be completely liquid ot
1200°C, the maximum temperature to which it was
heated. The ZnF,-YF, mixture (80.6 wt % YF,),
for which thermal data were previously presented,?®
contained only the starting materials, |t seems
certain that no compound is formed in this binary
system, |

 

10 p, Dergunov, Doklady Akad, Nauk SSSR 60,
1186 (1948).

Y

¢

 
 

PERIOD ENDING SEPTEMBER 30, 1957

2,2, CHEMICAL REACTIONS IN MOLTEN SALTS

F. F. Blankenship
R. F. Newton

EQUILIBRIUM REDUCTION OF NiF, BY H,
| IN NaF-ZeF,

C. M. Blood
Investigation of the equilibrium
NiF )(d) + H,(e)<=—Ni(s) + 2HF(g)

in the reaction medium Nc:l"'-Zl'F4 (53-47 mole %)
at 600°C was completed, and the data obtained are
summarized in Table 2.2.1. These results are
compared with those obtained previously! at 550,
575, and 625°C in Fig. 2.2.1 of the following

section.

L. G. Qverholser
G. M. Watson

ACTIVITY COEFFICIENTS OF NiF, IN MOLTEN
NdF-ZrF4 SOLUTIONS

C. M. Bloed

The determination of the activity coefficients
of NiF, in the molten mixture NaF-ZrF, (53-47
mole %) by the study of the equilibrium constants
for the reaction

NiF 5(d) + H,(g) —Ni(s) + 2HF(g)

1C. M. Blood, ANP Quar. Prog. Rep. Dec. 31, 1956,
ORNL-2221, p 120; ANP Quar. Prog. Rep. March 31,
1957, ORNL-2274, p 102; ANP Quar. Prog. Rep. June
30, 1957: 0RNL‘2340: P 131-

 

Table 2.2.1. Egullibrium Ratios for the Reaction NiF, (<) + Hz(g)fi Ni(s) + 2HF(g)
at 600°C in NaF-ZrF, (53-47 Mole %)

 

 

Ni in Melt - Pressure of H2 Pressure of HF .

(ppm) (atm) (atm) x

x 104
308 0.0292 0,489 1.56
310 0,0299 0.478 1.45
330 0.0296 0.482 1,40
313 0.0299 0.478 1.43
150 0.0358 0.375 1.54
137 0.0357 0.376 1.70
140 0,0360 0,371 1.60
143 0,0365 0.362 1.47
240 0.0316 0.447 1.55
325 0.0285 0.501 | 1.59
420 | 0.0263 0,540 1455
445 | 0.0253 0.557 | 1.62
430 | 0.0260 0.545 1.56
440 ' 040262 0,542 1.50
425 | 0.0271 0.527 1,42
445 . 0.0262 0.542 1448

AV 1053 i 0.07

 

 

 

 

*K, = Pl?iF/XNIszHz' wherg X is mole fraction and P is pressure in atmospheres,

117

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

has been brought to a successful conclusion. In
order to determine the effect of the solvent on the
behavior of NiF,, it is pertinent to compare the
experimental values of the equilibrium quotients
with corresponding values of equilibrium constants
obtained by extrapolation of the experimental data
to infinite dilution and by calculations based on
the vclues, available in the literature,? of the
free energy of formation of the solid NiF, and
gaseous HF, _

The values obtained at 550, 575, 600, and 625°C
for K_ for the reaction of NiF, with hydrogen are
plotted in Fig. 2.2.1 as a function of concentration
of nickel in the melt. The values of K, are,
within the accuracy of the experiments, inde-
pendent of the direction from which equilibrium
was approached and independent of the concen-
tration of Nil"'2 in the solution. Furthermore, as
may be observed, the extrapolation of the data to
infinite dilution presents no difficulty.

 

21, Brewer et al., Natl. Nuclear Energy Ser. Div. IV
198, 65~110 (1950).

The numerical value of the activity coefficient
of NiF , depends on the choice of standard state of
that compound. Calculations have been made and
values are tabulated for each of the following three
standard states: (1) solid crystalline NiF,,
designated NiF ,(s); (2) a hypothetical liquid
supercooled to the reaction temperature, following
a suggestion by Hildebrand and Scott,3 designated
NiF,(I); and (3) NiF, dissolved ot unit mole
fraction [designated NiIg o(@)] and assumed to have
the same activity coefficient as at infinite dilution,
that is, an activity coefficient of 1.

Values for the free energies of formation of
HF(g) and the standard states NiF,(s), NiF,(Z),
and NiF ,(d) are given in Table 2.2.2. The values
for NiF ,(s) and for HF(g) were taken from the
publication of Brewer et al.2 In order to compute
AF® for NiF ,(1), the melting point of 1300°K and
a temperature-independent heat of fusion of 8000

 

3), H. Hildebrand and R. L. Scott, The Solubilit
c]% lgonelectrolytes, 3d ed., p 12, Reinhold, New York,
50. |

UNCL ASSIFIED
ORNL-LR-DWG 25917

 

{x10%

625°C <
.

 

/A

 

K, {atm)

 

 

 

 

 

 

. [ == R R A A

 

 

 

 

 

 

 

 

 

0 50 100 150 200

300 350 400 450 500

NICKEL IN MELT {ppm)

Fig. 22,1, Equilibrium Quotients for the Reduction of NiF, by H, in NaF-ZrF , (53-47 mole %).

118

 
 

 

 

 

Table 2:.2.2. Free Energies of Formation
of HF and Nin

 

 

 

Temp?rature AF° (keal)
(°c) HF(g) NIiF,(s) NiF,(l) NiF,(d)
625 —6570 —127.14 =—124.65 ~l’!3g56
600  ~65.67 —127.88 —12525 ~11462
575,  ~6564 —128.64 —12584 —115.60
550 ~65.61 —129.38 —126:46 —116460

 

cal/mole, also given by Brewer et al,, were used.
The values of AF® for NiF,(d) were obtained by
extrapolation of the K, data to infinite dilution
to establish K_ for the reaction in which NiF ,(d)
is the standard state. The free-energy changes of
the reduction reaction were then calculated and
combined with the corresponding AF° for HF(g)
to yield the tabulated results. _

The values calculated for the equilibrivm
constants, K_, for the chemical reactions, along
with the activity coefficients of NiF, for each of
the three standard states, are shown in Table 2.2.3.
A  comparison of the calculated equilibrium

PERIOD ENDING SEFTEMBER 30, 1957

constants obtained for NiF _(s) as the standard
state and the experimental vu:iues, as presented in
Fig. 2.2.2, reveals a striking discrepancy between
the calculated and the experimental values. The
discrepancy is directly reflected in the very large
values of the activity coefficients obtained when
the solid and the supercooled liquid are chosen
as standard states for the NiF,. If NiF,(d)is the
standard state, however, the activity coefficient is

 unity as a consequence of the constancy of the

experimental equilibrium quotients obtained. This
is equivalent to stating that NiF, obeys Henry's
law over the concentration range studied, which
includes, ot 550°C, values not far from the concen-
tration of NiF, in the saturated solution.

The temperature dependence of the experimental
K, values and of the calculated K values for the
reactions involving NiF,(s) and NiF () is shown
in Fig. 2.2.3. The heats of reaction calculated
from the slopes of these lines and the corre-
sponding entropy changes are summarized in
Table 2.2.4.

The heat of reaction calculated from the temper-
ature dependence of the experimental values is in
substantial agreement with the corresponding
values calculated for the reduction reactions of

NiF ,(s) and of NiF,(I). On the other hand, the

Table 2.2.3. Equilibrium Constants for the Reduction of NI!"'2 by Hydrogen Gas and Activity Coefficients
of NiF2 in NaF-ZrF, (53-47 Mole %)

Reactions:* (1) Ni_Fz(s) + H,(g) = Ni(s) + 2HF {g)
(2) Nin(I) + Hy(g) = Ni(s) + 2HF(g)
(3) NiFy(d) + Hylg) = Ni(s) + 2HF(e)

 

.. Equilibrium Constant, 'Ka'

Activity Coefficient** for Nin

 

 

 

Temperature
| (°¢)' Reaction 1 R-n_cflcn 2 | _Reaction 3 W) y(l) y(d)
625 10.90 a8 21,90 * 2,000 2009 500 1
- 600 o 737 336 15,306 + 700 2076 455 1
Csis T am0 252 0 M,000 £ 600 2292 436 1
550 30 184 7,600 + 400 2468 413 1

 

*The notatlons s, g, 1, and d ‘rle‘fe;r to crystal-ling' 's_pl id, gaseous, supercooled liquid, and dissolved states,

respectivelys -

*%The ccf;viw"coefiicienfs yis) and"-y(l) were calculated byusing crystalline selid end supercooled liquid, respec-
tively, as the standard stotes for NIF2 in the molten solvent; y(d) is the assumed activity coefficient for NiF2 dis

solved at unit mole fraction as the standard state.

119

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL~LR—-DWG 25918

EXPERIMENTAL K, .

 

 

E
c
X %
CALCULATED 4, FOR NiF, (s) AS THE STANDARD STATE
625°C
600°C c
575°C
550°C
=
75 150 250 350 435
NICKEL IN MELT (ppm)
Fig. 22,2, Experimental and Calculated Equilibrium Constants for the Reduction f.d’,Nfl'-'2 by H2 in Nch-Zl-F4 ( ]

(53-47 mole %).

120
 

 

 

entropy change calculated from the experimental 7

measurements reflects the vast ‘discrepancy  be-
tween the equilibrium constants obtained by using
the different standard stotes. -

It is clear from the results of this. |nveshgaf|on
that NiF, in solution,” under the experimental
condltlons used, behaves in a vastly different
manner than that predicted by the free energies of

formation for solid NiF . This is rather surprising -

UNCLASSIFIED
ORNL-LR-DWG 25049 -

TEMPERATURE (%) 7
800 - 700 625 600 575 550

=T T T 113"

 

L]

X (atm)

 

 

 

 

. : T__.qof/r("d ..: Lo

F!g. 2.2.3. Tempercture Dependence ol’ Nin Re-

ducflon Equlllbrlu.

PERIOD ENDING SEPTEMBER 30, 1957

in view of the agreement previously found4 to

~exist “between the experimental and calculated

behavier of FeF, in the same solvent. The
striking dlfferenc:es in behavior between NiF, and
FeF, can be appreciated by comparing the achv:ty

" coefficients at 600°C. The values are listed in

Tdb|e 2‘2 050

. Table 2.2.5. Activity Coefficients of NiF, and of FeF,

in Molten NaF-ZrF, (53-47 Mole %) at 600°C

 

Standard State  yp o yNIF2 yNin/yFon

 

2
MF(s) 3,28 2076 633
MF () 0:666 455 689
MF,(d) 1 1 1

 

~ K {at)

The values presented in Table 2.2.5 are these
to be used for relating the behavior of FeF, and
NiF, in the NaF-ZrF, solvent to the thermo-
dynomlc estimates founj in the literature.2 The
experimentally determined activity coefficients for
NiF, are high enough to throw doubt on the validity

- of the estimates in the literature. A further

discussion of this point is presented in the
following section.

THERMODYNAMIC CONSIDERATIONS RELATING
TO THE ACTIVITY OF NiF,

‘M. Blander F. F. Blankenship
The activity coefficients Found for NiF, in fuel

' _mlxtures are quite |arge, as. reported in the previous
i 'sectlon.  Based on the standard free energy of

 

4C. M. Blood. cnd G. M. Watson, ANP Quar. Prog.
Rep. March 10, 1956, ORNL-zom, P 84s

Tnble 2.2.4. Heuts of Reuctlon und Enirepy Chunges for the Reducflon of NiF by Hydrogen

 

Stgn(!a\f.c.l_is.“l_'ef‘e ) . a

- -';Heafs' of -Recction,

Enh'opy Changes,
ASO (cu|/°K per mole of NIF2)

 

Nle(s) Lo w0 25,0000 L

MR, T T T T 00
NIF (d) | . 20,500

o '-fAH":-(car;pe._r‘- mole of NIF,)

226
2645

42.7

 

121

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

formation of solid NiF, (ref 2), the activity coef-
ficient of NiF, in NaF-ZrF, (53-47 mole %), as
determined by equilibrium measurements made with
H:2 as a reducing agent, is 2 x 10° at 600°C, ond,
based on o comparison of NiF, with FeF, by
means of emf measurements on Ni and Fe electrodes
in the NaF-ZrF, (53-47 mole %) solvent, the
activity coefficient of NiF ., at 650°C is 1.8 x 103
(see subsequent section o? this chapter on *‘EMF
Measurements in Molten Salts”). | .
The agreement between the results from the emf
studies and the equilibrium studies indicates that
the high values for yy,p cannot be attributed

entirely to experimental error. Since the activity
coefficient reflects any approximations in the
published estimates used in the computations, it
was of interest to examine the numerical factors
closely in a search for leads to revised estimates.

In order that the activity not exceed unity, the
activity coefficient, y\,¢ (sy based on the solid

as the standard state should not exceed 5 x 102 at
600°C, because the solubility at 600°C is greater
than 2 x 10~3 mole fraction. (Blood, Redman, and
Topol® found the NiF, solubility to be about
2 x 10=3 mole fraction at 600°C.) The saturating
phase is a compound of NiF, and other com-
ponents, and the compound Ni|=22 is more soluble
than the saturating phase, or, in other words, o
solution saturated with NiF, is metastable with
respect to g solution less rich in NiF, in equi-
librium with o ternary compound. Thus, if the
solubility data are correct and if the discrepancy
in the activity coefficients is assumed to result
solely from the estimated value of the free energy
of formation of NiF,, then the free energy of for-
mation must be shifted to a smaller negative value
by an amount sufficient to provide at least a
factor of 4 in the activity coefficient. At 600°C
this amounts to about 2.5 kcal, which would revise
the free energy of formation from —128.0 keal/mole
to —=125.5 keal /mole.

Further information on the plausibility of the
high values for the activity coefficients of NiF
can be obtained from considerations of coefficients
based on pure liquid as the standard state. For
the pure liquid, ideal behavior corresponds to an
activity coefficient of unity, and this sets an
approximate upper limit to the activity coefficient

 

5L, E, Topol, ANP Quar. Prog. Rep. June 10, 1956,

ORNL'2‘°6, P 103-1 05, esp Fig- 2:2.4.

122

| AH
(N |"-y—§3—= In () =~_-—f-(_l_-_]_)
Y

of Nin. This stotement is based on previously
discussed correlating principles that involve the
charge-to-radius ratio  Z/R and which . indicate
that positive deviations will not be -found in
mixtures of ions of low polarizability.é6 Moreover,
there is little reason to expect much difference in
the behavior of Ni** and Fe** in dilute solution.
The conversion of activity coefficients based on
the solid as the standard state, y(s), to those

based on liquid as standard state, y{I), requires a

knowledge of the melting point and the heat of
fusion of the solute. When Brewer's7 estimates
of the values of these quantities are employed,
YNIF, (1) is 560 at 650°C and YFeF 1) is 0.66
at the same temperature.

There is a good possibility that the estimates for
NiF, given by Brewer are incorrect, and therefore
an alternative method of estimation has been
developed. The alternative method depends on @
comparison between the behavior of MgF.,, for
which a phase diagram is available,® cmj that
of NiF ,, for which the requisite phase diagram data
are missing.

Since the charge-to-radius ratio Z/R of MgF,
is the same as that of NiF,, the NaF-NiF, system
should show the same deviations from ideality
as the NaF-MgF , system shows at the same compo-
sitions. At a given liquidus composition the
activity of MgF , in the NaF-MgF , binary should be
the same as the activity of NiF, in the NaF-NiF,
binary. The equation for the activity of the
separating component at the liquidus is approxi-
mately

R T T

0

AS, . To
R T /'

where a is the activity of the separating component,
AH, is the heat of fusion, AS, is the entropy of

 

$f. F. Blankenship, ANP Quar. Prog. Rep. March 31,
1957, ORNL-2274, chap. 2.3, p 123-127. .

L, Brewer, Natl. Nuclear Energy Ser. Div. IV 19B,
202 (1950),

8a. G. Bergman and E. P. Dergunov, Compt. rend.
acad. sci. URSS 31, 753-754 (1941); English version,
E+. M. Levin, H. Fs McMurdie, and F. P, Hall, Phase
Diagrams for Ceramists, American Ceramic Society,
Columbus, 1956, '

 
 

 

 

 

"

1]

fusion (AH /To), o is the melting point of the
pure component, and T is the liquidus temperature.
From Eq. 1, it is found that

oA T si (T
@ = T 'R T )

where the unprimed quantities refer to the NaF-

N:F system and the primed quantities to the -
s P ' . estimated melting temperature of NiF

NaF-MgF system. A further simplifying approxi-
mation is that, since MgF, and NiF, both have
the same structure (rutile), the entropy of fusion
should be the same for both substances. Based
on this assumption

 

a) T T
To To

The approximate value of the melting point of
NiF, can then be calculated from Eq. 3. The
melting point of MgF T, or IS taken to be 1543°K
(ref 8); the liquidus temperature, T, of a mixture
that is 44% NaF ond 56% NiF, is reported to be
1310°K (ref 9); and the estlmated liquidus temper-
ature, T, of a 44% NaF and 5% MgF, mixture is
about l]]0°|< This estimate was based on the
relation between the activity and the freezing-point
depression, with 0.18 as the activity of ‘Mng.
The value 0.18 was chosen in order to be con-
sistent with the deviations from ideality in

 

?H. A+ Friedman and B. S. Landau, ANP Quar. Prog.
Repo Jufle 30, 19570 0RNL'2340; P ‘28.

PERIOD ENDING SEPTEMBER 30, 1957

comparable systems and represents an improvement
over the direct visual extrapolation from the
published diagram.® The estimated melting point
of NiF, is then about 1825°K, or 1550°C. Pre-
liminary experiments indicate that this value for
the melting point of pure NiF, is approximately
correct. The calorimetrically measured heat of

fusion of MgF, is 13,900 cal/mole and the entropy

of fusion is 905 eu (ref 10). By utilizing the
and the
assumption that the entropy of fusion is the same
for NiF, as for MgF,, y(I)/y(s) for NiF, was
calculated at various temperatures, T, from Eq. 1.
The values obtained are presented in Table 2.2.6,
together with values based on Brewer's estimates.”
Apparently Brewer’s estimate of 1300°K for the
melting point was 500°K too low, and his value
of 8000 cal/mole for the heat of fusion was low by

o factor of 2,

The tentative values gwen in the second column
of Table 2.2.6 are much lower than those estimated
with the use of the values given by Brewer (third
column). A more definitive set of conversion
ratios must await a more detailed study of the
NaF-NiF, phase diagram and, more importantly,
a better value of the melting point of NiF

VYalues of the activity coefficients o? NiF, in
NaF-ZrF, mixtures based on the solid and the
liquid as standard states are presented in Table

2.2.7 that were obtained by utilizing the calculated

 

10k, «. Kelly, U.S. Bur. Mines, Bull. No. 476, 105
(1949).

Table 2.2.6. Tentatlve Ratios af the Activity Caefhefents oi NIF Based on Liquid as the Standard State,
'y(l % to 'l'hase Based an the Solid as the Standard State, Y(s), ot Various Temperatures :

 

-'l_'e'ntperatare

 

7 Calculated 'y(l )/y(s) for e ¥(1)/y(s) Obtained from Brewer's

o Ty=1s°C ‘Estimates
550 - o 0.0039 0.17
Ce0 .- o007l 0.22
a0 . ooz 0.28
700 009 0.35
70 0029 0,43
0. ooe2 0.52
850 0,059 0.62
900 04080 0.72

 

123

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 2.2.7. Activity Coefficients of Ni Fz in NaF . 62rl’~'4

 

 

Temperature y(!) Based on Brewer's Values of
(°C) )’(3)* yi) TO and AH f
550 2468 9.6 413
575 2292 12.6 436
600 2076 14.7 455
625 2009 1%.1 500

 

*The magnitude of this number depends on the correctness of the free energy of formation of pure Ni F2. An error

of 4 kcal/mole can lead to an error of a factor of 10.

values of the conversion ratio, as given in
Table 2.2.6. If the factor of at least 4 for the
correction to the standard free energy of formation,
discussed above, is divided into the values of
ysz(” in the third column of Table 2.2.7, the

revised values for the activity coefficients of
NiF_, bosed on liquid as the standard state are
numgers less than 5, and thus most of the dis-
crepancy is resolved.

REDUCTION OF UF‘ BY STRUCTURAL METALS
Jo Do Redman

Data obtained from filtration studies of the
reduction of UF , by vanadium metal in the reaction
medium NaF-LiF-KF (11.5-46.5-42 mole %), as
reported previously, !V indicated that V° is stable
in contact with UF , dissolved in this solvent.
Subsequent studies demonstrated that the analysis
of these materials was in error and that V° was not
nearly so stable as first reported. Additional
studies of this system in the NoF-LiF-KF mixture,
as well as in NaF-ZrF, (50-50 mole %), at 600
and at 800°C were made recently. Also, a brief
study of the reduction of UF , by tungsten metal in
the reaction medium NaF-Zrl'! (53-47 mole %) was
made in order to compare tli'ne behavior in this
solvent with that found earlier12 in the NaF-LiF-
KF mixture.

 

1, p. Redman, ANP Quar. Prog. Rep. Dec. 10,
1955, ORNL-2012, p 86+

12y, D. Redman, ANP Quar. Prog. Rep. March 10,
1956, ORNL-2061, p 93.

124

The results of the recent studies of the reduction
of UF, by V° at 600 and at 800°C in the two
different reaction mediums are presented in
Table 2.2.8. All runs were performed with 2 g of
V° added to the solvent in nickel equipment. In
the runs with the NaF-ZrF, mixture, 4.1 mole %
UF, (11.4 wt %) was present, and, in the runs
with the NaF-LiF-KF mixture, 2.3 mole % UF,
(15 wt %) was used. The studies of tungsten
were performed in a similar manner, with 4.0
mole % UF‘ (11.4 wt %) present in NaF-ZrF,
(53-47 mcle %). The results of the runs with
tungsten are given in Table 2.2.9.

The data given in Table 2.2.8 show that V°
is not stable in contact with UF, dissolved in
either type of fluoride mixture at 60?) or at 800°C.
There is some indication that higher equilibrium
vanadium concentrations exist in the NaF-ZrF
mixture at 800°C, but, in view of the relatively
few runs with the NaF-LiF-KF mixture and the
difference in the UF , concentrations, the data do
not show conclusively that V® is more reactive in
the NaF-ZrF, mixture than in the NaF-LiF-KF
mixture. The extensive attock of vanadium by
UF, in these fluoride mixtures indicates that the
use of vanadium alloys to contain these mixtures
would offer no real advantage. '

The results given in Table 2.2.9 show that low
equilibrium tungsten concentrations occur under
these conditions and demonstrate that W is stable
with respect to UF, dissolved in the NaF-ZrF,
mixture, These results are in marked contrast to
those found earlier with the NaF-LiF-KF mixture
as the reaction medium. Tungsten concentrations
of 1100 and 1400 ppm at 600 and ot 800°C, re-

spectively, were found in this medium.

 
 

 

 

 

W

PERIOD ENDING SEPTEMBER 30, 1957

Table 2.2.8. Data for the Reaction of UF4 with V° in NuF-ZrF4 (50-50 Mole %)
and NaF-LiF-KF (11.5-46.5-42 Mole %) at 600 and 800°C

 

 

 

 

Conditions of Equilibration Present in Filtrate

Reaction Medium Temperature Time u V* Ni
(°C) thr) (wt %) (ppm) (ppm)

Na F-Zl'F4 600 3 7.4 470 120

(50-50 mole %) 3 8.1 500 25

5 8.8 420 25

5 10.3 390 35

800 3 8.3 2200 70

3 8.3 2150 30

5 8.3 1850 30

5 8.3 1600 70

NaF-LiF-KF 600 3 11.4 550 25

(l 165-4645-42 mole %) 11.1 750 45

800 3 11.0 1450 25

3 10.6 1350 20

 

*Blanks of 185 and 150 ppm of V at 800°C in NaF-ZrF, and NaF-LiF-KF, respectively.

Table 2.2.9. Data for the Reaction of UF4 with W°
in NaF-ZrF, (53-47 Mole %) ot 600 and ot 800°C

 

 

 

Conditions of Equilibration Present in Filtrate
Temperature Time U W Ni
(°C) (hr) (wt %) (ppm) - (ppm)
600 3 9.2 70 10
3 9.1 120 130
5 9.0 95 45
5 941 70 50
800 3 9.1 130 - 45
3 93 100 . 40
5 85 95 30
5 ' 105 20

8.9

 

*Blank of 5 ppm of W at 800°C.

STABILITY OF CHROMIUM FLUORIDES
IN NoF-LiF-KF .

J. D. Redman

Previous studies'3 of the behavior of CrF
dissolved in NaF-LiF-KF (11.546.5-42 mole %)

 demonstrated that Cr*** is stable in this melt

at 800°C if contained in nickel. Similar studies
have now been made with chromium metal present.
Data given in Table 2.2.10 show the instability
of CrF; under these conditions. An equilibration

- period of 5 hr was used prior to filtration. Previous

studies 14 also showed that -Cr'F2 was not stable

 

13, p. Redman, ANP 'Qudr. Prog. Rep. Dec. 10,

1955, 0RNL'20]2' P 88" ANP Quar- Prog. REPO Sept. 10,

145, D, Redman and C. F. Weaver, ANP Quar. Prog.
Rep. March 10, 1955, ORNL-1864, p 61; ANP Quar.
Prog. Rep. June 10, 1955, ORNL-1896, p 63

125

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 2.2.10. Stability of Chromium Fluorides in
NaFeLiF-KF (11.5:46.5-42 Mole %) at 800°C

 

Initiclly Present Present in Filtrate

 

cett cettt cP cett Total Cr

(wt %) (wt %) (wt %) (wt %) (wt %)

 

1 0.87
1 0.06 0.86
3 0.38 2.4
3 0.27 2.6
5 0.68 4.1
5 0.75 4.4
1 5 0.30 1.29
! 5 0.32 1.33
2 5 0.64 2.53
2 5 0.47 2.58
3 5 0.37 4.1
3 5 0.95 4.2
3 S 0.49 4.3
3 5 0.98 4.0

 

at 800°C in the NaF-LiF-KF mixture and that it

probably disproportionated according to the reaction
3CrF, :_2CrF3 + C°,

Additional studies have now been made of this
system in a further attempt to determine the
equilibrium concentrations and for comparison with
results obtained when CrF, and Cr° are equi-
librated. The results are inclauded in Table 2.2.10.

Ifsthe Cr** values are ignored and only the total
chromium concentrations found in the filtrates are
considered, nearly constant values for the Cr**-to-
Cr*** ratic may be calculated. Such treatment
shows that about 60% of the chromium is present as
Cr** ond about 40% as Cr*** in all cases,
irrespective of whether CrF, or a mixture of CrF,
and Cr® was initially present. The concentration
equilibrium constants calculated from the mole
fractions of CrF, and CrF, present are not
constant, however, if this constant Cr¥**to-Crt*
ratio is used. They decrease from a value of
approximately 50 for the lowest total chromium
concentration to about 10 for the highest. These

126

vaolues are much greater than the 10—3 obtained
for the equilibrium constant by using the values
available for the standard free energy of formation
of CrF,and CrF, at 800°C. No reasonable balance
of Cr**, Cr***, and total chromium can be obtained
by using the experimentally determined valuves for
Cr**, and it thus appears that the experimental
values are probably inaccurate.

REDUCTION OF FeF, BY C° IN RbF-ZrF,
Jo Dc Redman

Data presented previously 'S for the reaction
FeF, + Cro=—CrF, + Fe®

in NaF-ZrF 4 (53-47 mole %) at 600°C showed that
the reduction of FeF, by Cr® was essentially
complete at the FeF, concentrations employed,
and, as a result, no reliable values for the CrF -
to-FeF, ratio (or K.) could be calculated. A
calculation of the CrF,.to-FeF, ratio from the
values found for K_ for the reaction

M + 2UF —MF, + 2UF, ,

where M is Cr or Fe, yielded o value of approxi-
mately 60 at 600°C with NaF-ZrF, (53-47 mole %)
as the reaction medium. This value is much larger
than the 6 x 10=3 obtained by calculating the
equilibrium constant from the values given for the
standard free energy of formation for CrF, and
FeF, at 600°C and shows that the system is far
from ideal.

A valuve of ~60 for the Crt*to-Fe** ratio was
found in the NaF-ZrF ; mixture at the highest FeF,
concentration (1.0 wt %) used in the previous
studies. Additional experiments have been per-
formed with higher l'-'eF2 concentrations to de-
termine whether the value for the Cr*‘-to-Fe**
ratio would increase with increasing FeF, ad-
ditions. Extremely erratic results were obtained
with 1.5 wt % FeF,. It appears probable that the
solubility of Cr** was exceeded and that the
presence of a solid phase consisting of a ternary
compound of NaF, ZrF , and CrF, was responsible
for the inconsistent results. |f this assumption is
correct, reliable values at the higher FeF, concen-
trations cannot be determined in this reactien
medium.

 

15}, D. Redman, ANP Quar. Prog. Rep. June 30, 1957,
ORNL-2340, p 132,

li

o

 
 

 

 

A study of this reaction was. also made with
RbF-ZrF , (52-48 mole %) as the reaction medium,
This medium sheuld yield a value of about 2 to
3 for the Crt*to-Fe** ratio, based on the data
available for the reduction of UF, by Cr° and Fe?,
and therefore be more adaptable to the experimental
determination of the Crt*-to-Fe** ratio. Blanks
were run in which 2 g of iron wire was hydrogen-
fired at 1000°C in a nicke! charge bottle; approxi-
mately 40 g of the RbF-ZrF, mixture and 2 g of
pure chromium were added; ond the system was
equilibrated at 600°C for 5 hr prior to filtration.
The experimental runs were made in a similar
manner, and the desired quantity of FeF, was
added. Data are reported in Table 2.2.11 fzor the
blanks and for the runs in which FeF, was added.

PERIOD ENDING SEPTEMBER. 30, 1957

The iron concentrations present in the blanks
approximate those found in the NaF-ZrF, mixture
and in the charge material. This agreement
suggests that the iron is present as metal suf-
ficiently fine to pass the filter rather than os
dissolved Fe**. The data show that the reduction
of Fe**is essentially complete at all Fe** concen-
trations studied. The low iron concentrations
present in these runs preclude any accurate
evaluation of the Cr**to-Fe** ratio, but it is
evident that this ratio is much larger than the
value of 2 to 3 calculated from the data available
for the reduction of UF by Cr®and Fe®. In contrast
to the good agreement between the experimentally
determined and the calculated chromium concen-
trations found in the NaF-ZrF, mixture, the data

Table 2.2.11. Data for the Reaction Fer + Cf":chz + Felin RbF.sz" (52-48 Mole %) at 600°C

 

Present in Filtrate

 

FeF, Added (ppm of Fe)

Calculated Cr* (ppm)

 

Ni (ppm) Fe (ppm) Cr (ppm)

Blank 30 90 315

40 100 215

40 85 235

35 100 280
995 55 100 1305 1185
845 20 95 1080 1045
890 20 90 1055 1090
850 35 95 1120 1050
1920 ' 110 135 2335 2050
1860 35 95 2305 1970
1900 20 90 2360 2030
196 30 105 2405 2070

4050 35 ous a5 4030
0 35 85 4610 3870

3900 25 %0 470 3880
30 30 0 480 390
o em0 50 105 . 7365 - 5880
00 .80 Mo 700 . 590
600 .. 75 s 7320 . 590
6050 30 90 - 7610 5850

 

*Complete reduction of FeF2 assumed.

127

 
 

 

 

ANP PROJECT PROGRESS REPORT

given in Table 2.2.11 show that in RbF-ZrF,
the calculated chromium concentrations are lower
than the experimentally determined concentrations.
This difference is approximately 20% for an initial
Fe** concentration in the range of 2000 to 6000
ppm. No satisfactory explanation can be given for
the large Cr**-to-Fe** ratic or for the difference
between the calculated and the experimentally
determined chromium concentrations. Additional

experimental work is planned in an attempt to -

resolve these discrepancies.

SOLUBILITY OF STRUCTURAL METAL
FLUORIDES IN RbF-ZrF,

J. D. Redman

The results obtained from studies on the solubility
of CrF, FeF,, ond NiF, in NaF-ZrF, (53-47
mole %),16 in LlF-ZrF (52-48 mole %,'7 ond
in KF-ZrF , (52-48 mole %)” were reported previ-

H

SOLUBILITY (wt % M)
W

0 1 2 3 4

 

ously. The data obtained at 600°C demcnstrated
that the solubilities of CrF, and of FeF., are
functions of the excess of structural metal fluoride
present,  The solubility of NiF, showed this
behavior in the NoF-ZrF, and LlF-ZrF mixtures
but not-in the KF-ZrF, m:xture. The change in
solubility at 600°C is due to the presence of a
solid phase consisting of a ternary compound
containing more ZrF, than alkali fluoride and the
resulting enrichment of the liquid phase with
respect to alkali fluoride.

More recently the solubility of the three structural
metal fluorides has been studied at 600 and ot
800°C with RbF-ZrF, (5248 mole %) as the
solvent. The results are presented in Table 2.2.12,
and the 600°C values are plotted on Fig. 2.2.4,

 

16, D. Redman, ANP Quar. Prog. Rep. June 10, 1956,
ORNL-2106, p 101.

7). p. Redman, ANP Quar. Prog. Rep. Sept. 10, 1956,
0RNL‘2]57. P 100-

UNCLASSIFIED
ORNL—-LR-DWG 25920

CrF, IN LiF-ZrF,
CrFy IN NaF=ZrF,

1€
o
&

s

<
0
\V\@

C,-(<

F, IN NaF-2rF,

eF, IN LiF-ZrF,
iF5 IN RDF-ZrF,

NiF IN LiF~Zrf, NiF, IN NaF — ZrF,

5 6 7 8 9 10

METAL FLUORIDE ADDED {wt % M**)

Fig. 22.4. Solubility of Structural Metal Fluorides in Alkali Fluoride=ZrF , Binary Systems at 600°C,

128

"

 
 

PERIOD ENDING SEPTEMBER 30, 1957

Table 2.2.12. Sclubility of Structural Metal Fluorides in RbF-ZrF4 (52-48 Mole %) at 600 end 800°C

 

 

 

 

 

 

 

. Additive Temperature . Present in Filtrate
lon  Quantity (wt %) (°C) " Fettwt%) Fewt® Cttwt®m) Crwt®  Ni(ppm)
. Nit* 1.0 600 2,400
1.0 600 2,400
- 4.7 600 5,400
4.7 600 5,200
1.0 800 5,400
. 1.0 800 5,400
4.7 800 13,800
4.7 800 13,800
crtt 1.0 600 0.54 0.76 55
1.0 600 0.68 0.76 ]
5.0 600 2.0 2.8 120
5.0 600 2.1 2.7 1
15.0 600 8.3 9.4 135
15.0 600 63 7.4 110
5.0 800 2.6 344 160
5.0 800 2.3 3.5 220
10.0 800 7.3 8.9 60
. 10.0 800 7.8 9.1 55
20.0 800 9.0 15.6 190
20.0 800 9.5 15.7 250
. Fett 1.0 600 0.77 0.99 155
1.0 600 0.87 1.00 175
5.0 600 1.2 1.8 150
5.0 600 1.4 1.7 110
4.0 800 3.4 4.0 215
= 4.0 800 3.2 4.2 240
10,0 800 66 7.9 430
100 800 61 91 660
15.0 - 800 T 9.9 110
. 15.0 800 7.4 9.2 130 -
ottt O 600 0,26 0.48 - 825
S 10 600 0.21 051 . 865
30 600 0,40 079 410
3.0 0 - - 600 0.76 078 390
L0 800" 0:37 071 1,970
100 - 7800 037 0,81 1,800
) &0 . 800 113 20 3,030
50 800 1408 149 2,640

129

 
 

 

 

ANP PROJECT PROGRESS REPORT

Table 2.2.12 {continued)

 

Additive

Present in Filtrate

 

 

 

Temperature . .
len Quantity (wt %) (°C) Fe**wt%) Fewt® Cettwt%) Cr(wt%) - Ni(ppm)
Fettt I W 600 1.1 16 1,890
1.5 600 1.3 1.4 1,490
2.0 600 14 1.7 2,700
2.0 600 1.1 2.2 _ 2,100
3.0 800 2.4 2.9 5,400
3.0 800 2.3 2.8 | 5,100
7.0 800 3.0 4.7 4,700
7.0 _ 800 3.7 4.7 4,500

 

along with data for the other solvents. As may be
seen in Fig. 2.2.4, the solubility at 600°C of all
the metal fluorides studied in the four solvents
increases ‘with increasing additions of the metal
fluoride, except in the case of NiF2 in the KF-
ZrF , mixture. In the case of FeF, the solubilities
in the different solvents for @ 5 wt % addition of
Fe** increase in the order Li, Na, K, and Rb,
with o total increase of about 50%. For a 1 wt %
addition of Fe** the order is different, and the
solubility in KF-ZrF  is a factor of 3 higher than
the solubility in NaF-ZrF,. The solubilities
found for CrF, in the NoF-, LiF-, and RbF-ZrF,
mixtures do not differ greatly, but the solubilities
found for CrF, in the KF-ZrF, mixture are con-
siderably higher. A comparison of the solubilities
at 800°C is not presented, mainly because the
FeF, and CrF, solubilities are so high. For
example, as ‘may be seen in Table 2.2.12,
solubility of approximately 16 wt % of Cr was
found for @ 20 wt % addition of Cr**, In the case
of NiF, the solubilities found in the NaF-, KF-,
and RbF-Z¢F , mixtures at 800°C range from 1.1 to
1.5 wt % Ni ** for 5 wt % additions of Ni**, whereas
the solubility is about 2.2 wt % in the LiF-ZrF
mixture. The difference is considerably less for a
1.5 wt % addition. ,
The results obtained from the runs in which
either CrF, or FeF; was equilibrated in the RbF-
ZrF, mixture show that both Cr*** and Fet*t
are partially reduced to Cr** or Fe** by the nickel
container. In virtually all cases no balance
between the Ni** found in the filtrate and the
quantity of divalent metal ion present was ob-
tained. This is, undoubtedly, due to the limited

130

solubility of Ni**. The presence of Cr** or Fe*?
represses the solubility of Ni**, and, as a result,
the solubility of Ni** in these runs is considerably
less than that reported for Ni** in Table 2.2.12,
where only Ni** was present.

ACTIVITIES IN ALLOYS
S. Langer

important factors that will influence the corresion
behavior of the new nickel-molybdenum alloys that
are being developed to contain molten fluoride
salts are the activities of the metallic. constituents
of the alloys. Ideally, it would be desirable to
know the activity of each constituent metal in the
alloy as a function of temperature and alloy compo-
sition. If, in addition, the thermodynamic properties
of the molten salts were known, it should be
possible to predict the corrosion behavior of the
molten salt—-alloy system. Data are now being
obtained relative to the activities of the con-
stituents of Cr-Ni-Mo alloys.

Panish!8 has studied the activity of Cr in the
Cr-Ni system ot 750 and at 965°C, and studies of
the activity of Ni® in Ni-Mo alloys have been
initiated. The cells being studied are of the type

Ni |NiCl, [Ni (alloy)
where the NiCl, is dissolved at a concentration of

0.5 mole % in f\laCI-KCI eutectic. The first goal
is to demonstrate the feasibility of activity

 

18y, B. Panish, ANP Quar. Prog. Rep. March 10,
1956, ORNL=2061, p 92; ANP Quar. Prog. Rep. June 10,
1956, ORNL-2106, p 93; ANP Quar. Prog. Rep. Sept. 10,
1956, ORNL-2157. p 100. ‘ .

 
 

 

 

 

measurements of the Ni<Mo alloys by the emf
method. The side reaction

NiCl, + Mo®==MoCl, + Ni°

appears to have a negligible influence, since
NiCl, is more stable by about 12 keel per gram-
atom of chlorine than is MoCl,. However, the
diffusion of Ni in Ni-Mo alloys may not be rapid
enough to prevent nonequilibrium surface con-
ditions on the alloy electrode. .

Preliminary experiments have been run, but no
relicble or reproducible data have yet been ob-
tained. Operation of the cells has been complicated
by the existence of a large temperature gradient
(about 15°C) along the electrode. Methods of
eliminating this gradient are now being tested.
Further, the time and temperature required to
produce a fully annealed alloy electrode are not
yet known.

EMF MEASUREMENTS IN MOLTEN SALTS
L. E. Topol
Daniell Cells in Fluoride Melts
Measurements of Daniell cells of the type

M |MF2(C|) I I.M'Fg(CQ)IIM'

where M and M’ cre the metals Fe, Ni, or Cr in
NaF-ZrF, (53-47 mole %) ond in KF-LiF (50-50
mole %) were continued. A helium atmosphere is
used for all experiments, and electrical conductivity
between the half-cells is attained with ZrO
bridges previously impregnated with the solvent.
The results obtained with iron and nickel couples
at 650°C, corrected for thermoelectric potentials
but with the liquid-junction emf’s neglected, are

listed in Table 2.2.13. The results are in accord

PERIOD ENDING SEPTEMBER 30, 1957

with previously reported values,!? and aill the
results lead to a value of (6 + 1) x 102 for the
ratio of the activity coefficients of NiF, and
FeF, in NaF-ZrF, (53-47 mole %) ot 650°C. If
the activity coefficient for FeF, is taken to be
3 (ref 20), the activity coefficient of NiF,, as
found from emf measurements, is (1.8 £ 0.3) x 103,
based on the solid as the standard state,

Results obtained for CreNi couples in NaF-ZrF,
at 650°C are presented in Table 2.2.14. All the
nickel half-cells were contained in nickel crucibles,
but various vessels were used for the chromium
half-cells.

When graphite, copper, or HfO, crucibles were
employed, the emf’'s were found to decrease rapidly
with time, and, if the initial potentials were used,
activity ratios of 50 to 300 could be calculated.
These results indicate that the choice of a con-
tainer material for CrF, melts is important. The
possibility that the inconclusive results were due
to the presence of Cr***, as well as Cr**, cannot
be completely ruled out. Further measurements
with chromium-plated nickel crucibles may resolve
this problem. It should be mentioned that ex-
aminations of the melts contained in Morganite
with the petrographic microscope revealed the
presence in the melts of some aluminum compound
other than A|20 . The melt in the boron nitride
vessel undoubtealy contained 8203 in sufficient
amounts to affect the measurements.

 

19, E. Topol |
« E. Topol, ANP Quar. Prog. Rep. June 30, 1957,
ORNL-2340, p 149, ’

20Bgsed on interpolations between values obtained at
600 and 700°C by C. M. Blood and G. M. Watson; see
ANP Quar. Prog. Rep. March 10, 1956, ORNL-2061, p 84.

" Table 2.2.13. Activity Coefficient Ratios §f Nin to FeF, in Fused Fluorides
Obtained from Daniell Cells at 650°C

 

Concenflaiifin.of NiF2 : Concentration of FeF2

Ratio of Activity Coefficients,*

 

. (mole %) (mole %) Solvent yN'FZ/yF.Fz
0375 | 0,255 NaF-Z¢F, 505

"~ 0.387 N 0.185 NaF-ZrF 4 ‘ 620
0.247 0,246 KE-LiF 162

 

*RBased on the solid as the standard state.

131

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REFPORT

Table 2.2.14. Activity Coefficient Ratios of NiF to CrF in NaF-Z¢rF . (53-47 Mole %)
Obtained from Daniell’ Cells at 650°C

 

Concentration of Ni F2

 

 

(mole %) c°““"(:::'|:";‘;’f CrFy Container for CrF, yN'leyc'Fz
0.284 | 0.220 ALO,* 173
0.283 0.226 Al 203 187
0.275 0.481 AL,O, 630
0.25 0.559 ALO, 580
0,283 1.042 AL, 780
0.277 0.752 BN ' 1050
0.273 0.246 Pt - 309
0.265 | 1.050 Ni 301
0.277 1.098 Ni | 223
0.235 0.521 Ni 339

*Morganite.

Emf Measurements of Chloride Melts

As discussed previously,21an emf study is being
contemplated of the effects of various compositions
of the solvent LiCI-ThCl, on a suitable solute in
dilute concentration. Although AgCl would be
useful as an index solute because of the good
repreducibility of the silver electrode, the acidity
of AgCl, as predicted by the ratio of its ionic
charge to its radius, is greater than that of LiCl.
Thus the activity coefficient of AgCl should
decrease continuously with increasing mole fraction
of ThCl ;. In order to obtain an activity coefficient
effect that will exhibit a predicted maximum when
plotted against the concentration of ThCl,, a
solute with an acidity intermediate between that
of LiCl and ThCl, will be necessary. The solute
NiCl, satisfies thls requirement, and the nickel
electrode is reproducible. However, the standard
free energy of formation of NiCl, ond thus its
standard potential for the temperature range in
question (about 700°C) are unknown. In order to
determine E °for NiCl,, the emf of the cell

Ni |NiCl (1) |CI,

 

2l E, Topol, op. cit., p 150,

132

or the cell
Ni | NiCl (sat), alkali chleride |Cl,

must be measurable at the desired temperature.
The first cell given above could not be used
becquse of the sublimation pressure of NiCl..
However, there was a possibility that saturated
solutions of NiCl, in KCI (or NaCl) could exist ot
700°C, as in the case of FeCl, (ref 22).
Accordingly, cooling curves were obtained
for various alkali chloride~NiCl, mixtures. The
mixtures were contained in platinum crucibles in
a helivm atmosphere, and the temperature effects
were measured with a Pt, Pt-=10% Rh thermocouple
that was dipped directly intc the melt. In the
KCI-NiCl, system there appeared to be an in-
congruentfy melting compound at 730°C for high
N|C| composmons, and thus there is apparently
no llquld phase in equilibrium with NiCl, below
730°C. Samples from the L|C|-N|C| cnd NaoCl-
NiCl, systems contained solid solutlons, which
ellmlnated these systems from consideration. Since
there was the prospect that CsCl and RbCl would

 

22C. Beusman, Activities in the KCl-FeCI and
LzCl-FeC'I Systems, ORNL-2323 (May 15, 1957).

 
 

 

form very stable compounds with NiCl, and thus
provide NiCl, equilibria only at high temperatures,
these systems were not studied, and the use of an
NiCl, cell was abandoned.

Another solute that could conceivably meet the
requirements is PbCl,. Lead chloride melts at
501°C, and its standard potential can be measured
directly. A lead electrode requires that provision
be made for containing the lead in the molten
state at temperatures above 327°C; and thus the
electrode, which has the advantage of being strain-
free, consists of a pool of lead contained in o
quartz cup. Electrical contact is made with a
10-mil tungsten lead insulated from the melt by a
1-mm quartz tube.

The cell |
Pb(!) | PbCl (1) | CI ,{graphite)

was used to determme E®° for PbCIz.- Granular

lead was heated slowly in a vacuum to remove gas

and moisture; the lead chloride was treated
similarly. A graphite tube was preconditioned for
use as the Cl, electrode by heating in Cl
1000°C for several hours. The cell voltages
obtained in the early trials, although close to the
predicted values, were found to decrease slowly
with time. For one trial the PbCl, was treated
with HCl (Mathieson) and fused in pyrex; this
gave a salt which was gray in color instead of
white.  Another sample of PbCl, was treated
similarly with HCl produced by tfie reaction of
H,SO, on NaCi. The PbCi, agcin appeared
gray|s11, presumably because o? the presence of
organic matter in the salt. This same product was
then fused under chlorine. Upon solidification, a

fairly clear, white product resulted. The ‘salt was
“removed from its container in a'dry box and ground -
into small chunks for use in a-cell. Flakes of

lead ‘were prefiielted in a Vycor crucible under

—helium. Since attempts to clean the surface of the
lead "by reduction with H, were unsuccessful,

clean metal from below the surface was. pipetted

‘into the electrode holders. A cell with such

purified materials gave potentials within a few
millivolts of the theoret;cal values and ‘maintained

reprodumb:llty ‘within' 1 mv. for two days. The =
experiment - was concluded when the tungsten'

lead-wire broke at the end of the second day.

The measured emf’s corrected for thermoelectric

PERIOD ENDING SEPTEMBER 30, 1957

potentials are given in Table 2.2.15, along with
the literature values of E° (ref 23).

In order to ascertain the solvent composition

~ limits at 700°C, cooling curves were run on various

LiCl-ThCl, samples. Vacuum-tight fused-silica
tubes were used as containers for the thermocouples
in the initial experiments, but the heat effects were
so small that they could not be detected even with
a Pt, Pt-Rh differential thermocouple and a sensi-
tive amplifier. In order to measure the small heat
effects, the Pt, Pt-Rh thermocouples were inserted
directly into the melts, which were contained in
platinum crucibles in a helium-atmosphere furnace.
The temperature measurements were calibrated
against the melting points of pure NaCl and pure

LiCl oand found to be low by 11°C for both salts.

All the temperature measurements were therefore
increased by 11°C. The phase diagram of the
system was not completed, but the following
important features were observed. The liquidus
temperature drops from 613°C for pure LiCl to

- 560°C for a eutectic containing 20 mole % ThCl .

A compound that appears to be 2LiCl.ThCl, melts
at about 720°C, and a second eutectic occurs at
670°C that contains 37 mole % ThCl,. A second
compound, LiCl 3ThC| melts at some fairly
high temperature, that is, 900°C or higher. The
melting point of ThCl, is 820°C. Thus the LiCl-
ThCI4 solvent is liquid at 700°C for compositions
with 0 to 29 mole % ThCl‘ and 36 to 39 mole %
ThCi .

 

23y, J. Hamer, Ms Ss Malmberg, and B. Rubin, J.
Electrochem. Soc. 103, 8 (1956)

Toble 2.2.15. Standard Potential for PI:CI2
as a Function of Temperature

 

Temperature E®° Measured E° from Literature

 

(°c) v) v)
502 1.270 1270
526 14256 1.257
55  lo2d4] 243
600 1.213 1.215
654 1.179 1.186
702 1.151 1.160

 

133

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

SOLUBILITY OF HELIUM, ARGON, AND XENON
GASES IN NoF-ZrF‘-

N. V. Smith

Determinations were made of the solubilities of
argon and of xenon gases in molten NaF-ZrF,
(53-47 mole %)as functions of pressure and temper-
ature. The experimental procedures and assemblies
were the same as those used previously24 for
determining the solubilities of helium and of
xenon in molten NaF-ZrF .UF, (50-46-4 mole %)
and of helium in NaF-ZrF, (53-47 mole %). The
results obtained for argon and xenon are summarized
in Tables 2.2.16 and 2.2.17, respectively. The
data continue to show that in this solvent (1) the
solubilities of the gases follow Henry's law,
(2) the solubilities increase with increasing
temperature, and (3) the solubilities decrease with

 

24N, V. Smith, ANP Quar. Prog. Rep. March 31, 1957,
ORNL-2274, p 111; ANP Quar. Prog. Rep. Dec. 31,
1956, ORNL-2221, p 130; ANP Quar. Prog. Rep. June
30, 1957, ORNL-2340, p 140.

increasing atomic weight of the gas. The tabulated

~ results for argon appear to be consistent to within

better than 5%, but, based on possible experimental
uncerfainties, the over-all accuracy of the measure-
ments will be assumed to be 110%.

The numerical consistency of the xenon data
appears to be of the same order of magnitude as
that assumed for the argon data. Additional ex-
experimental uncertcinties were introduced, how-
ever, by the need for recovering the xenon. The
initial purity of the xenon was 99.9+%, and after
seven experiments the purity had decreased to
72%. Because of the limited amount of xenon
available, the saturating gas was sampled only
at the start and at the end of the series of experi-
ments. Since no mishap occurred with the xenon
recovery system in any of the experiments, it was
assumed that the xenon was uniformly contaminated
to the extent of 4% per experiment, and the partial
pressure of xenon was estimated from the total
pressure by making the corresponding correction.
Numerically, the estimoted corrections amount to

Table 2.2.18. Solubility of Argon in NuF-ZrF4 (53-47 Mole %)

 

 

 

 

Temperature Saturating Argon Pressure Solubility K*
(°C) (atm) (moles of argon/cm3 of melt)
x 10~7 x 10~7
600 0.530 0.263 0.496
1.02 0.495 0.485
1.50 0.792 0.528
2,06 1.07 0.515
Av 00506 i 000]5
700 0.526 0.420 0.798
1.05 0.860 0.821
1.55 1.25 0.808
2.09 1.68 0.801
Av  0.807 1 0.008
800 0.432 0.555 1.28
1.04 1.22 1.18
1.53 1.68 1.10
1.99 2.43 1,22

 

Av  1.20 £ 0.06

 

*K = ¢/p in moles of argon per cubic centimeter of melt per atmosphere.

134

 
 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

Table 2,2.17. Sclubility of Xenon in NaF-ZrF, (53-47 Mole %)

 

Saturating Xenon Pressure
Temperature (atm)

 

Solubility

 

n

(°C) (moles of xenon/cm? of melt) ke
Total Net* '

x 10-8 x 1078

600 1.04 - 0.96 1.77 1.85

' 198 - 1.90 3.84 2,03

l - Av 1.94

700 : 1.04 0.92 - 3.25 3.56
2.03 171 6.08 356

Av 3+56

4,48 6.32

800 N N

 

*Estimated on the assumption of a decreaee of xenon purity of 4% per experiment for the first four experiments.

**Defined in Table 2.2.16.

4, 8, 12, 'end 16%,7 respectively, for the tirst,

second, third, and fourth experiments. The fifth

and sixth experiments were not completed because

of a power failure and a break in the transfer Ime, '

respectively. After the seventh expenment the
saturating gas was sampled, and it was found to be

72% pure. Accordmgly, the largest estimated

correction introduced was 16% with the exception

of a 28% correction which was based on the actual

purity of the gas at the completion of the experi-
ment. Since the xenon was found to be contami-
nated with nitrogen, the contamination was probably
the result of a leak to the atmosphere in one of the

gas cylinders in which the xenon was periodically

frozen. These cylinders were the only portions of
the equupment subjected -to prolonged periods of

-evacuation. The rest of the assembly was under a
positive xenon pressure at all times during oper-
"ations . Therefore the ‘numerical v'aluesv"of the
xenon so1ubtlit|es are. considered to be wnhln
_120% of the true solubllmes. o ‘

The values ofr Henry s low 'ce'nst"ants for .argon,

' helnum, and xenon in. NaF-ZrF (53-47 mole %)
have been complled 'in Table 2.2 18 for com-

parison. - - The temperature dependence of these
solubility constants is illustrated in Fig. 2.2.5,
on which the heats of solution determined from the
slopes of the lines are also presented.

The values of the enthalpy and entropy changes
for each of the reactions are presented in Table
2.2.19. |t may be noted that the heats of solution
increase with increasing atomic weight of the gas.
Furthermore, the entropies of solution in each case
are quite small (less than 2 eu). Similar behavior
was exhibited by helium and xenon in the related

solvent NaF-ZrF -UF , (50-46-4 mole %), as shown

in Table 2.2.20.

Since it is possible that the uncertainties in the
determination of AH and of AS may be as large as
1 kcal and 11 eu, respectively, it appears that
there ore no significant differences in the values

‘in Table 2.2.20 that may be attributed to differ-

ences in the solvents used. ~ The possible
significance of the small numerical magnitudes of
the entropies of solution has been studied. From

-a practical point of view, as has been indicated by

Blander,25 the small value of the entropy means

‘that the temperature dependence of the solubility

can be predicted fairly accurately from a single

measurement and a simple calculation of the
- entropy change undergone by an ideal gas in

isothermal expansion. Additional suggestions23
point out the possibility that the free volume of
the gas in solution is o large fraction of the total

 volume, as has been observed in other solutions

of rare gases.

 

25\, Blander, private communication to Ne V. Smithe

135

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 2.2.18. Henry's Low Constants, K, for Helium, Argon, und Xenon in NoF-ZrF 4 (53-47 Mole %)

 

K = c/p [(moles of gas)/(cm3 of melt) (atm)}

 

 

Temperature
(°C) Helium Argon Xenon
x 10~8 x 10~8 x 10~8
600 21.6 T 1.0 5.06 £ 0.15 1.9
700 29.2 * 0.7 8.07 * 0.08 3.6
800 42,0 * 1.3 120 T 0.6 6.3

 

UNCLASSIFIED
ORNL-LR-DWG 25924

TEMPERATURE (°C)
800 700 600
(x10~%)
50

AH = 6.2 kcal /mole

HELiUM

N
o

.
o

AH = 8.2 kcal/mole

14,1

AH = 444 kcal/mole

K = cip [(moles of gas) / (cm® of melt) (atm) ]

 

9 10 14 12
10000/T(°K)

Fig. 22.5. Temperature Dependences of Sclubilities
of Noble Gases in NuF-ZrF‘ (53-47 mole %).

136

SOLUBILITY OF HF IN NaF-ZrF, MIXTURES
J. H. Shaffer /

Studies of the solubility of HF gas in NaF-ZtF
were continued in order to determine the effect of '
solvent composition on the solubility. The
solubilities in solvents containing 35, 40, and
47 mole % ZrF , have been determined, and experi-
ments with so1vents containing 20, 55, 60, ond
65 mole % ZrF , are under way. The experimental
results are summarized in Tables 2.2.21 and
2.2.22. The values of the heats and entropies of
solution for HF in the different solvents are given
in Table 2.2.23.

The data indicate that HF gas follows Henry's
law within the precision of the measurements
throughout the composition range studied; the
solubility of HF decreases with increasing temper-
ature; and the solubility of HF increases with
increasing NaF concentration. In addition, the
calculated results indicate that the solution of
HF in these solvents is an exothermic process
which liberates more heat as the concentration
of NaF is increased. The entropy of solution
changes from -5.2 to —6.2 eu with an increase of
7 mole % NaF. Such losses of entropy are probably
primarily rotational losses in the dissolved HF
as compared with gaseous HF. Preliminary results
of studies of the solubility of HF in the NaF-KF-
LiF eutectic, as reported in the following section,
indicate @ AH and AS of solution of about —-17
kcal/mole and =10 eu, respectively. These values
do not appear to be out of line when they are

 
 

 

 

»

-

PERIOD ENDING SEPTEMBER 30, 1957

Table 2.2.19. .Eniropy Changes Related to the Solution of One Mole of Noble Gos in Molten
NaF-ZtF, (53-47 Mole %) at 1000°K and 1 atm

 

Equilibrium Concentrations Entropy Changes

 

 

 

Ga (moles/ liter) Heats of Solution {cal/mole «°K)
s (kcal/mole) - 3 =
Gaseous State Dissolved State AS]“ A.S‘2 ASB
x 102 X 10-5
Helium 122 33.0 | 62 6.2 7.5 -1.3
Argon I 022 ’ : 90] 802 802 9-8 - 106
Xenon w2 43 BRTR 1. 1.8 -0.7

 

. ®For. reaction 1, X (g) = X(d) at equulibrium at one atmosphere where X is helium, argon, or xenon.
bEor reucflon 2, X(g) - X (&), which tukes into account Isothermal expansion of an ideal gas.
cFor reaction 3, X(g) '== " X (d), that is, reaction 1 minus reaction 2, which represents the solution of 1 mole of gas
if the concentrations in the gaéeous and liquid phases are held constant.

Table 2.2.20. Heats and Entropies of Solution of Noble Gases in Molten Fluoride Solvents
- ot Equal Initial and Final Concentrations

 

In NaF<Z¢F, (53-47 mole %)

In NaF-ZrF -UF , (50-46-4 mole %)

 

 

Gas

 

 

 

" AH (keal/mole) As (ev) AH (kcal/mole) AS (ev)
Hefium 602 -1 03 704 ~ 0
‘ Argon 8.2 ~T1.6
| Xenon ma =07 114 ~0
® compared with a heat of reaction of —16.8 kcal/mole  the transfer line. Therefore the experimental

calculated2é for the formation of a mole of NaHF2
at room temperature from NaF(s) _o_nd HF(g). '

SOLUB“JTY OF HF iN NGF-KF-LIF MiXTURES
Ro Bo Evans ’

and a previously purified salt mixture, but all runs

 

26F. R. Bichowsky and - Fo D. Rosslnl, Tbe “Thermo-

 

; 1936, |
7 27§, H. Shaffer and C- M. Blood ANP Quar. Prog.
) g?’l ]Marcb 31, ‘1957, ORNL-2274, p 116, esp Fig.

Determlnahons of the solubulit} of HF in NuF-‘_
“KF-LiF were first uftempted ‘with the iyse of the . -
nickel transfer apparatus descnbed prevmusly” .

' folled becuuse the melt perslstently broke fhrough -
was based on the rate at which helium strips HF

from the saturated melt.
chemistry of . Cbem:cal Subs:ances, Relnbold, New Yorl: .

apparatus was changed from the transfer type of
system to a simpler, single, nickel reactor system.

. The fluoride mixture used in the latter system was

prepared from the component salts and was purified

= by dlternately purging it with H2 and HF. Large
- amounts of hydrogen sulfide were detected in the

evolved gas during the HF cycle; however, by
‘the third doy of purlficuhon, fl-ns impurity had been
' 'ehmlnated

Two sets of tentative HF solubility constants
were then obtained. The first set of constants

_ The second set of
constants was based on a direct determination of

“the total amount of HF dissolved in the melt
(without transferring).

Values taken from the
experimental curves are shown in Table 2.2.24.

137

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 2.2.21. Solubility of HF in NcsF-Zl'F4 Mixtures

 

Solvent Composition Temperature

Saturating HF Pressure Solubility

 

 

K*
(mole % ZrF ) (°C) {atm) (moles of HE /cm? of melit)
x 10-3 x 10~3
35 700 0.46 0.65 141
1444 2.13 1.48
1.84 2.68 1.46
AV 1046 i 0003
800 1.42 1.51 1.06
2.56 2.70 1.06
AV 1006
40 600 1.32 2.02 1.54
2-22 3-32 ' 1049
AV 1152 i’ 0003
700 1.44 1.47 1.02
2.27 2.38 1.04
AV ]003
800 1.33 1.07 0.80
2.24 1.81 0,81

Av  0.81 % 0.01

 

*K = ¢/p in moles of HF per cubic centimeter of melt per atmosphere.

Table 2.2.22. Henry's Law Constants, K, of HF

in NcF--ZrF4 Mixtures

 

Temperature Solvent Composition

 

(°c) (mole % ZtF ) K+
x 10=5
600 35 2.4+
40 1.52
47 1.23
700 35 1446
40 1.03
47 0.93
800 , 35 1.06
40 0.81
47 0,73

 

*Defined in Table 2:2:.21,

**Extrapolated from temperature-dependence plot, since
liquidus temperature of NaF-ZrF 4 (65-35 mole %)is about

610°C.

138

Both sets of data in Table 2.2.24 yield a straight
line when plotted vs the reciprocal of the absolute
temperature. Because of the close agreement of
the two sets of data, each representing an inde-
pendent set of experiments, it is highly probable
that the true values will not differ by an appreci-
able amount from these tentative values., A
comparison of the data with those presented pre-
viously for the NaF-ZrF, (53-47 mole %) system
reveals that the corresponding HF solubilities in
NaF-KF-LiF are roughly 50 times those for the
NaF-ZrF , system.

These preliminory studies have demonstrated
that molten NaF-KF-LiF can be contained in
nickel reactors for long periods of time if the
mixtures are subjected to exhaustive purification.

SOLUBILITY OF CeF; AS A FUNCTION
OF SOLYENT COMPOSITION

w. T. Ward R. A. Strehlow

Experimental data on the solubilities of rare-
earth fluorides in NaF-ZrF -UF , (50-46-4 mole %)

have indicated that small variations in solvent

 
 

 

 

*

4

‘Toble 2.2.23. Heats and Entropies of Sclution of HF
in NaF-ZrF Mixtures at Equal Concen'rafions
in the Gcseous and Liquld Phases at 1000 K-

 

 

Solvent Comp'osi"__flon S AH_ - As
(mole % ZeF ) (cal/mole) (cal/mole-°K)
35 -6400 —6:2
40 . -5800-  —6:2
47  —4500 ~-5.2

 

Table 2.2.24. Solubility Constants forVDeterm!nlng
the Amounts of HF Dissolved in the Molten
NaF«KF-LiF Eutectic Mlxture_

 

Solubility Constant
[ (moles of HF)/(em® of melt) (atm)]

 

 

Temperature
(OC) By Stripping-Rate By Direct
-Determination Determination
x 104 x 1074
650 5.0 649
700 33 : 5.0
750 2.2 36

800 145 2.8

 

composition may appreciably affect the SOlub'ili_ty
of the rare-earth fluorides. In order to determine

the magnitude of this effect, a series of experi- -

ments is now under way to investigate the

solubility - of CeF, in NaF-ZrF, solvents con-
taining from 42 to 63 mole % NaF. As in previous -
experiments, Ce 141 is being used as o radlptrqcer.r‘
Two experiments have been completed thus far. .

The data obtalned in these experiments are com-

pared in Fig. 2.2.6 with data obtained for other -
_solvents. As muy be seen the solubility of CeF,
.-decreases as the’ proportlon of tetravalent c0n--~_'_
smuents decreases. “These data indicate that the
CeF, ‘behaves - more like NaF or BuF2 m ihese'-
: ',solvents than like ZrF‘. - :

A further experiment was: ‘carried out to defermme— L
whether the evaporation of. ZrF from the solventv‘
had an effect on the CeF, solublllty. The uranium -

of the solvent was used as tracer, and it was
- 1956, ORNL=2221, p 125; ANP Quar. Prog. Rep. March

found thot dunng a typicol expenmentflm_volvrlng
four filtrations the solute (UF,) was enriched to

SOLUBILITY (mole % CeFz)

FPERIOD ENDING SEPTEMBER 30, 1957

=pESRES
ORNL—LR-DWG 25922

 

8 9 10 1 12 13
10,000/°K

Fig. 22.6. Effect of Solvent Composition on CeF,
Solubility,

the extent of about 4% of its original value and
that the condensed ZrF , vapor contained virtually
no UF, (<0.2%), as was expected. The data
obfcmea indicate that the evaporation of ZrF,
does not exceed the accuracy of the determmaflons

(£5%).
PRECIPITATION OF CERIUM AS Ce203 FROM
MOLTEN LiF-KF EUTECTIC
J. H. Shaffer

ConSIderable work pertaining to the chemistry of
fission-product fluorides dissolved in NaF-ZrF,

mixtures with and without UF, has been in progress
. for over a year and has been regularly reported by
_‘ Ward,28 One of the results of this investigation
-has been -the discovery of the possibility of
~removal of some injurious rore-earth fluoride
';.flsslon products by a process of solid solvent
‘extraction with a noninjurious rare-earth flucride
as_the _solid solvent,  The necessary conditions

 

2"w T Ward, ANP Quar éro 7
« T . .Re.set.10196
ORNL+2157, p 110; ANP Quar.g Prog. Ret',p. Dec. fo'

31, 1957, ORNL-2274, p 105; ANP Quar. Prog. Rep.

- June 30, 1957, ORNL-2340, p 135.

139

 
 

 

 

et e e e e e 3 A,

 

 

ANP PROJECT PROGRESS REPORT

for the feasibility of the solid-solvent extraction:

process are (1) limited solubilities of the rare-
earth fluorides in the solvent and (2) formation
of solid solutions by the saturating phases of the
solutes and the solid solvent., Previous measure-
ments of the solubilities of LaF, in the NaF-KF-
LiF ternary eutectic revealed that these solutes
are extremely soluble in this solvent {more than
20 wt % at 550°C). Accordingly, other methods
of removal of injurious rare earths must be found,
if satisfactory fuel reprocessing methods are
eventuclly to be developed for reactors that use
the NaF-KF-LiF fuel solvent. The possibility of
precipitating the rare earths as oxides was
therefore considered worthy of investigation. In
order to simplify graphical presentation of the
results, a eutectic mixture of KF-LiF was used
rather than the NaF-KF-LiF ternary eutectic.
It is believed that the chemical behavior of
solutes in these two solvents is very similar, if
not identical.

A systematic investigation of the solubilities
and relative stabilities of some monovalent,
divalent, trivalent, and quadrivalent oxides was
planned.  Simultanecusly, o study of different
possible methods of formation of the oxides in the
solution was scheduled. The particular oxides
proposed for study in the prelimincry phase of the
investigation were Li, O, BeO, Mg0O, CaO, SrO,
BaO, Ce,0,, La,0;, Sm,0, ZrO,, aond UO,.
The methods of oxnde formahon included for |nma|
study were reactions with: (1) carbonates,
(2) alkali hydroxides, and (3) oxides of less
stability.

The three methods of oxide formation have been
studied, and some experimental inconveniences
and advantages inherent in each method have
been observed. The addition of a carbonate such
as K 2CO; to the solvent in order to form and
prec:pltate some insoluble oxide sounded very
attractive. However, it was found in two experi-
ments that the carbonate does not decompose
appreciably unless the system is heated to temper-
atures above 1100°C. The second method studied
was the addition of LiOH to the solvent in order
to form and precipitate an insoluble oxide. This
method was found to be quite effective in the
precipitation of cerium from the solution. One
objectionable feature is the rapid liberation of
steam, which can cause the solution to spill over
unless special precautions are taken. The third
method consists of the addition of a less stable

140

oxide in order to form and precipitate an oxide of
higher stability. This method has been found to
be quite satisfactory to precipitate cerium by using
calcium oxide.

No numerical values of solubilities of oxides
have yet been obtained, but some useful quanti-
tative information has been gathered regarding the

solubilities of Li,0, Ce, O, and CeF,. It has
been found that at 550°C the solubility of Li,0
in KF-LiF (50-50 mole %) is greater than 1.6 wt %;
the solubility of Ce 0 at 600, 700, and 800°C
is less than 0.1 wt %, ond at 600°C the solublllty
of CeF, is greater than 8.2%.

An indication of the relative stabilities of some
oxides may be obtained by comparison of some of
the calculated free energy changes for the for-
mation of the oxide from the fluoride. The
calculated values may also be compared with the
degree of completion of the reaction as measured

experimentally by determining the concentration

of cerium remaining in the liquid after a reasonable
amount of time has elapsed after the addition of
the particular oxide to the solution containing
CeF,. Since the melting point of the oxide may
also be a good indication of its stability ond
relative solubility, the melting temperatures are
also included in the comparisons shown in
Table 2.2.25.

Since the values for the free energies of for-
mation of rare earth fluorides may be quite
unreliable, it is not expected that quantitative
correlations can be made from the AF° values
presented in Table 2.2,25. These values may,
however, be helpful in interpolation between
experiments. For example, Li,O has been shown
to react to completion with Ce%: g+ and the values
of AF and the melting points Ilsted for Na,,0 and
for K,O indicate that 100% completion is expected
of these two reagents for the corresponding
reaction.

Present plans call for the determination of the
solubility of BaO and its extent of reaction with
CeF, before a study of tetravalent oxide solu-
bilities and relative stabilities with respect to
trivalent oxides is undertaken. If sufficient
differences are noted, the possibilities2? of
effecting separation of tetravalent oxides from
trivalent oxides will be seriously considered.

 

29w.' R« Grimes, private cfimmun.icufloh to J. H.
Shatfer.

17

w

 
 

 

 

 

&

4

PERIOD ENDING SEPTEMBER 30, 1957

Table 2.2.25. Standard Free Energy Changes for the Formation of Ce.‘,(f)3 at 1000°K According to the Reaction

 

 

3M)0(s) 6MIF (s)
_or + 2CeF3(s):Ce203 + or

3MO(s) ' 3MF2(s)

Oxi Oxide Melflngl Point AF® Cerium Precipitated
xide (°K) (keal)® %)

Li,0 2000 -37 100
Na,0 1193 -116
K,0 800 -139
BeO | 2823 +59
MgO 3173 +53 gg?
CaO 2873 +2 100°¢
$r0 2703 ~42
BaO | | 2196 -52

 

A, Glassner, A Survey of the Free Energies of Formation of the Fluorides, Chlorides, and Oxides of the Elements

to 2500° K, ANL-5107 (Aug. 1953)
b1y over 24 hr.
“Reaction complete in 90 min at 600°C.

STABILITY OF UO,F, IN FUSED SALTS
Ro Jo She“

In connection with a possible method of
reprocessing fused salt reactor fuels by the
precipitation of the oxides of some of the fission
products, experiments were conducted to investigate
the stability of UO ,Fy ina fused salt mixture.

Crystalline UO_F was ‘mixed with @& purified-

mixture of NaF-BeF, (57-43 mole %) in a platinum

crucible to give a 'I mole % concentration of the _
~ uranium compound and then heated in an inert
- atmosphere in the visual-thermal apparatus. There

was no visible change in the mixture until liquid
appeared, at approximately 330°C. All the UOF,

“particles had disappeared at 400°C, and the melt |
had a light rust color which became darker when °
- the :temperature ‘was increased to the maximum of
 480°C.. Gas ‘bubbles were .observed on the walls -
~of the  platinum crucible. Upon~ “cooling, " the
sohdlfled ‘mixture had a -pinkish-brown calor, -

while a similor mixture given the same treatment
in a nickel crucible was gray-white in appearance
and there was a black deposit that adhered to the

surface of the crucible that had been in contact
with the melt. X-ray diffraction examination of
the deposit indicated that it was largely U,0,.
Ground samples of the fused materials were found
to contain UO2 or U308 when examined by means
of the petrographic microscope. Chemical analyses
of the melts are presented in Table 2.2.26.

A portion of the mixture fused in platinum was

transferred to nickel filtration - apparatus and

filtered at 470°C. The filtrate contained 0.44 wt %
U, and the residue cpr_flained 5.80 wt % U. The

‘Table 2.2.26. Results of Chemical Analyses of Fused

NdF-Bécmuoz F, Mixtures

 

 

Somela’ v F Ni

| amele Wt%)  (wt%  (ppm)

CAs fused inplatinum 536 55.5 140

As fused innickel 526 550 1070
Starting mixture 5.100 ~ 58.3

(calculated)

 

141

 

 
 

 

 

ANP PROJECT PROGRESS REPORT

theoretical uranium concentration of the mixture
was 5.10 wt % U if calculated os UO,F, and
5.83 wt % U if calculated as UD,. The uranium
compound in both samples was identified as either
UO, or U, 0, by microscopic examination. The
concentration was too low for positive identificafion
by means of x-ray diffraction. It oppears from
these observations that nickel will readily reduce
UO,F, in a fused salt medium by the reactions:

UOF, + Ni—> U0, + NiF,
or

4UOF, + 2Ni—> U,0, + 2NiF, + UF,

It seemed evident also that UO_F, is more
stable in contact with platinum, as might be
expected, - but that a portion of the compound
present was reduced by some means or underwent
thermal decomposition. In order to investigate
the likelihood of thermal decomposition, a portion
of the UO,F, was heated alone in an inert atmos-
phere in the visuval-thermal apparatus with o
platinum crucible as the container. When the
temperature was increased to 120°C, a large
thermal effect was noted, and the exposed surface
of the material gradually darkened. The temper-
ature was further increased to the maximum temper-
ature of 470°C, and the sample was then allowed
to cool to room temperature, placed in a tared
screw-cap bottle, and weighed immediately after
removal from the inert atmosphere box. The
compound had lost 0.4% in weight, and it was
quite hygroscopic. It gained about 1.1% in weight

after standing in the screw-capped bottle for

10 days. The analyses of the material before and
after heat treatment are given in Table 2.2.27.

The exact nature of the chemical change in the
UO,F, used in these experiments is not clear,

Table 2.2.27. Chemical Anclyses of UO,F, Samples

 

U F H,0
(Wt %) (wt%)  (wt %)

 

Sompled before heating  75.3 12.3 6.37
Sampled after heating* 7642 12.4 6:16
Theotetical content 774 12.3

 

 

*Analyzed after standing in screw-capped bottle for 10
days.

142

but it seems obvious that hydrolysis did not
occur, because the fluorine content did not change
on heating. It has been reported39 that hydrated
uranyl fluoride can be dehydrated at 120°C without
serious decomposition and also that uranyl fluoride
formed at low temperatures is very hygroscopic.3!
This experiment seems to confirm both obser-
vations, except that the observed weight loss was
much less than would be expected from the reported
water content of the starting material.

OXIDATION OF MIXTURES COMPOSED OF
SODIUM AND POTASSIUM

E. E. Ketchen

The results of some preliminary studies of the
oxidation of a sodium-potassium mixture at room
temperature were reported previously.32  The
data indicated that a sodium-to-potassium ratio
of about 22 was obtained in the oxides formed by
the oxidation of a mixture of 50.4 wt % sodium
with 49.6 wt % potassium. Additional experiments
have shown, however, that the values obtained
were probably inaccurate and that the precise
results were due either to a very careful repro-
duction of experimental conditions or, more
probably, pure chance. ‘

The apparatus used in the recent studies is a
modification of the glass apparatus described
previously.32 The upper portion of the apparatus
was enlarged to a diameter of 2% in. to eliminate
the need for shaking the mixture during the
oxidation. The oxidation and filtration are per-
formed in the dry box in order to reduce the amount
of contamination by water ond the resulting
hydroxide formation. At present, three capsules
are loaded in the dry box, and each is oxidized
by intreducing dry air at a pressure of 1.5 to 2 cm
of Hg, evacuating, and repeating this cycle wuntil
four portions of air have been used. Filtration is
achieved by applying helium pressure to the upper
portion of the apparatus, forcing the alkali metals
through the glass filter disk, and leaving the
oxides with a small amount of metal in the upper
chamber. The residual alkali metals are removed

 

3°G. R. Dean, R. Bradt, and M. S. Katz, Chemical
Research-P-9. Report for Period Ending January 5—
March 1, 1944, CC-1382, p 20, '

31), J. Katz and E. Rabinowitch, Natl. Nuclear
Energy Ser. Div. VIII 5, 569 (1951).

32E, E. Ketchen, ANP Quar. Prog. R
. . y . £ ep. ]une 30
1957, ORNL-2340, p 156. P '

 
 

 

W

»

 

”»

 

from the oxides by mercury according to the
procedure described previously.  The results
obtained by using this method for the room-temper-
ature oxidation of two different sodium-potassium
mixtures are given in Table 2.2.28.

The lack of preciseness of the results suggests
that it is extremely difficult to control all the
variables which offect the oxidation process.
Traces of moisture should tend to reduce the
sodium-to-potassium ratio, and consequently the

PERIOD ENDING SEPTEMBER 30, 1957

larger ratios are probably closest to the correct
valve. It may be noted that the values given in
Table 2.2.28 that were obtained for the 51 wt %
Na~49 wt % K mixture are significantly larger than
the value of 22 obtained previously. The lack of
precision in the results makes it difficult to
ascertain the effect of composition of the alkali
metal mixture on the sodium-to-potassium ratio in
the oxides, although it appears that the ratio de-
creases with increasing potassium concentration.

Toeble 2.2.28. Compesition of Oxides Formed by Oxidation of Sodium-Petassium Mixiures'uf 30 t;a 3s°cC

 

Composition of Na-K Mixtures

Composition of Oxides

Ratio of Na to K

 

 

(wt %) Na (mg) K (mg) : in the Oxides

51 Na—49 K 13.8 0.47 29
13.5 0.30 45

10.9 0.27 - 40

7.3 0.17 43

6.7 0.18 ' 37

10.4 0.29 36

8.5 0.18 47

22 Na-78 K 8.4 0.32 26
8.9 0.29 31

10.1 0.29 ' 35

8.1 0.27 30

7.4 0.23 32

86 0.31 28

 

143

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

2.3. PHYSICAL PROPERTIES OF MOLTEN MATERIALS
F. F. Blankenship

' VAPOR PRESSURE OF l'lfl'-'4
S. Cantor

An appreciable difference in properties between
analogous compounds of zirconium and hafnium
can be of considerable importance in the separa-
tion of the two elements. Therefore the sublima-

tion pressures of HfF, were determined, by using

the Rodebush-Dixon method,! for comparison with
the known sublimation pressure of ZrF, (ref 2).
The data obtained, as plotted in Fig. 2.3.1 and
given in Table 2.3.1, fit the equation

12,011
T (°K) °

As may be seen from Fig. 2.3.1, the sublimation
pressure of ZrF , is approximately three times that
of HfF, at o given temperature. This factor of
3 difference between the vapor pressures of these

Iogrp (mm Hg) = 12,631 -

 

UNCLASSIFIED
ORNL—~LR~-DWG 25923

TEMPERATURE (°C)

915 900 850 800 750
1000

800
600

400

200

100
80

60

40

PRESSURE (mm Hg)

20

10
8

6

 

a :
084 086 088 080 092 094 096 098 (.00

10/T (°K)

Fig. 2.3.1. Sublimation Pressures of HfF‘ and ZeF .

144

two fluorides could possibly be used as @ basis
for separating zirconium from hafium in a high-
temperature (above 925°C) process. It also might
be useful in the production of ZrF, from a starting
mixture of zirconium and hafnium oxides. The
oxides would be hydrofluorinated and ZrF, would
be separated by heating ond preferential subliming.

"YAPOR PRESSURE OF Ber
S. Cantor

An early determination of the vapor pressures of
BeF, was reported by Moore, but the scatter in
the measurements was larger thon is now expected
with improved techniques. Therefore new meas-
urements were made by using the Rodebush-Dixon
method. The values obtained, as given in Table
2.3.2, fit the equation

10,967
T (°K) ~

 

log p (mm Hg) = 10.487 -

The vapor pressures of BeF, in the range 800 to

1025°C have also been measured by Sense et al, %>
by using the carrier gas method and assuming that
the vapor species was monomeric. Their most
recent results? are in very good agreement with the
data given in Table 23.2. Comparisons of three
quantities derived from the vapor pressures are
presented in Table 2.3.3,

The agreement between the values obtained by

the direct and the carrier-gas measurements indi-

cates that BeF, vapor is monomeric, Accordingly
it should be possible to measure activities from
vapor pressures over BeF,-rich solutions, where
the vapor could be proved to be predominantly
BeF,. A svitable system for an investigation of
this type is LiF-BeF ..

1 « H. Rodebush and A. L. Dixon, Phys. Rev. 26,

W
851 (1925).

2
S. Cantor, ANP Quar. Prog. Rep. June 10, 1956,
ORNL-2106, p 113. Q & Rep J

3R. E. Moore, ANP Quar. Prog. Rep. Sept. 10, 1952,

 

"ORNL-1294, p 151.

4K. A. Sense, M. J. Snyder, and J. W. Clegg, J. Phys.
Chem. 58, 223 (1954).

K. A. Sense, R. W. Stone, and R. B. Filbert, Jr.,
Vapor Pressure and Equilibrium Studies of the Sodium
fgg%ide-Beryllium Fluoride System, BMI-1186 (May 27,

 
 

 

”n

n

Table 2.3.1. Vapor Pressures of HF,

PERIOD ENDING SEPTEMBER 30, 1957

 

Pressure {mm Hg)

Table 2,3.2, Vapor Pressures of BeF,

 

 

 

 

 

 

Temperature Temperature Pressure (mm Hg)
(?C) -Observed Calculated (°C) Observed Calculated
730.9 4.65 4.54 872.4 8.05 8.20
- 739.3 5,65 5.70 896.7 13.0 13.0
754.4 8.30 8.59. 912.5 17.1 17.3
755.5 8.75 8.81 917.6 19.2 18.9
764.1 10,9 1.0 936.9 26.5 26.6
775.9 14.5 46 " 958,5 38.4 38.3
778.0 153 15.7 ' 978.1 53.7 52,8
786.8 19,4 19.6 982.4 57.4 56.6
798.8 25.5 26.5 1000 76.2 74.8
806.6 32 318 : 1009 86.8 85.9
808.3 - 328 33.2 1024 - 108.6 107.9
818.4 - 43,1 42,1 1026 112.5 111.2
827.8 50.6 522 1038 133.4 132.4
830.7 554 560 1050 156.3 158.1
8319 56.4 515 - 1081 182.9 | 180.7
839.3 65.6 679 1070 210.9 209.9
840.2 - 67.3 69.5 - 1081 245.8 244.9
849.1 827 - 847 . 1088 267.5 269.8
859.5 1097 1062 1099 305.2 3119
8647 _120.2 . e
881.1 - - 206.8 1954 -
310 3027

- 909.7

 

' Tnb].e‘ 2.3.3. Com-pdflso_n of Quantities Derived from Ber Vapor Pressures

 

From Work of-’Sen_sE et al.

This Work

 

Vap_or pr:ejs‘su':re' 'eé;udfion_ e

 ‘Heat of vaporizaflon S

) Ex'l'ropolofed bolilng polni

- 10 943
| _Iogp = 10.466 -——T— -

50 1 kcal/mole o
| "_-_71170°c S

~ logp = 10.487 -

10,967

 

| :50.2 kcal/mole )

‘!169°C

 

145

 
 

 

ANP PROJECT PROGRESS REPORT

2.4. PRODUCTION OF PURIFIED MIXTURES

J. P. Blakely

PREPARATION OF VARIOUS
SPECIAL FLUORIDES

B. J. Sturm
Preparation of YF,

The amount of YF, required by the ORNL Metal-
lurgy Division in the experimental production of
ytrium metal has continved to increase. Plans
are being considered for changing the method of
preparing YF; from the wet process to the direct
hydrofluorination of Y,0, in order to meet the
increasing demands.

Approximately 6000 g of YF, was prepared by
dissolving Y,0, in aqueous HCI solution, filter-
ing, precipitating the fluoride by the addition of
aqueous HF (48%), washing by decantation, re-
covering the solid by centrifugation, and drying at
150°C. The partially dried material was given the
standard hydrogen=hydrogen fluoride treatment
following the addition of the desired quantities of
MgF, and LiF. Approximately 8000 g of the
processed YF,;-MgF,-LiF mixture was unioaded
from the receivers, crushed to a suitable size, and
loaded into charging vessels supplied by the
Metaliurgy Division.

Preparation of Other Fluorides

Approximately 5 Ib of CrF, was prepared by
treating hydrated chromic fluoride with NH,F.-HF
and decomposing the resulting (NH,),CrF, at
450°C. Most of the CrF; was converted to CrF,
by hydrogen reduction at 700 to 800°C. A silver
reactor was used in a recent hydrogen reduction
run, but the resulting CrF, has not been analyzed.
The complete dissolution of CrF, frequently is

G. J. Nessle

L. G. Overholser

difficult to obtain, and, as a result, the analytical
values are not as reliable as desired. A suitable
flux that will not oxidize CrF, but will render it
more readily soluble is being sought. One run, in
which equimolar quantities of ZrF, and CrF,
were fused at 900°C in a sealed capsule, yielded
a product that was readily soluble. ' Additional
work is necessary before the stability of GF, in
this medium can be established.

A small batch of VF, was prepared by con-
densing HF on ¢ V,0, powder, heating to remove
the water and excess HF, and then hydrofluorinat-

ing at 200°C.

PRODucnoN",jA-»_ND DISPENSING OF FUSED
SALTS AND LIQUID METALS

Experimental Batch Production
C. R. Croft - J. Truitt

Experimental batches of various compositions
were prepared for use inphase equilibrium studies,
physical properties testing, and corrosion testing.
A total of 442 kg of halide salts was prepared in
41 batches. Seventeen of the 41 batches were
NoF-KF-LiF-UF, (11.2-41-45.3-2.5 mole %, fuel
107) for use by the Metallurgy Division in studies
of nickel-molybdenum base alloys. These 17
batches averaged 20 kg each.

Conversion of 10 Ib of low-zirconium HfCl, to
HfF, was completed with a total recovery of
96.3%. The total input was 4300 g; the total

_output was 3288 g; and the theoretical output was

3415 g. Chemical analyses of the three batches
in which the material was converted are presented

in Table 2.4.1.

Table 2.4.1. Chemical Analyses of Three Batches of HfF ; Prepared from Low-Zirconium HfCI,

 

 

 

 

Botch Major Constituents (wt %) Contaminants {(ppm)

Number Hf F Cl Ni ‘ Cr Fe
1 68.7 30.0 0.30 175 115 1295
2 69.4 29.4 0.27 175 160 1225
3 69.3 29.5 0.29 125 175 1270

 

146

]

s

 
 

 

 

 

 

0

Four batches of LiCI-KCl were prepared for
corrosion and physical properties testing, Three
batches of YF,-LiF-MgF, and seven batches of
KF-BeF ,-ZrF , were prepared for chemical studies,
In addition, batches of BeF, and NaF.BeF, were
prepared by over-the-surface treatment (non-

bubbling) with HF.

Dispensing and Servicing Opgidfiqns
F. A. Doss D. C. Wood

During the quarter, 100 filling and draining op-
erations were performed that involved 785 kg of
liquid metals and 1229 kg of salts. The 1229 kg
of salts was made up of 419 kg of new material
and 810 kg of salvaged material. The salvaged
material is being used in testing of the fluoride-
volatility process for the recovery of uranium.
This salvaged material is not included in the mo-
terial balance given in the following section.

A series of experiments for determining the
amount of fuel which will be held up on various
internal surfaces of the ART was initiated. The
surfaces to be studied are various sizes of tubing
held at various angles to the horizontal, machined
flat plates held at various angles to the horizontal,
a section of the heat-exchanger tube-to-header
junction, a mockup of a tube bundle, a 3-in.-dia
bellows, and various right-angle welded joints.
The pieces being studied are appropriately sup-
ported in a small flange-top receiver, the receiver
is filled to immerse the piece in NaF-ZrF ,-UF,
(56-39-5 mole %, fuel 70) at 1500°C for 2 hr, the
temperature is allowed to drop to 1200°F (the
draining temperature of the ART), and the receiver

"PERIOD ENDING SEPTEMBER 30, 1957

is emptied through a dip line to simulate draining.
Tests have been completed on five of the 17 pieces
to be examined.

Materials Available
J. E. Eorgan F. A. Doss D. C. Wood

“Upon termination of operation of the production-

scale facility at the end of June 1957, the mo-

terials on hand consisted of 11,559 Ib of NaF.ZrF ,,
1810 Ib of low-hafnium ZrF,, 1685 Ib of normal
ZrF,, 200 Ib of NaF, two new 58-in. reaction
vessels, two reworked 48-in. reaction vessels,
one new furnace liner, and seven new storage
containers. The higher-than-expected amount of
NaF.ZrF, on hand resulted from the receipt of
10,840 ib from Kawecki Chemical Company, who
took advantage of the 4000-1b overage allowed in
a purchase order for 30,000 Ib of material.

The types and emounts of purified fluoride mix-
tures on hand on July 30, 1957, are listed below:

‘Composi- Amount
tion No. Type (kg)
30 NaF-ZrF ,-UF 4 (50-46-4 mole %) 5860
31 NaF-ZrF, (50-50 mole %) 933
70 NaF-ZrF ,-UF 4 (56-39-5 mole %) 900
107 NaF-KF-LiF-UF, (11.2-41-45.3-2.5 119

mole %)
VYarious special mixtures 552
Total '8_3-6:4—

147

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

2.5. ANALYTICAL CHEMISTRY
J. C. White

DETERMINATION OF TRACES OF NaK IN AIR
A. S. Meyer, Jr.  J. P. Young

Additional modifications were made in the optical

system of the instrument for the detection of NaK

in air by measurement of the absorption of sodium
resonance radiation by sodium atoms.! These
modifications, which were first discussed in
the previous report in this series,? include the
insertion of an additional light stop which is
designed to exclude the continuum radiation of
the heated furnace from the photomultiplier tube.
The continuum radiation caused a near-saturation
phenomenon with respect to the light seen by
the phototube and thereby severely reduced the
sensitivity of the instrument when the temperature
of the furnace exceeded 700°C.

It was suggested® that the light stop consist
of a plate with two small apertures approximately
0.08 in. in diometer. The aperture plate (shown
in Fig. 2.5.1) is placed in the focal plane of

 

1A, s. Meyer, Jr., and J. P. Young, ANP Quar. Prog.
Rep. March 31, 1957, ORNL-2274, p 137.

2. s, Meyer, Jr., and J. P. Young, ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340, p 166.

3G. K. Werner, private communication to A. S. Meyer, Jr,

INLET AIR
SAMPLE

FURNACE
QUARTZ WINDOWS

the image of the quartz end windows of the ab-
sorption cells; the image is formed by the con-
densing lens. The two circular holes in this
plate correspond precisely to the images of the
end windows. With this arrangement, all radiation
except that from the heated quartz windows
is masked by the light stop. An additional
condensing lens is required to refocus the beams
on the photocell. The positions of both the
aperture plate and the small. condensing lens
can  be adjusted along the axis of the optical
system to provide sharp focusing. A multilayer
interference filter and several possible com-
binations of glass and gelatin color filters have
also been included to isolate more effectively
the resonance radiation at 589 my.

Since the openings in the added cperture plate
are relatively small, much more precise focusing
and alignment of the optical system are necessary
than were required for the original system. Ac-
cordingly, o base assembly is being fabricated
which provides for precise adjustments of the
positions of both the chopper and detector com-
ponents. The arrangement is designed according
to the principles delineated by Strong,4 and it

 

4J Strong, Procedures in Experimental Physics,
p 585-592, Prentice-Hall, New York, 1938.

UNCLASSIFIED
ORNL-LR-DWG 25924

ABSORPTION CELLS

 

SODIUM VAPOR LAMP

 

 

INTERFERENCE FILTER

 

 

APERTURE V 7 //// // //4 APERTURE~

 

 

 

 

 

/
@ Z:f_"___f__

 

 

 

 

 

 

 

 

ANALYZER
e FURNACE
CHOPPER EFFLUENT
ASSEMBLY AIR SAMPLE

QUARTZ WINDOWS

 

 

—y—

DETECTOR
ASSEMBLY

Fig. 2.5.1. Instrument for the Detection of Submicrogrom Quantities of NaK in Air.

148

of

 
 

 

=}

L

@

 

provides for movement of the optical components
in any desired direction with a single degree of
freedom. All the components are mounted on -a
7-ft section of 18-in. channel iron, which serves
as a rigid optical bench. Each end of the channel
iron is -slotted to accommodate bushings which
locate the base plates for the chopper and detector
assemblies. These base plates can be moved
along the optical axis, for coarse focusing, or
can be rotated about a vertical axis. A second
plate is supported on each base plate by ball
bearings in opposing V.grooves. The second
plate is spring-loaded with respect to the base
plate and is driven by o screw-drive mechanism
to provide motion at right angles to the optical
axis of the system. The chopper and detector
sections are each supported by three vertically
adjustable - screws whose spherical ends rest
in radial V-grooves in the second plates. Ad-
justments of these screws can produce tilting
motions -in any desired direction with negligible
translational motion. Fabrication and gssembly of
the modified instrument are about 90% completed.

DETERMINATION OF OXYGEN IN
METALLIC LITHIUM

A. S. Meyer, Jr. 'R. E. Feathers

Further studies of the butyl iodide method for
the determination of oxygen in metallic lithium
have revealed that several unexpected reactions
occur during the dissolution of the sample in

"ethereal solutions of iodine and butyl iodide

and the subsequent extraction of the inorganic
compounds  into -aqueous. solutions for titration.

It was demonstrated that oxygen, “when present
‘as LiOH, cannot be. defermmed by this. method.
No ucud was consumed in the titration ‘of the .
‘aqueous extract of- the soluhon which was formed‘ :
“ by the reactlon of 500 mg of LiOH with the
ethereal reagents. The following reaction of ._L!OH
with iodine is proposed to explain the elimination

of the hydrox:de durlng the dissolution reaction:

L|0H + 1, -—>L|I + HOI

The hypo:odous acud is oo weak for tltrcmon cnd'

is probably decomposed to -water, iodine, and

oxygen in the nonaqueous medium. The similar - -

reaction which may occur with Li,0,

Li20 + I2~—-—>Li0| + Lil

PERIOD ENDING SEPTEMBER 30, 1957

does not interfere with the determination of oxygen
as Li,0. In the titration with acid, the hypoiodite
consumes a quantity of acid which is equivalent
to that required to titrate the oxide from which
it is formed:

- - L4 '
oI + I~ + 2H 7——>I2+H2O

The butyl iodide method appears therefore to
be essentially specific for the determination of

. 'Li20' The elimination of the alkalinity of Li 3N

and Li,C, by reaction with iodine was dlscussed
prevuously. ~ Lithium carbonate is the only re-
maining alkaline constituent which might be
titrated. It is unlikely, however, that Li,CO,
will be present in sngnlflcant concentrahons m
samples which have been heated to elevated
temperatures, since it is not compatible with
metallic lithium.

A small negative bias has been observed in
determinations by the butyl iodide method that
is believed to be caused by traces of acid which
are liberated by the reaction

H20 + Iz-% HI + HIO

which occurs during the extraction of the ethereal
solution with water. The equilibrium constant
for this reaction has been reported® to be approxi-
mately 10~13,  On the basis of this constant,
less than one microequivalent of acid would be
liberated by this reaction under the conditions
of the titration. When ether is present as a third
phase, however, the equilibrium is displaced to
the right and hydrogen ions are liberated in

‘quantities - equivalent to 50 to 100 ppm for a
~1-g sample. The magnitude of this error is a
“function of the amount of excess iodine which

remains  after the dissolution of the sample and

_'fhe'refore 'is not easily corrected for by blank
~ determinations. The only explanation which has

been postulated for the increase in the hydrolysis

" reaction ‘is that hypoiodous acid is complexed

or preferenhally extracted to a high degree by

 ether.

The error can be eliminated by removing the

'free -iodine before the organic solution is ex-
“tracted with water. A distillation method for the

 

SA. S. Meyer, Jr., et al., ANP Quar. Prog. Rep.

Sept. 10, 1956, ORNL-2157, p 127.

SN. V. Sidgwick, The Chemical Eléménts and Their
Compounds, vol 1l, p 1213, Clarendon Press, Oxford,
1950.

149

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

removal of the iodine and the ether was tested
and was found to be impractical. The iodine
was rapidly removed, however, by reduction with
finely divided silver or with mercury. Mercury
was found to be the more convenient reagent,
since an exfensive preparation procedure must
be carried out to obtain silver in o sufficiently
finely divided form for ropid reaction. The effec-
tiveness of this procedure in eliminating the
negative interference was demonstrated by titration

of reagent blanks end by carrying 500-1g sarnples,
of Li, O through -the procedure. - Approxlmutely-'
80% of the odded Li, 0 was titrated. . In view.
of the possible confumlnaflon with moisture during”

transfer of small quantities of Li,O, a recovery
of 80% was considered to be satisfactory.

in the presence of iodide, the excess silver
or mercury is oxidized rapidly by atmospheric
oxygen, These reactions,’ indicated in ‘the fol-
lowing equations, can consume hydrogen’ ions
during titration and introduce: posmve errors:

4Ag + 41 + O2 + 4H* ——>4Ag| + 2H20

2Hg + 81~ + 0, + 4H*-.->2Hg| == + 2H,0

The positive - errors can. of course be - avoxded"

by excluding. oxygen durmg fhc ‘extraction of

The errors can be eliminated more convemently,
however, by the addition of enhydrous HgCI
after the reduction of the iodine. The HgCI,
complexes the iodide according to the equcfion

217 + HgCl,—> Hgl, + 2Cl~

The extraction and titration are then, in effect,
carried out in a chloride system,

A proposed procedure based on the considerations
discussed above is now being tested on samples
of metallic lithium. The dissolution is carried
out in the opparatus shown in Fig. 2.5.2. With
the cooling jacket of the condenser drained and
an empty reaction vessel in position, the con-
denser is flushed by transferring 25 ml of anhy-
drous ether to the reaction flask. lodine {15 g)
and a magnetic stirring bar are transferred to
a second reaction vessel immediately after it
is removed from a drying oven, and, with a
positive pressure of helium in the condenser,
the reaction vessel containing the iodine is

substituted for that containing the flushing.

150

solution, The system is then evacuated and
purged with helium; ond a 100-ml wvolume of
n-butyl iodide, which is stored over activated
aluming, is transferred to the reaction vessel,
along with 100 ml of diethyl ether, which is
stored over metallic sodium. The solution is
then stirred until the iodine is mixed thoroughly.

The sample of 1 to 2 g of metallic_ lithium is
added through the condenser from a Jomesbury

“valve sampler.” The solution is heated at reflux
. temperatures and stirred until dissolution of the

metal is completed. An excess of iodine should

be evident at this time. A 10-ml aliquot of the

organic solution is withdrawn, and the con-
centration of iodide in the solution is determined
by titration with AgNO,.
sample is calculated from this titration. A
10-ml portion of mercury is then added, and the
solution is stirred until the color of the iodine

is discharged. - A 30-g quentity of anhydrous

HgCl, is added,” and the solution in the flask
is snrred for 5 min to permit complexing of the
iodide ion. The solution is transferred from the

flask to o separotory funnel and extracted with

recently boiled distilled water which has been

pretitrated to a pH of approximately 5. The
. aqueous extract is then 'ritroted with 0.01 or

- 0.002 N'HCI.

the organic solution .and . dunng _the fitrgtion.

There has, as yet, . been no opportunlty to test
this modification of the. procedure by analyzing

'hlghly purified samples of metallic lithium.

Excellent recovery was attained when a 0.5-mg

~sample of Li,O was added to a simulated sample

solution which was prepared from anhydrous
Lil, butyl iodide, ether, and iodine. As soon
as a sufficiently reproducible series of samples
is available the results of determinations by
this procedure will be compared with unolyses
carried out by other methods

DETERMINATION OF OXYGEN IN.FLUORIDE
SALTS AND IN METALS
A. S. Meyer, Jr.  G. Goldberg

The apparatus for the determination of oxygen
as oxides in fluoride salts and in metals by the

 

7A. s, Meyer, Jr., R. E. Feathers, and G, Goldberg,
Ahlli' Quar. Prog. Rep. March 31, 1957, ORNL.-2274,
p

BA, S. Meyer, Jr., G. Goldberg, and B. L. McDowell,
ANP Quar. Prog. Rep. June 30, 1957, ORNL-2340,

- p 167

The weight of the

u}

 
- GLASS-TO-METAL SWAGELOK FITTING

- PRESSURE R ELl EF
BELLOWS VALVE

  
  
  
  
 
 

HELIUM ——==

 

'ETHER

WATER-COOLED JACKET—

 

"~ SERUM

 

. Blliter
FLASK -

 

FRITTED FILTER

   
 

STOPPER

   

_fi

UNCLASSIFIED
ORNL—LR—DWG 25925

SAMPLE PORT FITTED TO JAMESBURY-VALVE TRANSFER CHAMBER

INLET FOR HELIUM, VACUUM

 

14Y5 in.

   
 

 

 

r\l-
wfijw\

sTO PPER‘—\

™~ WATER- q

SERUM

COOLED

 

 

 

 

|

 

 

 

 

CONDENSER -

BUTYL
IODIDE

REACTION VESSEL

™~—MAGNETIC STIRRER

FLASK

 

 

 
 
 
   

  
   
  

  
  

3-liter

FRITTED FILTER

Fig. 2.5.2. Apporatus for the Determination of Oxygen in Metallic Lithium,

1S1

£561 0 ¥IGWILJ3IS ONIANI Q0Id3d

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

fluorination of the oxides with KBrF, was as-
sembled. This apparatus, which is adapted from

the work of Hoekstra and Katz,? is shown in

Fig. 2.5.3.

A detciled procedure was formulated and is

presently being given an initial test. Briefly,
the reagent, KBrF,, is formed by the addition
of BrF; to KF in the reaction tube. Sufficient

KBrF, is made to permit the andlysis of a number
of samples without. recharging the system with
fresh reagent for each determination. The reaction

tube is cooled to the temperature of liquid
nitrogen, and, with argon flowing into the tube,
the sample is quickly added to the tube, which
is then caopped. The temperature of the reactor

is then raised to operating temperatures, about -
400°C, so that the KBrF, can react quanmotlve!y‘
with the metallic oxides present and liberate oxy-

gen gas. This gas.is passed through @ series of

traps to remove bromine and excess recgent. "An *

automatic Toepler pump is used to transfer the
oxygen fto either of two pressure-measuring de-
vices - a mercury manometer or, if greater sen-
sitivity is desired, an oil manometer. .

Since the volume of the measuring system cannot

be obtained directly, the system must be calibrated

by carrying out the procedure with samples of

known oxygen content. Zirconium oxide has been

selected as a standard for calibration. On the

basis of estimated wvolumes of the measuring
system, quantities of oxygen as low as 5 pg

can be detected. The range of the apparatus caon
be extended upward by severdl orders of magnitude
by increasing the volume of the measuring system.

DETERMINATION OF NICKEL IN
METALLIC SODIUM

A. S. Meyer, Jr.  B. L. McDowell

The method for the determination of nickel in
which the absorbance of its chelate complex with
4-isopropyl-1,2-cyclohexanedionedioxime in xy-
lene'® is measured was applied to -onalyses of
samples from studies which are being carried
out by the Metallurgy Division in order to deter-
mine the solubility of nickel in molten sodium.

 

4. R. Hoekstra and J. J. Katz, Anal. Chem. 25, 1608

 (1953).

104, 5. Meyer, Jr., and B. L. McDowell, ANP. Quar ‘
“""is converted to a compression fitting by facing

Prog ‘Rep. June 30, 1957, ORNL-2340, p 168."

152

The samples are prepared by heating metallic
sodium under an atmosphere of helium in a nickel
crucible which is contained in a welded Inconel
capsule. After the sample hds been maintained
at the desired temperature for a period sufficient
to establish equilibrium, the molten metal is
transferred to an opposing molybdenum bucket
by inverting the Inconel capsule. The sample
is submitted for analysis in the molybdenum

bucket.. The sample includes the contents of .

the bucket together with ‘any nickel deposrfed
on the inner walls of the bucket.

- Initial determinations were carried out by dis-
solving the metallic sodium in 50 m! of methanol.
An aliquot of the methanol solution was titrated
with standard acid to determine the sample weight.

The moiybdenum ‘bucket was -rinsed with warm

1:3 HCI, ond the washings were combined with

the methanol solution, which was acidified and -
‘evaporated toa volume of approximately 25 ml

to remove the methanol.

In the determinations, an unusual interference by
molybdenum, which was dissolved from the con-
tainer, -was experienced. It was found that molyb-

~denum -(an orange-colored extract) .interfered only
when methanol or some contaminant thereof was

present and when cyanide and Sulfi-Down, added

_as -masking agents, had been used. No colored

extract was observed when the chelating agent,
4-isopropyl-1,2-cyclohexanedionedioxime, was ab-
sent, and the interference appeared to be in-
tensified by the presence of nickel. Since the
interfering reactions appeared to be exceedingly
complex and could be eliminated by dissolving
the samples in water rather than in methanol,
the nature of the reactions has not yet been
studied in detail.

Because the outer walls of the molybdenum
bucket may be contaminated with nickel as a
result of contact with the Inconel capsule, it is
necessary to exclude from the sample the water
used for the washing of the outer surfaces of
the bucket. This requirement presented special
problems when the samples were dissolved in
water and required the design of the new ap-
paratys, fabricated from fiuorothene, which is
shown in Fig. 2.5.4. The apparatus is con-
structed by threading a modified ‘/2-ih. flared

. fluorothene fitting through the bottom of a 250-ml
' heavy-walled fluorothene beaker, The flared fitting

‘l
gsl

VACUUM GAGE

 

 

UNCLASSIFIED
ORNL—LR—-DWG 25926

 

 

VACUUM

. le 53

 

 

 

 

PLvdb ety rr by rrrrer

   

 

 

 

MANOMETER

ey

o=t

I' &

  
       
   

 

oI ManoMETER

1P

 
  

 

 

 

 

   
  

i

hs;

S
N /
blFFERENTlAL
MANOMETER

 

 

 

 

N TOEPLER PUMP

  

MANIFOLD

GLASS-TO-METAL SEAL

 

 

 

 

V—INCONEL DIAPHRAGM VALVES (HOKE, INC)
S—HIGH-VACUUM STOPCOCKS

   

 

 

   

 

 

'. Flg. 25.3. Apparatus for the Determination of Oxygen In Fluoride Salts and in Metals,

=0

 

{?S.

 

 

PRESSURE
REGULATOR
FOR HELJUM SUPPLY

(¥ 24,

PYREX TRAP

LS61 ‘0c ¥IIWILHIS ONIANI goIyad

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 
  
  
   
  

BRASS NUT

UNCLASSIFIED
ORNL—~LR-~DWG 25927

INERT GAS INLET

P RuBBER STOPPER

FLUOROTHENE BEAKER

MODIFIED FLARE FITTING

RUBBER SLEEVE
RETAINING RING

MOLYBDENUM BUCKET APPROXIMATELY ‘/3 FULL OF SODIUM

Fig. 2.5.4. Apparatus for Dissolution of Sodium Samples in Water,

off approximm‘el¥ ‘/8 in. of the tapered end ond
turning into it a /2-in.-!D recess 7 in. deep. The
molybdenum bucket is held in place in this recess
and sealed by a Y-in. section of 3/8-in. rubber
tubing. The rubber is backed by o brass ring
(back ferrule of a Swagelok fitting) and compressed
by the brass nut of a flared fitting. The apparatus
can easily be modified to accommodate cylindrical
containers of smaller diameters.

In the dissolution of the samples the apporatus
is loosely covered with a rubber stopper which
ig fitted with on inlet for inert gas. An inert
gas is passed ropidly through the fitting for
approximately 30 sec, after which the sample
bucket -is inserted and locked into position.
After the bucket is cooled in an ice water bath,
approximately 25 g of ice, which is prepared by
freezing demineralized water in plastic tubes,
is added to the becker. While the flow of inert
gas is continued, the stopper is removed suffi-
ciently for water to be added cautiously until

154

the sample storts to react. The beaker is again
covered, and an atmosphere of inert gas is main-
tained until the dissolution is completed. Approxi-
mately 5 min is required to complete the reaction.

FORMATION OF CARBIDES BY THE REACTION

OF PUMP LUBRICANTS AND NaK AT
ELEVATED TEMPERATURES

A. S. Meyer, Jr.  G. Goldberg

During the dissolution of alkali metdls and
alkali metal residues, an odor that is commonly
associated with acetylene has frequently been
detected. This phenomenon has been most pro-
nounced during the chemical examination of cold
traps by the method described previously,'!
When water was added to cold traps from NaK
systems after most of the residual alkali metals

 

A, s. Meyer, Jr., and G. Goldberg, ANP Quar. Prog.
Rep. June 10, 1956, ORNL-2106, p 128.

 
 

 

 

 

 

"

had been converted to the halides by reacting
them with butyl bromide, an intense odor was
detected and minor explosions were experienced,
even though an attempt was made to maintain an
inert atmosphere within the system. The presence
of acetylene in the evolved gases was sub-
stantiated when precipitates of Ag,C, were formed
upon passage of the gases through solutions of
AgNO

Recenfly an explosion occurred when an ex-
perimental NaK system was opened after operation
had been terminated because of a leak in the
pump seals. In view of the well-known acetylene
explosion hazard, it was deemed necessary to
investigate the possibility that carbides had
formed as a result of a reaction between the
pump lubricant and the alkali metals, with the
subsequent liberation of acetylene.

In earlier tests of the compatibility of lubricants
with molten alkali metals,!? there was no ob-
servable reaction between the alkali metals and
the lubricants. The tests were carried out, how-
ever, to determine whether a reaction would occur
if alkali metals at 1400°F leaked into the {ubricant
at 400°F. The conditions of these tests did not
approximate, therefore, the temperatures which
would prevail during a leak of the pump oil into
the NaK being pumped.

In order to test for the formation of carbides,
the reaction was investigated in the apparatus
shown in Fig. 2.5.5. The reaction tube is «
section of 3/ -in.-OD, 0.065-in.-wall Inconel tubing,
which can be heated in a vertical tube furnace.
A thermocouple is spot-welded to the tubing.
The upper section of the apparatus, which is
fitted with a cooling coil, provides an expansion

chamber and reflux surfaces for the lubricant

and prevents the formohon of dangerous pressures,
A pressure gage is fitted fo the ‘/4-|n. Swagelok

- fitting.- A bellows valve is fitted to the 3/8-|n.'

Swagelok fitting after the reochon components

are added to the apparatus. =
For the test,2 ml of NaK and 2 ml of Gulfspln-35

lubricant were added ‘to the reactor, “which was

“then evocuuted ond purged with helivm. . The
bellows valve was. closed and the lower section
of the reaction tube was heated to a temperature

of 650°C. No. sugmhca_nt increase in pressure

 

12G, Goldberg, A. S. Meyer, Jr., and J. C. White,
Compatibility of Pump Lubricants with Alkali Metals
and Molten Fluoride Salts, ORNL-2168 (Dec. 28, 1956).

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR-DWG 25928

a—3/g-in. SWAGELOK FITTING
Yfa=in, SWAGELOK FITTING ~* &

COPPER COQOLING COIL. e
_———

 

) 2in
=L
)\|~ ri z ‘
REACTION TUBE ———=
4in.

 

 

 

 

 

Fig. 25.5. Appoaratus for Sfudying the Reaction of
NaK with Pump Lubricants.

was observed wuntil the temperature reached
approximately 600°C, at which time the pressure
increased abruptly to 450 psi. The gases were
later analyzed mass spectrographically and found
to consist primarily of methane, with a trace of
hydrogen. After the reaction tube had been main-
tained at a temperature of 650°C for 4 hr, it
was cooled and frozen with dry ice. A 3-in.
section of the tubing was then cut off and placed

"in‘ a stainless steel bomb which contained 200 ml
- of frozen water. The bomb was sealed and allowed

to warm to room temperature. When the reaction
was completed, the ewvolved gases were passed

'through a 1.5 M solution of AgClO,, according’
-to the procedure which was developed for the

determmcmon of Li C2 in metallic lithium.13

Copious quantities of Agzc were precipitated.
The acetylene present was not determined quan-
titatively because part of the precipitate was
lost when a portion of the Ag, C, exploded. It
is estimated that approxumcrfely 2% of the alkali

 

137, W. Gilbert, Jr., A. S. Meyer, Jr., and J. C. White,
Spectrophotometric Determination of Lithium Carbide in
Metallic Lithium as the Acetylene-Silver Perchlorate
Complex, ORNL CF-56-12-111 (Dec. 27, 1956).

155

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

metal was present as the carbide. A heavy
deposit of carbonaceous material remained in
the reaction tube. Negligible quantities of
carbides were found in a blank run in which the
test was repeated with only NoK added to the
reaction tube.

The test is being repeated in order to obtain
quantitative measurements of the carbides, and
tests with sodium are proposed.

DETERMINATION OF ALUMINUM IN MIXTURES
OF FLUORIDE SALTS

J. P. Young J.R. French

An investigation of a spectrophotometric method
for the determination of trace amounts of
aluminum' in mixtures of fluoride salts with
pyrocatechol violet was continued. The repro-
ducibility of the data was improved by washing
all glassware with ¢ dilute solution of pyrocatechol
violet prior to use and by the addition of thiogly-
colic acid to the test solutions. These modifi-
cations lowered the coefficient of variation of
the method to approximately 2%.

A study was made of interferences that would
normally be encountered in corrosion tests with
NaF-KF-LiF-UF, and the newly developed nickel-
molybdenum-base alloys. These alloys contain
trace amounts of iron plus a number of nonferrous
materials. Since pyrocatechol violet is a sensitive
reagent for iron, essentially complete removal of
iron is necessary before application of the reagent
in the determination of aluminum. Removal of the
iron is best accomplished by means of the mercury
cathode. The interference by titanium is partially
eliminated by measuring the absorbance of the
solution at both 582 and 700 mp and correcting
for the contribution at 582 mu from ftitanium.
The simultaneous determination of titanium and
aluminum in the scme test portion appears to
be possible and is to be the subject of further
work. Nickel, chromium, molybdenum, and niobium
do not interfere in determinations by this method.

Initial results obtained in the determination
of aluminum in synthetic solutions of samples
from corrosion tests were about 10% high. These
results indicate possible accumulative inter-
ferences, ond therefore further work is planned
to improve the reliability of the determinations.

 

14, p. Young and J. R. French, ANP Quar. Prog.

" Rep. June 30, 1957, ORNL-2340, p 168.

156

SOLVENT EXTRACTION OF TITANIUM AND
NIOBIUM FROM ACIDIC SOLUTIONS WITH
TRI-z-OCTYLPHOSPHINE OXIDE

Studies of the application of tri-n-octylphosphine
oxide (TOPO) in the solvent extraction of metals
of interest in the ANP program were continued.
Since sulfuric acid is commonly used in the dis-
solution of fluoride salts, the extraction char-
acteristics of sulfuric acid by TOPO were also
investigated.

Extraction of Sulfuric Acid
T. W. Gilbert

Solutions of 1 to 8 M sulfuric acid were ex-
tracted with 0.1 M tri-n-octylphosphine oxide
(TOPO) in cyclohexone. The quontity of sulfuric
acid in the organic phase was determined directly
by titration with alcoholic potassium hydroxide
ond also by back-extraction of the organic phase
with water. The results were the same by either
method. The quantity of sulfuric acid extracted
was found to increase almost linearly with in-
creasing concentration of sulfuric acid. From
7.0 M acid, 0.77 mmole of sulfuric acid was
extracted by 1.0 mmole of TOPO. A third phase
formed when 8.0 M acid was extracted. The
third phase was analyzed for total acid and for
phosphorus. The ratio of H,50, to TOPO obtained
was 1.05.

Extraction of Titanium
T. W. Gilbert

Tri-n-octylphosphine oxide (TOPO) dissolved in
cyclohexane has proved to be an efficient ex-
tractant for titanium from solutions of sulfuric
acid. The efficiency of extraction was found to
be highly dependent on the concentration of
sulfuric acid in the aqueous phase. In the ex-
traction of 0.16 mmole of titanium with 1.0 mmole
of TOPO, only 6% of the titanium was extracted
from 1 M sulfuric acid. From 7 M sulfuric acid,
however, 99% of the titanium was extracted. From
8 M acid, 99% of the titanium was removed from

the aqueous phase, but a third liquid phase was

formed.

The extraction of titanium from solutions 0.5 M
in tartaric acid in the presence of varying amounts
of sulfuric acid was also investigated. The
resuits were very similar to those obtained in

 
 

 

 

 

 

the presence of sulfuric acid alone; only a slight
improvement in the extraction of titanium was

observed. -,
A study of the extracthn of increasing amounts

of titanium from 6 M sulfuric acid with 1.0 mmole
of TOPO showed that saturation of the organic
phase occurred when 0.5 mmole of titanium had
been extracted. This result, when combined
with the results of an acidimetric study of the
sodium fluoride solutions used to strip the
titanium from the organic phase, indicates that
the principal species present in the cyclohexane

phase at saturation is TiOSO,.2TOPO.

Extraction of Niobium
C. A. Horton

Niobium(V) extracts best into TOPO in cyclo-
hexane from hydrochloric acid solution, fairly
well from sulfuric acid, and' almost not ot ali
from nitric or perchloric acid, with approximate
extraction coefficients for approximately 4 M
acid of 200, 20, 0.02, and 0.01, respectively.
For stable acidic solutions of over 0.2 mg of
niobium, the presence of citric, lactic, tartaric,
or oxalic acids or thiocyonate, peroxide, or
catechol are required to keep the niobium in
solution. For 500 pg of niobium in 0.2 M tartaric
acid, maximum extraction is obtained from 4 M
hydrochloric acid, as shown in Fig. 2.5.6. The
complexing agents must be added before the
inorganic acids in order to keep the niobium
in solution. The niobium in the organic layer
can be estimated by adding catechol and butyl
acetate to an aliquot of the extract and measuring
the absorbance at 360 mu. :

INSTALLATION OF A MICROBALANCE
IN A YVACUUM DRYBOX

A. S. Meyér, Jr.  R. E. Feathers

The welghlng compartment of a Cahn Electro-

balance 13 has been installed in a vacuum drybox.
This’ electrobalance can be adjusted for weighing

~ quontities of . the - order of 5, 10, 20, and 50 mg. -

The balance operates on the principle of equating

- the- torque of the sample with an opposmg electro-

magnehc torque.
The components which are mstulled w:fhm the

drybox mclude the weighing compartment, four -

connecting leads, a ‘terminal strip, and the torque

ISCuhn Electrobalance, Model M-10, Cahn Instrument
Company, Downey, Calif.

 

EXTRACTION COEFFICIENT (orbitrary units)

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR-DWG 25929

200

100
80

60

40

20

 

o i 2 3 4 5 & T 8 9 10
CONCENTRATION OF HCI (M)

Fig. 2.5.6. Effect of HCl Concentration on Extraction

of Niobium with Tri-n-octylphosphine Oxide.

motor and balance beam, which are similar in
construction to the movement and pointer of o
galvanometer.  All these components can be
subjected to evacuation without damage and do
not offer any problems of excessive outgassing.
Other components, which cannot be conveniently
evacuated, including batteries, switches, and
Helipot precision resistors, are retained in the
original instrument cabinet outside the drybox.
All  balancing operations are thus carried out
remotely by adjusting the currents to the torque
motor. Electrical connections are effected through
four insulated conductors, which are sealed
through the walls of the drybox.

This balance will be used principaily for the

~preparation of synthetic samples to which trace

quantities of a constituent must be added with-
out exposure to atmospheric contamination or
moisture. It has already proved invaluable in
the preparation of simulated somples for tests
of the proposed method for the determination
of ~oxygen in metallic lithium (see preceding
section of this chapter). In these studies sub-
milligram quantities of Li,O were combined with
20-g samples of anhydrous Lil.

157

 
 

 

Part 3
METALLURGY

1 S¥

 
 

 

B

 

 

 
 

 

 

 

 

3.1. NICKEL-MOLYBDENUM ALLOY DEVELOPMENT STUDIES

MATERIAL DEVELOPMENT

T. K. Roche
J. H. Coobs -

H. Inouye

The developmental work on a new container alloy

for the liquid fuel, NaF-KF-LiF-UF , (11.2-41-45.3-

2.5 mole %, fuel 107), being considered for use in
odvanced. circulating-fuel reactors has been con-

cerned primarily with three major problems. First,
studies have been made for determining the optimum:
alloy composition for service temperatures up to
1800°F. The second problem involves determining
whether the most promising alloy, INOR-8, can be
produced as a ‘‘commercial’’ item. = Since all
nickel-base alloys exhibit mass transfer in sodium
at the service temperatures of interest, the third

problem is the fabrication of duplex tubing of -

INOR-8 clad with stainless steel for use in heat
- exchangers,

The research and developmental studies of the -
alloys had consisted of a-

nickel-molybdenum _
screening program until recently, when sufficient
data had been accumulated to indicate that INOR-8

was the most sat_isfdctory' of the alloys studied .
in terms of the original objectives. The data on
which the selection was based were obtained for -

INOR-8 and seven other INOR-type alloys at tem-
peratures up to 1500°F. Since forced-circulation

loop tests, both in-pile and out-of-pile, have n_bt ‘

been completed, and, in addition, comprehensive
weldability and strength evaluations remain to be

made, two other compositions, INOR-8 modified

and INOR-9, are aiso being studied as possible
alternates if INOR-8 shows excessive corrosion

~attack.

“or poor creep strength at 1800°F, The composi-

tions of INOR-8, INOR-8 modified, and INOR-9 are
given in Table 3.1,1. No data on INOR-8 modified
are as yet available.

-~ Tests of thermal-convection loops fabricated of

INOR-8 in which fuel 107 was circulated showed

‘acceptable corrosion resistance of the alloy at

1500°F, but no tests have been run at 1800°F. It
may be assumed, however, that if excessive cor-
rosion occurs at 1800°F it will be due to the
chromium content of the alloy, since in the tests
at 1500°F the chromium content of the fuel in-
creased with increases in the depth of corrosive
Therefore INOR-9, which contains no
chromium but is strengthened with niobium, is
being studied as a possible substitute in the event
that INOR-8 is not sufficiently corrosion resistant
at 1800°F. On the other hand, INOR-8 may prove
to be satisfactory from the corrosion standpoint
and unsatisfactory in terms of creep strength, and

- therefore the possibility of increasing its creep
‘strength with niobium additions, such as in INOR-8

modified, and with carbon additions is being in-

vestigated.

Although it is known that more than 0,10 wt %
carbon usually prevents the production of tubing
because of splitting along carbide boundaries,
present indications are that more than 0.10 wt %
carbon can be tolerated in the fabrication of plate
and sheet. Thus it is important that the effect of

“the carbon content on the creep strength of these
alloys be determined.
~ content is found to be beneficial, thin-walled tubing

If increasing the carbon

Table 3.1.1. -Comppiitiohs of INOR-Type ‘Alloys- Being Considered as Container Materlals for
Fuel 107 in the Tempercture R&ngg of 1500 to 1800°F -

 

: ;'Allo)"f L

- Nominal ‘Cc‘:mp‘osition.(wt %)

 

 

~ Designation” Mo Cr " Fe  Nb Ni

INOR-8 15-19 68 4-7 0 Balance
~INOR-8 modified 5-19 - 68 4-7 2-4 Balance
INOR-9 4-7 2-5 Balance

12-19 0

 

({89  (Co) 18

 

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

of a suitable carbon content may be produced by
carburizing tubing containing 0.10% or less carbon.

Properties of INOR-8

Physical Properties. — The density of INOR-8
has been determined to be 8.79 g/cm® or 0.317
Ib/in.3, and its modulus of elasticity at room tem-
perature in the annealed condition is 32 x 10% psi.
Data on its thermal expansion over various tem-
perature ranges are given below:

Tempercture Range Expansion
CF)  (pin./in. °F)
212--1832 8.6
212-752 7.0
752-1112 8.4
1112-1800 9.9

Fabricability. — The principal difficulty that has
been experienced in the fabrication of INOR-8 has
been the formation of large shrinkage cavities in
cast ingots. While the tendency for ingots to form
such cavities is inherent in the nature of freezing
of all molten metals, the tendency seems to be
somewhat accentuated in the case of INOR-8. It
is not known at present whether this casting diffi-
culty is attributable to the casting variables, the
expulsion of dissolved gases, or the temperature
range of freezing of the alloy. Irrespective of
their cause, cavities in an ingot can, in the worst
case, cause center bursts during forging, or they
can, in less drastic situations, cause center-line
defects that are detectable by ultrasonic inspec-
tion. Present approaches to a solution of this
problem include changes in casting practice, con-
sumable arc melting, and subjecting the molten
metal to a carbon boil. In all other respects, the
forgeability of the INOR-8 alloy is excellent.

Oxidation Resistance. — The oxidation resistance
of the nickei-molybdenum alloys in air is strongly
influenced by the chromium content. The studies
made thus far have indicated that no chromium is
better than a slight amount and that unless a
critical amount is present the oxidation rate is
not decreased to a level that can be considered to
be acceptable. As the data in Table 3.1.2 indi-
cate, the critical chromium content is approxi-
mately 6 wt %.

. Corrosion Resistance. ~ Tests of thermal-convec-
tion loops fabricated of INOR-8 in which fuel 107

was circulated at @ maximum temperature of 1500°F

162

Table 3.1.2. Oxidation of Nickel-Molybdenum (17 wt %)
Alloys at 1800°F as a Function of the Chromium
Content of the Alloy

 

 

Chromium Weight Gain of
Content of Alloy Specimen in 168 hr
(wt %) (mg/cm?)*

0 10.44
2.83 16.46
4.61 1245
5.43 17.83
6.19 0.44
6.86 (INOR-8) 0.60
7.97 0.31

 

*One mil of oxidation corresponds to a weight increase
of about 4 mg/cmz.

revealed corrosive attack to a depth of 1 to 3 mils
in 500 hr, and mass transfer was observed when
sodium was the circulated fluid. Forced-circulation
loops fabricated of INOR-8 are to be operated in
order to determine the effect of various amounts of
iron and chromium on the corrosion resistance of
the alloy when exposed to fuel 107. In-pile cap-
sule tests have also been initiated in order to
study the effect of radiation on the corrosion re-
sistance of the INOR-8 alloy. |

Mechanical Properties. — Details of the studies
being made of the stress-rupture characteristics of
the INOR-8 alloys are presented in a subsequent
section of this chapter. Most of the tests thus
far have been made ot a stress of 8000 psi ot
1500°F in fuel 107, The results of two tests at
1800°F and 3000 psi in fuel 107 are presented in
Table 3.1.3. Data for Inconel tested at 1650°F

are included for comparison.

Properties of INOR-9

Alloys in the INOR-9 classification have a nickel
base, 12 to 17 wt % molybdenum, up to 5 wt %
niobium, and up to 7 wt % iron. The constituents
of the INOR-9 alloy were chosen on the basis of
corrosion resistance to fuel 107, Preliminary tests
have indicated that the oxidation resistance of
INOR-9 is about equal to that of Hastelloy B; the
creep properties are about the same as those of
INOR-8; and it is attacked by fuel 107 to a depth
of 1 to 3 mils in 1000 hr ot 1500°F. In contrast to

"

 
 

 

»

PERIOD ENDING SEPTEMBER 30, 1957

Table 3.1.3. Stress-Rupture Properties of INOR-8 Tested at a Stress of 3000 psi ot 1800°F in Fuel 107

 

 

 

Time to Specified Strain (hr) Time to Elongation
Heat No. Source Rupture at Rupture
1% 2% 5% 10% (hr) (%)
SP-16 Haynes Stellite Company 9 40 68 132
30-34 ORNL 3.4 8 24 50 94 20
Inconel* 6 16 40 130

 

*Tested at 1650°F (rather than at 1800 in NaF-ZrF“-UF4 {(50-46-4 mole %, fuel 30).

the attack on the chromium-bearing alloys, such as
INOR-8, the attack on the INOR-9 alloys does not
increase when the exposure time is increased from
500 to 1000 hr, Further, the rate of attack does not
seem to be affected by the niobium content up to
5% niobium,

Studies at University of Tennessee!

Studies of the solubility of chromium and molyb-
denum in nickel have been made with the use of
x-ray and metallographic methods. The composi-
tions studied ranged from 20 to 25 wt % molybdenum
and 3 to 15 wt % chromium.

The beta phase, Ni Mo, of the nickel-molybdenum
system was not observed in any of the ternary alioy
specimens. It was concluded that the addition of
chromium suppressed the formation of the beta
phase. As the alloying content was increased,
evidences of the existence of the gamma phase,
Ni,Mo; the delta phase, NiMo; and the primitive
orthorhombic *‘p'’ phase were observed. A report
in the form of a thesis has been prepared by T.S.
Lundy of the Umversnfy of Tennessee on this sub-
ject. :

Studies at Bafielle Memorial Institute 2. |
The semicnnual ANP materials conference was
held at Battelle Mernorlal Institute on June 17,

1957, and reports on. progress made during the |
preceding six months were presented. A summary - ceived from Superior Tube Company for creep-

of the information presented has been published.?

In general, the work at BMI has ‘stressed the de-
velopment of alloys with superior strength; thus

 

lUnder subcontract 582,
2Under subcontract 979.

3E. M. Simmons, Semiannual ANP Meeting, Battelle
Memorial Institute, ORNL CF-57-6-77 (June 17, 1957).

~ studies have been made of alloys containing alumi-

num and molybdenum as the principal strengtheners.

Several compositions were evaluated, and heat
B-3277 (nominal composition: 20% Mo-7% Cr-
1.5% Al-2% Nb-0.15% C) showed outstanding
creep strength, Thermal-convection loop tests of
this alloy with fuel 107 as the circulated fluid
showed that it was prone to attack by the salt be-
cause of its aluminum and chromium contents,
Attempts to operate forced-circulation loops fabri-
cated of alloys of this type were unsuccessful
because of the poor quality of available tubing.
This contract has been terminated and a final
report is being prepared.

Studies of Stress-Rupture Properties at New
Englond Materials Laboratory*

Specimens from the production heats of INOR-1
through -6 (ref 5), which were prepared by the
International Nickel Company, were tested to
obtain their stress-rupture characteristics at 1350,

1500, and 1650°F in air at the New England

Materials Laboratory. The results of the tests are
summarized in Table 3,1.4. A final report on the
work done under this subcontract is being prepared,

INOR-8 Alloys Produced by Supenor Tube
| Company®
A series of alioys of the INOR-8 type was re-
tupture evaluation.  These alloys were air-melted

at Superior and subsequently rolled to 0.065-in.-
thick strip. Creep-rupture tests are to be run in

 

4Unr.ler subcontract 584.

SFor compositions of these alloys see ANP Quar,
Prog. Rep. March 31, 1957, ORNL.-2274, p 177.

6-Un_r:!er subcontract 1112,

163

 
 

 

 

ANP PROJECT PROGRESS REPORT

fuel 107, A limited amount of 0.500-in.-OD,
0.045-in.-wall tubing was also processed from
these alloys that is sufficient for the fabrication
of three thermal-convection loops for operation
with fuel 107, Fabrication of the loops will be
started soon if inspection shows the quality of
the material to be satisfactory., Descriptions of
the alloys and the quantities awvailable are pre-
sented in Table 3.1.5, These heats were prepared
primarily for studies of the effect of zirconium
additions on the strength and corrosion resistance
of the INOR-8 alloys.

The work at Superior Tube Company has also
been concerned with the processing of tubing from

material supplied by ORNL, International Nickel
Company, Haynes Stellite Company, and Westing-
house Electric Corporation. This work is described
below under the contributions of the individual
companies to the over-all alloy development
program.

Status of Preduction Heats

Westinghouse Electric Corporation. ~ The work
accomplished to date on the production of pilot
heats of INOR-8 by Westinghouse was recently
reviewed and their first periodic progress report on
this development work was submitted to ORNL.

Table 3.1.4. Summary of Results of Stres#-Rupfure Tests of INOR-Type Alloys

1 hr at 2000°F, air cooled
16 hr at 1500°F, air cooled

Heat treatment:

 

Stress to Rupture in 1000 hr (psi)*

Elongation at Rupture in 100 hr (%)**

 

 

 

_:‘:’:’: At 1350°F At 1500 At 1650°F At 1350°F At 1500°F At 1650°F
INOR-1 13,000 3,900 1800 18 19 19
INOR-2 8,000 3,600 1600 1 12 20
INOR-3 15,000 5,100 2800 S 19 9
INOR-4 17,000 6,300 3400 1 5 n
INOR-5 15,000 4,500 1700 22 26 32
INOR-6 18,000 11,000 2600 S 8 . 20

 

*Extrapolated values.

**|nterpolated values.

Table 3.1.5. Air-Melted INOR-8 Alloys Prepared by Superior Tube Company

 

Nominal Composition (wt %)

 

Number of 0.065-in.-Thick Total Length of 0.500-in.-

 

 

Heat No. Creep-Rupture Specimen 0D, 0.045-in.-Wall
Mo C Fe C Zr Ti Ni Blanks Received Tubing Received

5712  16.99 6.64 3.14 0.070 0.08 * Bal 5 14§ 2in.

5713  17.20 6.66 3.08 0.064 0.07 * Bal 5 18 ft

5714  15.13 6.71 2.18 0.032 0.03 Ba! 4

5715 1527 4.61 2.16 0.024 0.02 Bal 7 |

5716  16.60 6.64 5.34 0.046 Bal 2 2 1in.

5717 16.13 6.79 5.34 0.094 0.02 Bol 3

5718  15.88 5.33 2.27 0.058 0.03 Bal 1 6ft 1in.

 

*Titanium present in heats 5712 and 5713, but analysis not yet complete.

164

 
 

 

 

 

v

In summary, 1hree heats of the composmons specu-
fied below have been prepared by lnduchon meltmg
in air; .

Amounts", -

.Cem.ponenf:s. _ L (wt %)
Molybdenum 1517
Chromium ' ) 6-8 . }
Cdren e et S
Carbon - o _.0-‘04-_""0-08‘.:7
Silicon 0.5 (max)
Manganese 0.8 (mox) |
Nickel ‘Balance

The first heat (INOR-8M, ~5200 Ib) was poured

into a 3000-1bmold, a 1500-1b mold, and two 300-ib
molds. The composition of the alloy was within~
the specified range, except for the carbon content,
which was found to be 0.13%, and: therefore forging

studies were carried out on the maferlal One

300-1b ingot was worked on a 1000-ton press forge

and the other onan 18,000-1b hammer forge. Both

forging ‘methods worked the metal; but the press.
forge moved the material easier than did the hammer -
forge. The press forge worked one 300-Ib ingot to -
approximately 4 in. in diameter, and the hammer
forge made a 4-in. square.. Both ,forg_l_ngs were
cooled and machined to 3-in.-dia bars. The bars

were inspected by ultrasonic techniques and were

found, for the most part, to be unsound, . The forger
ingots were then sectioned and were also found to -
contain internal flaws, The unsoundness of the |
forgings was therefore ctmbufed to shrmkage_

cavities in the or:gmai mgots. S

Two pOSSIb|e methods for. prevenhng 1he formu-_'_,.,_, R
tion of cavities ‘during solldlflcuhon were proposed el
Fiest, it was ‘suggested thut the mold be surrounded_'-;,._i__.‘~__"-7L':‘fi:]~ T
by an insulating jocket so that the .ingot would
~ cool slowly and - 1haf the mold be desugned o'

‘vide~'an ingot with a. smaller. length-io-wndth ratio,

The- second method mvolved mcreaslng the fhlck :

" ness of the mold’ walls, accentuahng the taper of
'-vdcuum-remelhng on fhe creep-rupture character-
istics of this hlgh-corbon Westinghouse heat, a
feeding the molten core“of the ingot and assure "~ 3-Ib ingot was prepared from the air-melted ma-
~- terial and tested, The results of the tests, which

the mold -and chill- castmg the ingot.on a copper
stool.  Both of these methods ‘should facilitate

progressive sohcl:flcaflon from the boflom to :lhe
top of the ingot.

The second heat (iNOR-SMZ ~500 |b) was cust'

into a mold of the type suggested in the first

  
 
  

 

PERIOD ENDING SEPTEMBER 30, 1957

“method proposed above, but- faulty pouring resulted
~in ‘mold-wash ‘and contamination of the alloy with
- _cast iron. This ingot was not forged.

~ The “third heat. (INOR-8M3 ~2000 Ib) was cast

into-a.mold of the second proposed type and was
.subsequently press-forged at 2150°F from a 15-in,
. square to a 9Q-m. round. Ultrasonic inspection

indicated internal flaws in the forging, but it was

o possible to obtam three 8-in.-dia extrusion billets
" from the maferlal These billets have been sent
- to ‘the In$ernuhonal Nickel Company to be extruded

~ to tube blanks,
~ Work for the immediate future at Westmghouse
“will be concerned with (1) the preparation of an

ingot of the initial high-carbon heat (M-1327) by

“the consumable-electrode method for determining
“ingot soundness, forgeability, and strength proper-

ties relative to those of the air-melted material;
(2) the preparation of plate, sheet, bar, and wire

products of air-melted INOR-8; and (3) an evalua-
tion of the casting characteristics of the alloy,

~ with -possible application in the shell molding of
pump impellers,

-As indicated .previously,” an evaluation of the

 high-carbon air-melted heat (M-1327) of INOR-8 is

. in progress. The results of thermal-expansion

measurements made on the material from room

~temperature to 1832°F, in cooperation with the

Ceramics Group, are shown in Fig. 3.1.1. The

‘average coefficients of thermal expansion, a, ob-

tained from the data in Fig. 3.1. l are, as a func-
tion of temperature:

Temperuture Range ' a
(°F) - {in./in.+°F)
o xwd

_‘.2'].2_7”5_2. e e 0.70

Co7s2-1m2 - © o 0.84
T izossr N X

‘212-133'2 . 0.86

ln order to obtam an’ mdlcahon of the effect of

 

7H. Inouye, T. K. Roche, and J. H. Coobs, ANP Quar,

~Prog. Rep, June 30, 1957, ORNL-2340, p 175.

165

 
 

 

P

 

 

ANP PROJECT PROGRESS REPORT

were carried out on the air melt end the vacuum
.melt under identical conditions, are given in

Table 3.1.6 for comparison. The data showed that
a definite improvement in rupture life and elonga-

‘tion was obtained by vacuum remelting, However,

the creep rate of the air-melted alloy was superior
to that of the vacuum-remelted material,
Haynes Stellite Company. — A 10,000-Ib air-

mglted heat® of INOR-8 has been purchased from__

 

8Designafed Haynes Stellite experimental dliby No.

' 8284 (heat SP-16).

UNCLASSIFIED
ORNL—LR—DWG 21929

TEMPERATURE (*F)
392 - ™2 e 1472 1832
0.0i800

0.01400

0.01200

TRUE EXPANSION

& 0.01000 FOR
S ~ 0.989-in. LENGTH
z
o
2 aoosoo
&
»
ul .
- PLOTTED DATA
S FOR 0.989-in.
g 000600

0.00400

000200 INOR—8
COMPOSITION (wi %):
Mo Cr Fe c Mn Ni
16.90 421 04 0.4 023 BAL

 

0 200 400 600 800 1000 1200
TEMPERATURE (°C)

Fig. 31.1. Lineoar Thermal Expansion of INOR-8
as Determined with the Use of the Vacuum ‘'Atcotran’’
Differential Transformer Recorder.

the Haynes Stellite Company to obtain the material
needed. for corrosion, welding, and creep testing.
The composition of this heat is given below:

Amount

Components (wt %)
Nickel 70.50
Molybdenum 15.82
Chromium 6.99
~ lren . 4.85
. Carbon 0.02

Tungsten 0.35
Silicon | 0.32
Cobalt 0.51
Manganese 0.34
Copper 0.03
‘Phesphorus - 0.009
Sulfur o 0.014
Boron = - 0.04

Casting ond fabrication of the alloy by hot and
cold working fo bar, plate, sheet, and strip products
were accomplished with little or no difficulty.
To date, the various shapes of this materia!

“which have been received include: a 3-in.-dia

forged bar; | 1-, ‘/2-, and ‘/4-in.-thick hot-rolled
plates; 0.063-in.-thick hot-rolled sheet; ?(‘-in.-dia .

* hot-rolled bar; six production-size extrusion billets;

and rolled strip for the processing of weldrawn
tubing. The extrusion billets will be sent to
Allegheny Ludlum Steel Corporation for conversion
to type 316 stainless steei—INOR-8 composite
tubing. Superior Tube Company has received the
rolled strip ond is currently processing the alloy
into small-diameter weldrawn tubing. Upon com-
pletion of the orders by the various fabricators
involved, enough material should be received to
fabricate approximately 30 forced-circulation loops,
one in-pile test loop, and specimens for a rigorous
program of creep and weld tests,

Table 3.1.6. Creep-Rupture Characteristics of the Westinghouse High-Carbon INOR-8 Heat M-1327
Tested at 1500°F in an Argon Atmosphere at a Stress of 10,000 psi

 

Time to Specified Strain (hr)

 

 

Type of Specimen Time to Rupture - Elongation
: ' 1% 5% 10% (hr) (%)
As air-melted material 32 _ 120 135 6.25
Vacuum-remelted material 25 80

125 214 54.5

 

166

 
 

 

 

U

International MNickel Company. —  As reported
previously,’ eight 9-in.-dio, - 12-in.-long billets,
two each of INOR-1, -2, and -5 ond cne each of
INOR-3 and -6, were extruded to 3)-in.-OD,
0.500-in.-wall tube shells  at the International
Nicke!l Company. Recently the reducing of these
tube shells was completed at INCO, and the re-
duced shells have been sent to the Superior Tube

Company for processing to 0.500-in, -OD 0. 045-m. :

wall seamless tubing.

Considerable delay was experlenced in completmg
the tube reducing of these shells at INCO for
various reasons. Primarily, the delay was caused
by unexpected difficulty in annealing the tube
shells following cold working on the tube reducer.
It was found that the alloys were subject to thermal
splitting when a total reduction of 30% or less was
given prior to annealing. This behavior was en-
countered at annealing temperatures between 1700
and 2150°F. However, by increasing the total cold
reduction between anneals to over 50%, the alloys
were successfully annealed at 2150°F  without
splitting. The tube shells, which were reduced to
1.66-in.-0D, 0.148-in.-wall shells at INCO,. have
been sent to Superior Tube Company, Descrip-
tions of the tube blanks sent to Superior Tube
Company for further processing are given in

Table 3.1.7.

Composite Tubing |

All the nickel-base alloys exhibit mass trnnsfer
when used as containers for flowing sodium at

 

H. Inouye and T. K. Roche, ANP Quar._Prog. Rep
March 31, 1957, ORNL 2274, P 134. . -

Table 3 I 7. |NOR-Type Tube Sbells Produced by
Internuhonal Nlckel Compuny

 

Approxtmufe Lengfhs of

 

H'eatf,No.‘ Alloy  1.66-in.-0D; 0.148in.-
o °="9"°"°" _ Wall Tube Sheils (ft)

Y-8195  INOR-1- 64

Y-8197.  INOR2 64

Y-8196  INOR3 . . 32

Y8200  INOR5 64 °

Y.8199  INOR-6 32

 

*Products from 4800-1b air melts.

PERIOD ENDING SEPTEMBER 30, 1957

high temperatures. The extent of mass transfer is

-temperature dependent, being slight at 1300°F and

tolerable -at 1500°F, For service temperatures of
1650°F and ‘above, it appears that a composite

‘tube, composed of a nickel-base alloy and type
316 stainless - steel, is desuroble and, possibly,

necessary. :

There are three basic problems assocmted wi th
the use of such composite tubing in fluoride fuel-
to-NaK heat exchangers. The first problem is the
fabrication of the composite tubing. Experiments
indicate that tubing of acceptable quality can be
made by following the practice of coextruding
composite billets. The high temperatures and
pressures used during coextrusion result in the
formation of metallurgical bonds. Dimensional
control is maintained by a suitable choice of the
starting composite and by conditioning of the tube
shell after extrusion. The outlook is optimistic,
since it is believed that composites of type 316

 stainless steel and INOR-8 should not be more

difficult to prepare than composites of Inconel and
type 316 stainless steel, which have been made in
substantial quantities.

The second problem is the welding of the com-
posite tubing. The mechonical properties of the
welds and the resistance of the weld nugget to
sodium mass transfer must be investigated. Con-
siderations of the properties of alloys intermediate
between INOR-8 and type 316 stainless steel may

provide information on these points.

The six intermediate alloys described in Table
3.1.8 were cast, fabricated, and aged to determine

_the equilibrium phases and the effect of various

: T_u'ble 3.1.8. Compositions of Alloys Intermediate
-Between INOR-8 and Type 31§ Stainless Steel

 

 

 

Alloy ~ Nominal Compéshion’ (wt %)
Designation Mo Fe Cr Ni

~ INOR-8 7 5 7 71
Atloy-1 15 n 8 66

a2 13 22 9 56
-3 1 33 1 45

4 9 44 12 35

-5 6 56 13 25

-6 3.5 67 14 15

Type 316 25 70 15 12

stainless stee!

 

167

 
 

 

 

 

ANP PROJECT PROGRESS REFPORT

aging femperatures on their tensile properties.
Thus far only the microstructures of the cast-and-
aged. alloys have been determined. It was found
that the alloys 1 and 6 were solid solutions, and
therefore they should not exhibit undesirable
mechanical properties. The other four alloys
showed small quantities of a precipitate. The
effects aond nature of the precipitate have not yet
been determined. Based upon the preliminary
study, the configuration for a tube header sheet

should be one in which the stainless steel is:

diluted by a minimum omount of the nickel-base
alloy after welding. ' '

The third problem involves an investigation of

the diffusion of the elements of one alloy into the
other, Since diffusion is a temperature-dependent
phenomenon, its effect is expected to be greatest ot

high temperatures. ' Of prime importance is the -

determination of the depth to which the various

elements will penetrate each other under a variety -

of conditions, since this will determine the thick-
nesses of the layers comprising the composite.
Since the nickel-base alloy is stronger thaon the
stainless steel at high temperatures, the composite
tube should contain as large o percentage as pos-
sible of the nickel-base alioy. The diffusion
effects have been studied in a series of tensile
tests of a composite composed of 50% |INOR-8 clad
on both sides with type 316 stainless steel. The
data obtained are shown in Table 3.1.9.

JOINING EXPERIMENTAL INOR-8 ALLOYS

R. E. Clausing G. M. Slgughter
P. Patriarca

Ten heats of experimental nickel-molybdenum
alloys similar to INOR-8 have been tested to de-
termine whether these alloys can be successfully

fabricated into complex components by means of
brazing and inert-gas-shielded arc welding and
whether, once fabricated, the material will have
suitable mechanical properties. Since the compo-
sition of Battelie Memorial Institute heat 3766 was
closest to the composition range of INOR-8, the
experimental results obtained in the study of speci-
mens of heat 3766 are summarized here. The
analysis of heat 3766 is compared with the compo-
sition range of INOR-8 in Table 3.1.10,

It was readily determined that specimens from
heat 3766 could be brazed in a dry-hydrogen atmos-

phere with Coast Metals brozing alloy No., 52 and
similar

brazing alloys by -using conventional

techniques.  The determination of weldability,

 however, was inherently more complex and time-

consuming. Many tests were required to determine

Table 3.1.10. Comparison of Analysis of
Battelle Memorial Institute Heat 3766
with the Composition Range of INOR.8

 

Quantity Present (wt %)

 

 

Components INOR-8 Heat 3766
Molybdenum 15-19 13.3
Chromium 6-8 6.05
Iron 4--6 6.07
Carbon 0.04-0.10 0.065
Manganese 0.78
Siticon 0.43
Titanium 0.11
Magnesium 0.05
Nicket Balance Balance

 

Table 3.1.9. Room-Temperature Tensile Properties of INOR-8 Clad on Both Sides with Type 316 Stainless Steel

 

Heat Treatment Tensile Strength

Yield Strength
at 0.2% Offset

Elongation

 

(psi) (psi) (%)

~ Annealed 89,400 33,600 59
500 hr at 1300°F 94,400 36,800 51
500 hr at 1500%F 93,600 34,600 42
500 hr ot 1650°F 83,000 32,400 - 28
500 hr ot 1800°F 79,800 31,800 43

 

168

 
 

 

 

 

 

\'5.

‘mental Studies of Inconel,

(1) that sound weld deposits could be made, that is,
deposits free of cracks and pores, as well as
excessive segregation, (2) that there would be no
deleterious regions in the heat-affected zone,
(3) that the base metal would not be subject to
hot tearing, and (4) that the mechanical properties
of the joint would be satisfoctory, both in the
as-welded and the aged conditions.

A test specimen was designed which permitted
nondestructive determination of the soundness and
freedom of cracking of the weld metal and the
heat-affected zone. This specimen also provided
material for metallographic and hardness speci-
mens, as well as the bend- and tensile-test speci-
mens needed for determining the mechanical
properties of the welded joint. The specimens

were machined from plates that had been welded

under a high degree of restraint. ~ The methods
used for obtaining the specimens were similar to
those shown in Fig. 3.2.18 of Chap. 3.2, '‘Develop-
" this report.

Examinations of many specimens of weld deposits
made by inert-gas-shielded tungsten-arc welding
procedures on BM!| 3766 base metal with filler
wire of the some material failed to reveal any
significont porosity and only one small root crack.
On the basis of these examinations, which were
made with dye-penetrant, x-ray, and metallographic
methods, it may be concluded that it should not be
difficult to make sound weld deposits in BMI 3766
joints,

Hardness measurements indicated that very little,
if any, hardening occurs either in the weld metal or
in the heat-affected zone as a result of aging for

200 hr at temperatures from 1100 to 1700°F, Aging

at 1700°F for 200 hr reduced the hardness of the_

weld metal to that of the base metal, while aging
at 1500°F for 200 hr results in considerable soften-
ing of the weld metal, Aging at 1100, 1200, and

1300°F for 200 hr did not appreciably change the

hardness. Metallographic examination of hardness
specimens indicated that the eutectic material -
which was present in the welds initially was 170
seemed fo have nearly identical base-metal and

partially spheroidized upon aging at 1500°F and

completely spherocdrzed upon agmg at 1700°F for i

200 hr, ,

Bend test specimens were eut from the - Welded
joints and tested in the as-welded and aged condi-
tions at temperatures ranging from 1100 to 1700°F
in order to evaluate the mechanical properties of

‘the joints. It is possible in a bend test to test not

PERIOD ENDING SEPTEMBER 30, 1957

only the joint as a whole but also the various com-
ponents of the joint. For instance, the deformation
can be made to occur in a selected area, such as
the base-mefal—weld-metal interface, which might
not otherwise deform in the tensile test. The data
obtained from these tests are presented in Table
3.1.11 in terms of deflection when the load had
dropped to two-thirds of its maximum value. The
deflection values multiplied by a factor of 100
roughly correspond to the elongation values ob-
tained in tensile testing (that is, a 0.100-in, de-
flection corresponds to approximately 10% elonga-
tion in the standard 0.252-in.-dia tensile specimen).

- The maximum possible deflection with the particu-

lar machine used is 0.3 in, Therefore, when the
reported deflection is 0.3 in., the specimen is not
considered to have failed. An analysis of the data
in Table 3.1.11 reveals that none of the specimens
tested at 1100°F failed, that the specimens tested
at 1300°F in the aged condition exhibited con-
siderably less ductility than those tested in the
unaged condition, and that the specimens tested at
1500°F in the aged condition showed considerably
more ductility than specimens tested in the unaged
condition, This may be correlated with the ob-
served change in microstructure, that is, the dis-
appearance of eutectic material. Further, the speci-
mens tested at 1700°F also exhibited more ductility
in the aged condition than in the unaged condition.
This is again associated with the disappearance of
eutectic material, At 1300 and 1500°F the heat-
offected zone showed a loss in ductility when
compared with the base metal. However, the
ductility was equal to or greater than that of the

-Vweld metal,

A number of tensile specimens cut from the
weldments were tested at temperatures ranging

from room temperature to 1700°F. The results of
these tensile tests, as shown in Table 3.1.12,

. agree very well with the results of the bend tests

(Table 3.1.11). It may be noted that the speci-

~ mens which were aged for 200 hr at 1500°F and at

1700°F and were tested at the aging temperatures

This is indicated by the

weld-metal properties.

- fact that one specimen failed in the weld at each

temperature, while duplicate specimens failed in
the base metal, with similar strengths and elonga-
tions. It may also be noted that very few of the
failures occurred at the gage marks, and thus it
appears that this material is not particularly notch

169

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 3.1.11. Results of Bend Tests of Specimens Cut from Welded Joints of BMI 3766 Filler Wire
on Base Metcl of the Source Material

 

Test Temperature

Area Tested

Final Deflection when Load
Dropped to Two-Thirds of

 

(F) Maximum Value {in.)
1100 Weld centerline 0.300+
Weld centerline 0.300+
Weld centerline* 0.300+
Weld centerline* 0.300+
1300 Boase metal 0.290
Heat-affected zone and weld interface 0.220**
Weld centerline 0.240
Weld centerline 0.220
Weld centerline* 0.120
Weld centerline* 0.145
1500 Base metal 0.300
Heat-affected zone and weld interface 0.165
Weld centerline 0.080
Weld centerline 0.070
Weld centerline* 0.200
Weld centerline* 0.190
1700 Weld centerline 0.200
Weld centerline 0.100
Weld centerline* 0.300
Weld centerline* 0.300

 

*Specimens aged 200 hr ot the test temperature prior to testing.

**Specimen broke in base metal.

sensitive. Specimens aged at 1300 and at 1400°F
failed in the welds, and it seems that the weld
metal is weaker than the base metal, at least in
the aged condition. Unfortunately, the solution
heat treatments of the base materials tested at
room temperature and at 1500°F were not identical
with the solution treatment given the base material
used in making these welds. Therefore, a direct
comparison between the properties of the base-
metal tensile specimens and the base metal in the
transverse-test specimens is not truly valid. In
order to make such a comparison, base-metal
tensile tests should be made of specimens given
the same heat treatment as that given to the base
metal used in the tronsverse tensile tests, The
base-metal specimens should then be tested at all
the temperatures listed in Table 3.1.12, It would
also be desirable to test unaged tensile specimens

at 1300°F in order to verify the results of bend

170

tests, which indicate that aging decreases the
ductility at this temperature. |t should also be
noted that the room-temperature results indicate
that failures should occur in the base metal,

A comparison of the tensile properties of weld-
ments of Inconel, Hastelloy B, and the experimental
INOR-8 alloys is given in Table 3.1.13. Paort of
the data were interpolated and therefore may be
subject to some deviation from the actual values,
and the Inconel data for aged specimens are actually
data for unaged specimens, since it is thought that
any changes which would occur on aging for 200 hr
at the specified temperatures would not appreciably
alter the date. It may be noted that at nearly all
temperatures the values for the INOR-8 specimens
are between those for Inconel and those for Hastel-
loy B. In general, the strength is less than that
of Hastelloy B but greater than that of Inconel,
while the ductility lies between that of Inconel

b

 
 

 

 

 

 

9

Qs‘o

PERIOD ENDING SEPTEMBER 30, 1957

~ Table 3.1.12. Results of Tensile Teits of BMI 3766 Weldments

 

 

Test Type of Tensile Strength at Reduction
Temperature Tensile Test Strength ~  0.2% Offset Elongation of Area Location of Fracture
(°F) Specimen (psi)  (psi) (%) (%)
Room All weld metal 108,500 77,800 40 34 Weld metal; no defect
All weld metal 107,900 64,600 40 32 Through small inclusion
: ' in weld metal
Transverse® 108,800 = . 57,200 42 44 Base metal
Transverse 108,600 - - 58,700 43 37 Base metal
Base metal® 107,100 41,800 38 Base metal
Base metal® 105,500 39,400 50 38 Base metal
1300 Transverse® 56,400 . 37,300 12 13 Weld; no defect
Transverse® 59,900 _ _27,000 13 13 Weld; no defect
1400 Transverse® . 59,000 37,600 15 14 Weld; no defect
Tronsverse® 57,000 41,700 13 8 Weld; no defect
1500 All weld metal 51,500 36,800 12 7 Weld metal; no defect
All weld metal 51,700 .37,900 12 5 Weld metal; no defect
Transverse 51,500 | 36,800 12 6 Weld; ot gage mark
Trnhéversg 51,700 37,900 12 6 Weld; no defect
Transver;ec _ 46,700 31,400 23 27 Weld; ot gage mark
" Transverse® 45,800 | 32,200 35 - 34 Base metal
Base metal® 47,700 25,100 40 35 Base metal
Base metal? 47,100 - 26,300 40 32 Base metal
1700 Transverse® 28,0000 . 27,400 30 ' 25 Weld; no defect
32 15 Base metal

 

Transverse® 27,2_00 26,700

“Specimens were cuf to contain weld metul bose metal, and the weld-metal-base-metal interface within a uniform

gage length,

This material was given a longer soluhon treatment than was the joint-base material.

€Specimens were aged 200 hr at the test femperutufe prior to testing.

and that 6frHc»x‘é;fe'liloyi'B_.';'-:J-Tfié" lowest ;:!uc't_ility

value reported for INOR-8 is above 10% at 1300°F, .
In addition, numerous- fl'termal-convechon loops

L by these methods will not be subject to any un-
“usval restrictions as a result of the fubrlccmon
Qprocesses.

have been fabricated of olloys similar to INOR-B'-"'-_
without difficulty. Corrosion tests of many of ihese. e

loops have been completed, and there have been
no. md:cahons of any deleterious effects near or
“in the welds.l Although the filler metal used in
“ali these loops was Hastelloy. W rather than being
the same as the base metal, it is thought that the
- successful fabrication and operation of these loops
'is a good indication that these materials should

“not be difficult to fabricate by welding. . In summary, -
it appears that material of the INOR-8 composition

will be readily fabricable by both welding and
brazing techniques and that structures fabricated

_ STRESS-RUPTURE PROPERTIES OF NICKEL-
.. 'MOLYBDENUM BASE ALLOYS IN A
FUEL ENVIRONMENT

J.W Woods ' - D, A, Douglas

Evuluuhon studles of rhe creep characterlsflcs

_of the nlckei-molybdenum alloys under a constant
load .in the fuel mixture NaF-KF-LiF- UF, (11.2-
~ 41-45,3-2.5 mole %, fuel 107) were conhnued The

results of tests made during the quarter are pre-

sented in Table 3.1,14.

171

 
 

 

ANP PROJECT PROGRESS REPORT

Table 3,1.13. Comparative Tensile Properties of Inconel, Hastelloy B, and INOR-8 Weldments

 

 

 

 

As Welded Aged 200 hr at Test Temperature
Tempercture . Tensile Yield Strength Tensile Yield Strength
Material Elongation Elongati
(°F) Strength  at 0.2% Offset gatio Strength ot 0.2% Offset gaiton
(psi) (psi) %) (psi) (psi) (%)
Room {nconel 88,000 50,100 - 40
Inconel 90,700 52,700 40
Hastelloy B 118,300 21
Hastelloy B 124,500 23
INOR-8 108,800 57,200 43
INOR-8 108,600 58,700
1300 Inconel 50,000* 35,000* 40*
Hastelloy B 108,300 5.0
Hastelloy B 109,100 2.0
INOR-8 56,400 37,300 12
INOR-8 59,900 27,000 13
1500 Inconel 37,000 28,000 45
Inconel 37,000 28,000 45 37,000* 28,000* 45*
Hastelloy B 66,700 9 67,200 : 19.5
Hastelloy B 62,300 12 63,800 10
INOR-8 51,500 36,800 12 46,700 31,400 23
INOR-8 51,700 37,900 12 45,800 32,200 35
1700 Inconel 22,000* 20,000* 50*
Hastelloy B 40,000** 30**
INOR-8 28,000 27,400 30
27,200 26,700 32

 

*Data for as-welded specimens.

**These are conservative figures based on data obtained ot 1650°F.

Tests of specimens from Battelle Memorial
Institute alloy heats B-3874 through B-3879 were
completed.  As previously reported,'® no data
could be obtained on alloys B-3874, B-3875, and
B-3876 because of their poor welding characteris-
tics. Alloys B-3877, B-3878, and B-3879 ex-
hibited good stress-rupture characteristics but the
ductility was poor. A metallographic examination
of the specimen from heat B-3879, which was
tested in the annealed condition, revealed a small
grain size and a heavy concentration of carbides
in the grain boundaries (Fig. 3.1.2). In the past,
the alloys that have exhibited the best creep
characteristics have also shown this concentration

 

105, W. Woods and D. A. Douglas, ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340, p 179.

172

of carbides in the grain boundaries. Little, if any,
corrosive attack is evident. Even though this alloy
looks promising from the stress-rupture standpoint,
dynamic corrosion studies have indicated con-
siderable corrosive attack by fuel 107 of all me-
terials containing substantial quantities of alumi-
num,

In an effort to determine the effect of annealing
temperature on the creep properties, a series of
specimens from ORNL vacuum-melted heat 30-63
(13% Mo-5% Fe-0.06% C-5% Nb-—bal Ni) were
tested at 8000 psi and 1500°F in fuel 107. The
data obtained are presented in Table 3.1.15,
There is little difference in the creep properties of

‘the material annealed at 2100°F and that annealed

at 2200°F. However, when these annealed speci-
mens were aged at 1300°F prior to testing, the

*

&

 
 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

Table 3.1.14. Stress-Rupture Properties of Nickel-Base Molybdenum Alloys Tested at 1500°F and o Stress of
8000 psi Exposed to NcF-KF-LiF-UF‘ (11.2-41-45.3-2.5 Mole %, Fuel 107)

 

 

 

Heat No. Composition (wt %), Bfilance Nickel Rl;_p:;:e Elongation Heat
Mo Cr Fe C Al" Mn Ti Other (he) (%) Treatment*
Battelle Memorial Institute Alloys, Air Melted
B-3877 174 6.9 09 007 21 09 006 1.3Nb 904 16 4
B-3878 19.9 4.5 0.06 1.5 0.9 0.9 Nb 751 5 3
B-3878 19.9 4.5 0.06 .5 0.9 0.9 Nb 801 | 6 4
B-3879 19.8. 6.9 43 0.05 1.5 09 008 1.0Nb 1641 14 3
B-3879 19.8 69 43 0.05 1.5 0.9 008 1.0Nb 2034 11 4
ORNL Alloys, Vacuum Melted
30-62 16 7 5 0.06 0.5 0S5 0.15 Zr 492 53 1
30-62 16 7 5 0.06 0.5 0.5 0.15 Zr 1189** 66 2
VT-66 17 7 4 0.14 0.8 195%* 51 1
VT-66 17 7 4 0.14 0.8 330** 51 3
7 Haynes Stellite Company Alloy, Air Melted
SP-16 158 7 4.8 0.02 0.3 530 13 )
SP-16 0.02 0.3 536 12 2

 

15.8 7 4.8

*Heat treatment No. 1:
Heat treatment No, 2:
Heat treotment No. 3:
Heat treatment No. 4:

**Test discontinued before rupture occurred.

solution annealed at 2000°F

rupture life and ductility were decreased, The
specimen annealed at 2200°F before aging was
less affected than the one annealed at the Iower
temperafure. '

In order to obtain an indication of the effecf of
vacuum remelting on the creep properties of the
“high-carbon-content Westinghouse Electric Corpo- .
| a vacvum-melted “3-1b
casting, VT-66 was' prepared from the air-melted -
. material.
casting were tested in fuel 107, The ductility of_ ,

ration - heat of INOR-8, .
- Stress-rupture specimens - from -this

‘the remelted material was supeflor to that of the

original melt, and there was a smuH increase in

the -rupture life; however, the creep rate was not

improved over that ‘obtained for the alr-melted'.
The VT-66 specimen, which -

Wesh_ngh_ous._e heat.
was tested in the annealed condition, is shown in
Fig. 3.1.3. In view of the high-carbon content

solution annealed at 2000°F for 1 hr.
for 'I hr and aged at 1300°F for 50 hr.
solution annealed at 2100°F for 1 hr.
solution annealed ot 2100°F for 1 hr and aged at 1300°F for 50 hr.

(0.14%), surprisingly few carbide precipitates are
present. Those present are dispersed randomly
throughout the sample, with no concentrations in
the grain boundaries. The lack of carbides prob-
ably accounts for the poor stress-rupture properties.

~There was little conosive attack.

Two tests were conducted on sheet material

obtumed “from a 10,000-1b qir-melted heat of the

INOR-8 -composition, heat SP-16, alloy 8284, pro-

- duced by the Haynes Stellite Company. The alloy

exhibited a shorter time-to-rupture than was ex-
pecfed but fl'ns could be attributed to the low
carbon content (0.02%). The ‘specimen tested in

~ the annealed condition is shown in Fig. 3.1.4. As

was expected, there are few carbide precipitates,
and those present in the grain boundaries do not
form a continuous network, The properties of the
Haynes Stellite heat are compared with those of

173

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

 

RN e m*‘
UNCLASSIFIED &

s"l-i_p ¢

 

TRE
- ._.ns — e A}
s > ox o g o 2 2 INCHES

 

 

 

 

 

 

Fig. 3.1.2. Stressed (8000 psi) and Unstressed Sections of a Speélmen of Alloy B-3879 Tested in the Annealed
Condition at 1500°F in Fuel 107. Etchant: chrome regia. 500X. Seerativith-captton)

Table 3.1.15. Results of Study of Effect of Prior Heat Treatment on Creep Properties of Specimens
from ORNL Vacuum-Melted Heat 30-63 Tested at 8000 psi and 1500°F in Fuel 107

 

 

Rupture Life Elongation
Heat Treatment (hr) (%)
Annealed at 2100°F for 1 hr 750 50
Annealed at 2200 for 1 hr 740 25
Annealed at 2100°F for 1 hr and aged at 1300°F for 50 hr 350 15
Annealed ot 2200°F for 1 hr and aged at 1300°F for 50 hr 670 13

 

other heats of INOR-8 in Table 3.1.16, and the
compositions of the heats are given in Table
3.1.17, It may be seen that the Haynes alloy,
8284, has the best creep properties of all the
alloys listed. It is felt that with an increase in
carbon content, this alloy will have a rupture life

of 1000 hr or more at 1500°F and 8000 psi.

174

Specimens from the vacuum-melted heat 30-62
were tested to determine the effect on the stress-
rupture behavior of adding zirconium to the INOR-8
clloy. Previously, specimens of 30-lb heats of
INOR-8 had failed in 680 hr with 30% elongatien,
but testing of the zirconium-bearing specimen was
discontinved ot 1189 hr and the elongation was

 
PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

4
*
-«
-«% o
‘..: - ’V : gfl
{ AL "
& o ! o o ‘ o ' .
33 S 500X 3 S g 3 § INCHES
Fig. 3.1.3. Stressed (8000 psi) and Unstressed Sections of Specimen from Vacuum-Melted Heat VT-66 Tested in
" the Annealed Condition at 1500°F in Fuel 107, Etchont: chrome regia. 500X. (Seeret-with-cuption)
»
. - lo ) l . e .-
i ooolet 100X 0 o - o e o INCHES - ';
‘ . a . Sy . - w n : ) J
: u Fig. 3.1.4 Stressed (8000 psi) and Unstressed Sections of Specimen of Hayes Alloy 8284 Tested in the

: Annealed Condition ot 1500°F in Fuel 107. Etchant: chrome regia. 100X, {Seeret-with captien)

175

 

 
 

 

 

ANP PROJECT PROGRESS REPORT

Table 3.1.16. Comparison of Creep Properties of Several Heats of INOR-8 Tested at
1500°F and a Stress of 8000 psi in Fuel 107

 

 

 

Time to Specified Strain (hr) ‘Rupture Total
Alloy Designation Source Life Elongation
) : 1% 2% . 5% 10% (hf) (%)
8284 (heat SP-16) Haynes Stell.ite 36 100 280 510 530 13
Company
M-1327 " Westinghouse Electric’ 14 32 64 92 110 16
: Corporation

30-62 ORNL 29 72 180 340 1200 66
VT-51 ORNL 28 62 200 450 1243 49

75 180 340 670 26

 

30-34 ~ ORNL 36

Table 3,1.17. Compositions of Yarious INOR-8 Heats

 

Nominal Composition (wt %), Balance Nickel

 

Alloy Designation

 

Mo Cr Fe C Other
8284 (heat SP-16) - 16 5 0.02 0.35 W
M-1327 | 17 4 0.14
30-62 16 7 5 0.06 0.15 Zr
VT-51 17 10 7 0.06
30-34 15 6 5 0.06

 

66%. The creep rates of both specimens were
similar, and thus the only difference was the in-
creased rupture life of the zirconium-bearing
alloy, which may be ctiributed to an increase in
ductility, The zirconium-bearing specimen, which
was tested in the annealed and aged condition is
shown in Fig. 3.1.5. Bands of various grain sizes
ranging from coarse at the surface to fine at the
center of the specimen may be seen. The usual
heavy concentration of carbide particles in the
grain boundaries that is associated with all the
high-strength nickel-molybdenum alloy systems is
also present. The specimen shows corrosion attack
to a depth of approximately 0.5 mil.

Although the INOR-8 alloy does not represent the
ultimate in mechanical properties which can be
cbtained from the nickel-molybdenum alloy systems,
it is superior to type 316 stainless steel and

176

Inconel at 1500°F in a fuel environment, as is
shown by the representative creep data in Table
3..18, Tensile tests of alloy 30-62 in air at
1500°F indicated a tensile strength of 46,000 psi
with 20% elongation, The maximum corrosive
attack observed corresponds to the corrosion

found for Hastelloy B in the same fuel environment.

CORROSION STUDIES
J. H. DeVan

Forced-Circulation Loop Test of a Nickel-
Molybdenum Base Alloy Exposed to Fuel 107

J. H. DeVan R. S. Crouse

The first forced-circulation loop fabricated of an
experimental nickel-molybdenum base ciloy com-
pleted 1000 hr of operation with fuel 107 at a

maximum fuel-to-metal interface temperature of

a4

 
 

 

 

1760°F, The loop alloy had the nominal ¢omposi-
tion 17% Mo—-6% Fe-bal Ni. Other operating
conditions for the loop, 7641-9, are given below:

Maximum bulk fuel temperature 1610°F

Fuel temperature drop 300°F
Reynolds No. ' 10,000
., Flow rate 1.4 gpm
Ratio of surface area of heated 2.2 il1.2/in.3
section to total loop volume

 

PERIOD ENDING SEPTEMBER 30, 1957

An examination of a hot-leg section of this loop
showed attack by the fuel to be confined almost
entirely to grain boundaries. The attack reached
a maximum depth of 4 mils at the point of maximum
wall temperature. ‘As may be seen in Fig. 3.1.6,
the attack was, in general, in the form of small,
discontinuous voids, although shallow surface pits
are also evident along the exposed surface, Con-
siderable oxidation occurred on the outer surface
of the hot leg, which was in contact with air during
operation of the loop. In addition to forming a

 

 

 

 

 

 

 

Flg. 3.'! 5. Speclmen of AI on 30-62 Tufed In the Annnaled and Aged Condlflon at 8000 psi and 1500°F in
Fuel '|07. _ Etchanf' ' chrome regiu. soox. (&cmHh-eep'hon)

chlo 3. 'l 18. Repruentuflve Creep T est Data Obfclned at 'ISDO°F

 

 

 

LT : B g B . , . ' Rrar.lptrure‘ Total
L : sl T : _ Time te Specified Strain (hr) )
“Alloy Stress - Test ' : . - Life Elongation

: (psi) Envi_ronrnen'f_r 0.5% _1%, 2% 5% 10% (hr) ' (%)

" CAlley'30-62 < 8000  Fuel107 7 29 72 180 340 1200 66
Type 316 . - 6300 Fuel 30 4 20 95 300 - 540 o 977 55

stainless steel - : ) : :
Inconel 3500 Fuel 30 22 70 180 600 1000 1000 10
u ‘ ' 8000 Fuel 30 3.5 15 30 47

 

177

 
 

 

 

ANP PROJECT PROGRESS REPORT

heavy, uniform oxide film, the oxidation had pro-
ceeded preferentially along grain boundaries to
depths of as much as 9 mils.

Examinations of cold-leg sections showed numer-
ous surface pits and some areas of intergranular
attack to a depth of 2 mils. A very thin deposit of
tiny metal crystals was present over most of the
cold-leg surface, and, in addition, widely dispersed
clumps of metal deposits up to 6 mils in thickness
were found. The heaviest deposits found are shown
in Fig. 3.1.7. The deposits interfered with drain-
_ age of the fuel from the walls of the loop along
these portions of the cold leg, and therefore the
deposits were partially covered with a heavy
fluoride-mixture film, Samples of this film which
contained sizeable quantities of metal crystals
were analyzed and were found to contain the fol-
lowing elements: |

Fuel Quantity Found

Constituents (wt %)
K 5-10
Na 2-5
Li 2-5
) 10

Quantity Found
Contaminants

(wt %)
Al 0.05
Fe 0.3
Cr 0.1
Mo _ 0.1
Ni | 5

I+ was possible to make a separation of the
metal crystals from the residual fuel by using an
ammonium oxalate solution, and analyses of the
metal particles are in progress. The separated
metal particles were quite magnetic, and, on the
basis of the above cnalysis, it would appear that
they contain large percentages of nickel, with
some iron, \

Before-test and after-test samples of fuel from
this loop showed no change in nickel and iron
contamination, but a slight increase in chromium
contamination was noted in the after-test sample.
It is assumed that Hastelloy W, which was used
as a filler wire in making the loop welds, served
as the source of chromium,

178

 

 

Fig. 3.1.6. Hot-Leg Attack in Nickel-Molybdenum
Base Alloy Forced-Circulation Loop 7641-9 Operated
with Fuel 107. Etchant: 250X, (Gecres

with-voptior
BRI UNCLASSIFIED.
R T-13256

aqua regia.

 

Fig. 3.1.7Z. Cold-Leg Sections of Nickel-Molybdenum
Base Alloy Forced-Circulation Loop 7641-9 Operated
with Fuel 107. Unetched. 250X. G&m‘flfihfl%

 
 

 

 

 

Thermal-Convection Loop Tests of Nickel-
- Molybdenum Base Alloys Exposed to
Fuel 107

D. A. Stoneburnerl' J. R. DiStefano

Two '_thermal_-convecztion loops constructed of
experimental nickel-molybdenum base alloys were
tested with the fuel 107 at 1500°F for 1000 hr,

The compositions of the alloys from which these

loops (Nos. 1136 ond 1155) were fabricated are

presented in Table 3.1.19, dlong with the results
of chemical analyses of the fuel circulated in these
and three previously operated loops. "
Metallographic examination of loop 1136 revealed
heavy surface roughening and pits along the hot-
leg surface to a maximum depth of 5 mils. In
addition, very large and irregular-shaped voids
were present at depths as great as 9 mils below
the surface. As may be seen in Fig. 3.1.8, these
voids were elongated in. the direction of the tubing
axis and did not appear to be connected with the
attacked areas above, They probably reflect fab-
rication defects rather than actual areas of attack,
The cold leg of loop 1136 showed light surface

roughening with no evidence of deposits. . The

analysis of the fuel after the test, as shown in
Table 3.1.19, indicated considerable reaction of
the fuel with aluminum and a slight reaction with
titanium, : :

lLoop 1155 showed heavy surface roughening and
surface pits, with heavy intergranular subsurface
void formation to a depth of 3 mils in the hot-leg

 

1D, H. DeVan and D. A. Stoneburner, ANP Quar.,

Prog. Rep. June 10, 1957, ORNL-2340, p 185,

PERIOD ENDING SEPTEMBER 30, 1957

section.  The cold leg showed moderate surface
roughness, with second-phase material concen-
trated near the surface. No cold-leg deposits or
layers were visible, Fabrication defects in the
form of voids and cracks were found in the samples
examined. Such defects occurred generally to a
depth of 7 mils, although one crack extended from
the outer surface into the sample to a depth of

20 mils.

 

Fig. 3.1.8. Hot-Leg Attack in Nickel Molybdenum
Base Alloy Thermal-Convection Loop 1136 Which
Operoted with Fuel 107 for 1000 hr at 1500°F. Etchant:
copper regia. 100X, (&eeret-with copiien)

Table 3.1.19. Anglyses of Corrosion Products in Fuel 107 After Circulation in
'Nickel-Molybdenum Base Alloys at 1500°F :

 

 

o ' ' ' 70pe;rufing

Loop - Alloy Composition Time Quantity of Alloy Constituents

Ne.. O (mR) (hr) in After-Test Fuel Samples
1125 17 Mo—d V=bal Ni 1000 210 ppm

1123 17 Mo—4 V—bal Ni 500 290 ppm

"1135 16 Mo—6 Cr—1 Nb—1 Al-bal Ni 1000 455 ppm Cr, 20 ppm Nb, 1455 ppm Al
1136 16 Mo—2 Al=1.5 Ti=ba! Ni 1000 1400 ppm Al, 370 ppm Ti |
1155 INOR-8 plus 0.5 Al-5 Fe~6 Cr-0.5 Mn--0.06 C 1000 760 ppm Cr, 415 ppm Al

 

179

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

A summary of thermal-convection loop test re-
sults has been prepared to provide a comparison
of the corrosion properties of the various nickel-
molybdenum base alloys tested thus far in fuel
107. In Tables 3.1.20 ond 3.1.21, the alloys are
grouped according to the level of attack found after
500- and 1000-hr tests at 1500°F. As may be seen

in Table 3.1.20, the maximum attack did not ex-
ceed 3 mils for any of the alloys tested, There-
fore it appears that none of the additions seriously
affect corrosion resistance from the standpoint of
total depth of attack. . In particular, additions of
chromium in amounts up to 9% and niobium in
amounts up to 5% had little effect on the depth of

Table 3.1.20. Comparison of Attack of Nickel-Molybden&m Base Alloy Thermal-Convection Loops
Operoted for 500 hr with Fuel 107

 

Alloys with Very Limited Attack
of <1 mil, Nomina! Composition

Alloys with Limited Attack of
1 to 2 mils, Nominal Composition

(wt %) {wt %)

Alloys with Noticeable Attack of
2 to 3 mils, Nominal Composition

(wt %)

 

17 Mo=2 Ti-bal Ni
17 Mo~2 V-bal Ni
17 Mo—3 Nb-—bal Ni
17 Mo--5 Nb—bal Ni 17 Mo=5 Cr—bal Ni
17 Mo—4 Fe-bal Ni 17 Mo—7 Cr-bal Ni
15 Mo-0.5 Al-3 Nb=3 W—bal Ni 20 Mo—7 Cr-bal Ni
16 Mo=9 Cr—bal Ni
17 Mo-2 W=bal Ni
17 Mo—4 W-bal Ni

17 Mo=3 Cr—bal Ni
20 Mo-3 Cr-bal Ni

17 Mo-0.5 Al-bal Ni

17 Mo-2 Al-boal Ni

16 Mo—~2 Al-1.5 Ti-bal Ni

16 Mo—6 Cr—1 Nb—1 Al-bal Ni -

16 Mo—6 Cr--1 Nb—1 Al-bal Ni

15 Mo-5 Cr—3 Nb-3 W-0.5 Al-bal Ni
17 Mo—4 V~bal Ni

17 Mo~1 Al-1.5 Ti-bal Ni

18 Mo—1 Al-1.5 Ti-bal Ni

20 Mo-7 Cr—2 Nb—-1 Fe~bal Ni

20 Mo—-7 Cr~1 Al-2 Nb—1 Fe—bal Ni
20 Mo—1 Nb—-2 Ti—0.8 Mn—bal Ni

 

Table 3.1.21. Comparison of Attack of Nickel-Molybdenum Base Alloy Thermal-Convection Loops
Operated for 1000 hr with Fuel 107

 

Alloys with Limited Attack of
1 to 2 mils, Nominal Composition

Alloys with Noticeable Attack of
2 to 3 mils, Nominal Composition

Alloys with Heaviest Attack of
3 to 5 mils, Nominal Composition

(wt %) (wt %) (wt %)

 

15 Mo-85 Ni

17 Mo—-5 Cr—=bal Ni

17 Mo~-2 Ti-bal Ni
17 Mo—4 W-bal Ni

17 Mo—3 Nb—bal Ni

11 Mo—-2 Al-bal Ni
20 Mo—~1 Nb—2 Ti—8 Mn—bal Ni
20 Mo—~7 Cr—2 Nb-1 Al-1 Fe-=bal Ni

17 Mo—-2 Al=bal Ni

17 Mo—4 Y—ba! Ni

17 Mo-5 Nb-bal Ni

16 Mo—6 Cr—1 Nb-1 Al-bal Ni
16 Mo—2 Al-1.5 Ti-bal Ni

15 Mo—6 Cr-0.5 Al-5 Fe—-0.5 Mn—bal Ni

 

180

 
 

 

 

W

corrosion; however, the presence of these elements
as corrosion products in the fuel increased notice-
ably as the percentage of the element in the base
metal increased. It is interesting fo note that in no
case did alloys containing 20% Mo appear in the
column of noticeable attack (2 to 3 mils), whereas
alloys with similar additives but with a decreased

percentage of molybdenum did show notaceable.

attack.

Increasing the time of the test from 500to 1000 hr

caused the attack to increase 1 to 2 mils, as shown
in Table 3.1.2]. The maximum attack recorded
was 4 mils, and agcin none of the alloy additions
seriously affected the corrosion resistance,

Compatibility of Nickel-Molybdenum Base
Alloys with Molybdenum

It is expected that the use of nickel-molybdenum
alloys as materials of construction in molten fluoride
salt reactors will make possible a considerable in-

crease in the operating temperature attainable in -
The operating temperature of the

such systems,
moderator material will, of course, increase cor-
respondingly. At the temperatures being con-
sidered, moderating materials such as beryliium
oxide and the hydrides of zirconium ‘and yttrium
must be protectively clad with molybdenum. It is
necessary therefore to determine whether the molyb-
denum cladding will be compatible with nickel-
molybdenum alloys in a common flucride fuel
circuit, If it is found that-excessive dissimilar
metal mass transfer occurs between molybdenum
and these alloys, an outer cladding of a nickel-
molybdenum alloy over the molybdenUm w:ll become
an added requirement, .. .

Preliminary compatibility studles of molybdenum :
with nickel-molybdenum alioys are being made by -
~using Hostelloy W thermal-convection loops. with -
molybdenum inserts contained in the hot leg, One
~ such loop has been operated with fuel 107 at o
maximum temperature of 1500°F for. 1000 hr,: anda
second loop has operated at. 1650°F for 517 hr.
The latter loop was terrmnated prior to fhe scheduled >

“ately roughened,

PERIOD ENDING SEPTEMBER 30, 1957

1000 hr when a leak developed in the hot leg.
Metallographic examinations revealed no evidence
of attack or metallurgical change of any kind along
the inner surface of molybdenum samples taken
from the top, middle, and bottom of the inserts
from both loops. Numerous fabrication defects
were found along. the inner surface of the insert
from the loop operated at 1500°F, and the wall
thickness of the insert varied considerably. The
outer surfaces of the molybdenum inserts revealed
no attack, but the insert from the loop operated
at 1500°F had an extremely thin metallic layer
in some areas. The samples from the bottom of the
insert showed less of the deposited layer than did
the middle and top insert samples,

The Hastelloy W surfoces that were exposed
to fuel showed heavy surface pits and shallow
void formation to a depth of 3 mils after the
test at 1650°F, and 2 mils after the test ot
1500°F, The Hastelloy W tested at 1500°F was
attacked in some areas and yet other areas were
completely free of attack. Also, the metallic
layer on the molybdenum insert was found opposite
the area of least attack of the Hastelloy W. The
Hastelloy W tested ot 1650°F had a very thin,
discontinuous line of particles immediately above
some areas of attacked surface. This line appeared

to be associated with foreign phases originally

present in the tubing which withstood attack by
the fuel.

The Hastelloy W cold-leg surfaces were moder-
and in the loop operated at
1500°F there were moderate surface pits and some
general subsurface voids to o depth of 1.5 mils.

- No mass transfer deposits or layers were found.

The COrquion'rate' and appearance of the after-
test samples showed no evidence of serious inter-

~ action between molybdenum and Hastelloy W under
‘these ' test conditions in either loop,
~samples of the molybdenum insert and the Hastel-
- loy W sleeve have been prepared for spectrographic
“examination to further check for alloying between
'r-'these metals. '

However,

181

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

3.2. DEVELOPMENT STUDIES OF INCONEL

STRAIN-CYCLING AND STRESS-RE LAXATION
' STUDIES

C. R. Kennedy' D. A. Douglas

The studies in o program for investigation of the
strain-cycle properties of fine- and coarse-grained
Inconel were continued, with a series of tests
being made to determine the effect of strain re-
versals upon the creep properties of Inconel. Also,
other studies were initiated to defermine the strain-
cycle properties of Inconel weldments, of cor-
burized Inconel, and of reactor-grade beryllium,
The tests to be performed, with the exception of

the tests of carburized Inconel, will provide

engineering data to be utilized in determining the
life expectancy of structural members subjected to
strain reversals, The carburized-Inconel test pro-
gram is being run in conjunction with the alloy
development program (Chap. 3.1, ‘‘Nickel-Moiyb-
denum Alloy Development Studies’’) to determine
whether corburization is deleterious to the strain-
cycle properties of Inconel.

Results of strain-cycle tests of fine- and coarse-
grained Inconel tubular specimens at 1600°F in
NaF-ZrF,-UF, (50-46-4 mole %, fuel 30) are pre-
sented in Fig. 3.2.1, in which the plastic strain-

 

lOri assignment from Pratt & Whitney Aircraft.

UNCLASSIFIED
ORNL~-LR-DWG 25930

S
w

[¢.]

AT
TESTED AT

0%
o S

S m

n

COARSE - GRAINED INCONEL-

PLASTIC STRAIN PER CYCLE,
o

o
o .

0.2

of :
0102 05 1 2 5 t0 2 5 1022 5 100°2 5 0
NUMBER OF GYCLES TO FAILURE, ¥

 

Fig. 321 Stru‘ln-Cycle Properties of Fine- and
Coarse-Grained Inconel Tubular Specimens Tested in
NuF-ZrF l.ll‘-'4 (50-46-4 mole %, fuel 30) ot 1500 and
1600°F. (Se-sr.gt_w.uth-snflnn-).

182

per-cycle (€,) is plotted vs the number of cycles
to failure (I\f Data from previous similar tests?
at 1500°F are also shown for comparison. As may
be seen in Fig. 3.2.1, the data indicate that the
strain-cycle properties ot 1600°F in fuel 30 are
very similar to those at 1500°F and again show
that coarse-grained Inconel is greatly affected by
this environment,

 The evaluation study of the effect of prior strain
cycling upon the creep properties of Inconel is
under way. The test procedure includes strain
cycling the specimens at 0.83% plastic strain per
cycle for different numbers of cycles at 1500°F and
then creep testing the specimen at 7000 psi. The
test results obtained thus far are shown in Fig.
3.2.2 as creep curves of specimen with 0, 100, and
300 strain cycles prior to creep testing. ' The
curves indicate that the creep strength and rupture
life are immediately reduced by the first 100 cycles.
The effect of 300 cycles does not further reduce
the creep strength; however, the total elongation

and rupture life are less than after the 100-cycle

test. The type of plot that will be produced when
the program is completed is shown in Fig. 3.2.3,
where the number of cycles prior to creep testing
is plotted vs the percentage loss of rupture life

 

2c. R, Kennedy and D. A. Douglas, ANP Quar, Prog.
Rep. June 30, 1957, ORNL-2340, p 190.

UNCLASSIFIED -
ORNL-LR-DWG 25934

100
NOQ PRIOR
50
FOR. PRIOR CYCLES
20 RUPTURE FOR 300 PRIOR CYCLES
10
3 500 PRIOR CYCLES
‘é 5
o
=
n 2
4
05
a2
04

 

04 1 10 00 . 1000
TIME (hr) -

Fig. .22, Creep Curves of Inconel Tubes Tested at
1500°F in Acrgon After Prior Strain Cycling at 1500°F.

o

O

 
 

 

 

and percentage loss of creep strength, The creep
strength value used is the time to 1.0% strain,

This is the first attempt to study the interrelation
of strain cycling and simple creep, and these early
results emphasize the importance of the inter-
dependence. Apparently the effects can be con-
sidered to _be additive in that the consumption of

UNCLASSIFIED
ORNL-LR-DWG 25932

 

\

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

8

T 20 \

5

=

i 40 \RUPTURE LIFE

v * .

g . \ TIME TO 4% ELONGATION

W 60 ‘\

< \

S ~—

@ 80 e — STRAIN-CYCLE FAILURE

3 FOR 0.85% «,
100

 

0 200 400 600 800 1000 4200 440Q 1600 800
NUMBER OF STRAIN CYCLES PRIORTO CREEP TESTING

Fig. . 3.2.3, The Effect of Prior Strain Cycling ot
1500°F on the Creep Properties of Inconel ot 1500°F.
Specimens creep tested at 7000 psi in argon.

 

PERIOD ENDING SEPTEMBER 30, 1957

life expectancy by one of these factors also re-
sults in reduced life under the other condition.
This implies that design calculations based on
only one of these conditions will be quite optimistic
for cases where both strain cycling and creep are
present,

A test program has been initiated for evaluating
the strain-cycle properties of Inconel weldments in
both argon and in fuel 30. Two tubular specimens
of conventional size were machined from all-weld-
metal bars, The as-received weld metal is shown
in Fig. 3.2.4, The results of two tests at 1500°F
are listed in Table 3.2.1. The data from the test

Table 3.2.1. Results of Strain-Cycling Tests of
All-Weld-Metal Inconel Specimens at 1500°F

 

Plastic Strai
asfic Strain Number of Cycles

 

 

 

 

Environment per Cycle to Failure
(%)
Argon 1.0 870
FUBI 30 0‘94 45
: / | UNCLASSIFIED
o T Yes3s o o
o
—w-—
T
0
=z
0.02
|0.03
e
-o-
o

 

 

Fig. 3.24. As-Received All-Weld-Metal inconel Specimen. Etchont: 10% oxalic acid. 100X,

183

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

run in argon, as shown in the table, compare
favorably with the standard tubular test data, The
all-weld-metal specimen tested in fuel 30 was
given a 100-hr soak in this environment prior to
testing to allow corrosion to occur. As shown in
the table, the fuel environment seriocusly reduced
the number of cycles to failure. A metallographic
specimen, shown in Fig., 3.2.5, does not indicate,
however, that corrosion caused the premature
failure. In fact, it appears that the major inter-
granular cracks initiated at the inner surface where
no corrosion had occurred. |t theréfore appears
that the loss of life expectancy may have been
caused by defects in the weld or unfavorable
orienfation of the dendritic grains. Future tests
are proposed to check these findings.

During the operation of a reactor with e beryllium
reflector, the beryllium will be subjected to thermal
gradients which wiil resolve into strain, There-
fore, a test program was begun to investigate the
strain-cycle properties of hot-pressed, reactor-
grade beryllium. Strain-cycle tests at 1250°F
were completed and the results are shown in

 

 

 

Fig. 3.2.6 and compared with results obtained for
Inconel at 1500°F, The data shown illustrate the
magnitude of the difference between the strain-
cycle characteristics of a brittle metal and those
of a ductile alloy.

UNCLASSIFIED
ORNL-LR-DWG 25033

-
o
~n

Q’"o'
»w g N o™

o
o . N

PLASTIC STRAN PER CYCLE,

o
n

 

o4
0102 05 { 2 5 1©© 2 8 2 s 2 BixioY
NUMBER OF CYCLES TO FAILURE, ¥

Fig. 3.2.6. StraineCycle Properties of ReactorGrade
Beryllium Tested at 1250°F in Argon Compared with
Those of Incene! at 1500°F,

 

 

 

"UNCLASSIFIED |
v}
= tu)—
I
[ 3]
z
.02
8l lo.o3
>
"o-
S

 

 

Fig. 3.25. All-Weld-Metal Inconel Specimen After Strain-Cycle Testing ot 1500°F in Fuel 30. Etchant: 10%

oxalic acid. 100X, (Secretwith captian).

184

|"

 
 

 

In conjunction with the alloy development pro-
gram, a series of strain-cycling tests of carburized
Inconel at 1500°F have also been run, The results
of these tests are shown in Fig. 3.2.7, where the
data are compared with the data for coarse- and
fine-grained Inconel, These results show that for
the level of carburization of the specimens fested
there was no impairment of the strain-cycle
properties.,

The program under wcly -at the University of
- previous tests.

Alabama for exploring the effect pof thermally-
induced strain cycles is pregressing, A series
of tests in which the specimen is thermally strain
cycled about a mean temperature of 1300°F has
been completed, and the test results are shown in
Fig. 3.2.8 and compared with the mechenicaliy-

UNCLASSIFIED
2 ORNL-LR-DWG 25934

® CARBURIZED

PLASTIC STRAIN PER CYCLE, &, (%)

 

C102 05 4 2 5 10 2 S 1 2 5 10 2 saod
NUMBER OF CYCLES TO FAILURE, &
Fig. 3.27. Sirdifi-Cycle Properties of Carburized
Inconel Tubes Tested at 1500°F in Argon Compured
with Fine- and Course-Grulned Inconel, S

UNCLASSIFIED -
 ORNL-LR-DWG 25935

*
@ MECHANICAL STRAIN CYCLE TEST

- PLASTIC STRAIN PER CYCLE, «5 (%)

 

2 5 2 540% 2 510° 2 5 w02 :mo“

" NUMBER OF CYCLES TO FAILURE N -

Fig. 3.2.8. Thermal Sircin—Cycle Properties of lnconel
Tested at a Mean “Temperature of 1300°F. -

PERIOD ENDING SEPTEMBER 30, 1957

induced strain cycle data obtained at 1300°F under
isothermal conditions at ORNL. As may be seen,

.the correlation of the results is excellent. A

similar series of tests at a mean temperature of
1500°F is in progress.

. Relaxation tests of fine-grained Inconel at 1100°F
have been completed, and the results are shown in
Fig. 3.2.2. The test procedure was described pre-
Yiéygly,a The results obtained this quarter agree
well with the values expected on the basis of the

BIAXIAL CREEP STUDIES
C. R, Kennedy D. A. Douglas

The performance of a particular type of test does
not always reliably indicate the manner in which
material will react under operating conditions,
The uniaxial creep test yields quantities of useful
information; however, the applicability of the data
to multiaxial stress conditions is questionable,
Therefore, in an attempt to achieve a better under-
standing of the behavior of metals under multiaxial
stress conditions, bioxial creep studies were
initiated. This program has as its aim the corre-
lation of the strain rate and the rupture life data
obtained in tests under uniaxial stress conditions
with data obtained under biaxial stress conditions,

A convenient way to vary the principal stress
ratios is to superimpose an axial load on a tubular
specimen, Thus, the circumferential and radial
stresses will be dependent upon the internal pres-
sure, and the axial stress will depend upon both
the pressure and the axial load. The machines for
strain cycling tubular specimens, which were
described previously,? are used for these tests.
This equipment provides a simple means for ob-
taining strain measurements in the axial direction

~ with-the use of a. dial ‘gage fastened to the pull

rod.  Rupture is determined by a drop of internal
pressure, which indicates the first complete crack

- through the wall of the tube. .The strain in the

tangential direction cannot be measured during the
test and can only be roughly obtained after failure.
.. The . tangential ~-stresses - applied- in. ‘various

“ biaxial creep tests were calculated by considering

that the stresses are elastic and that their distri-
bution is represented by fhederne’ equations. * The

 

3¢. R Kennedy, ANP Quar. Prog Rep. March 31,
1957. ORNL-2274 p 216.

J. R. Weir, Jr., ANP Quar. Prog. Rep. Dec. 31 1956,
ORNL~2221, p 246.

185

 
 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 25936

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

- P
A NO PRIOR STRAIN
22,000 = 8,000 psi AND 0.4% STRAIN
B 0.04% PRIOR STRAIN _
T C 008% OR GREATER PRIOR STRAIN
20,000 (] D 045% OR GREATER PRIOR STRAIN
\\\
~ 48,000 ~
16,000 P N
3 "
@ 14000 | 13:800 psi AND 0.05% STRAIN N
W o< 13,700 psi AND 0.05% STRAIN \
- "-n—.__ ! : l
12,000 < 13,600 psi AND 0.05% STRAIN -._,___"' Sw >
T~ T~ "~
r— \K . c
10,000 ~ M.,
[~~~ TR \\ NN
\\-.“ , \ ) \\\ 7
8,000 ™t ™
=~
6,000 X
s B \\..
4,000
002 004 006 008  Of 2 5 100 20 50 100 200 500 1000
TIME (hr)

Fig. 3.29. Reloxation Characteristics of Inconel Tested at 1100°F and Stressed to Produce a Constant Strain.

distribution of the stresses is indicated in Fig.
3.2.10, where 0, 0,, and g, are the radial, axial,
and tangential stresses, respectively. It should be
pointed out that

o, + 0, = constant for all values of r ,

(at)a - (al)b =p,

|

 

(**thin-wall”’ formula) ,

(0)ey = £ 57—

o constant for all values of r .

z

The latter condition results from the axial strain,
€,, being constant for all values of r and o, + o,
being constant. The actual pressure applied
internally was determined by the ‘‘thin-wall’
formula.

It must be realized, however, that, when the
material ceases to behave elastically, the stress

distribution will change. This distribution must

186

~ Fig. 3.2.11 {ref 5).

be calculated before o correlation of the uniaxial
stress properties with the multiaxial stress
properties can be made, The two most important
factors governing the plastic-stress distribution
are {a) the stress-strain-time properties of the
material, such as creep and relaxation, and (%) the
portion of the stress system which directly con-
tributes to plasticity, For example, in a closed
cylinder made of a perfectly plastic material (under
a noncreep condition) in which yielding follows the
Henky-von Mises criterion,

M (o, - av_)2 + (o, - oz)2 + (o, - crt)2 =. 2002 ,

where g, is the yield stress in a simple fension
test, the stresses are distributed as shown in
It may be seen in Fig. 3.2.11
that maximum valves of o, and o, occur on the

 

5a. Nddai, Theory of Flow and Fracture in Sol:ds.
vol 1, 2d ad., McGraw-Hill, New York, 1950.

 
 

 

 

 

 

 

 

 

 

 

 

3

gy
b2+02‘ 7 # |
p———-—

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR-DWG 25937

 

 

 

  
 

 

 

-/ L ]
—— —-

 

 

 

 

Fig. 3.210. Elastic Distribution of Stresses in a Thick-Walled Cylinder Under Interncl Pressure.

outside surface of the pressure vessel.. The

variaticn of 0, across  the . wall “thickness
o), ~ (ot) 1 equafs the magnltude of the internal
pressure,  The g, 0, . and o, vs r curves are equs- 3
_dnstnnt loganthmnc curves. ;’ S | |

“For  &" moderqfely fluck-wailed cylmder (us

opposed to a heaw-wclled cylmder) the distribu-
. tion of all’ stresses across the wall of the cylinder i
" can_be assumed to.vary Imearly rather than. loga-
ST ,nthmscally. "Also, for the condmon of creep, the ’
.. material will probobly ‘behave . more like the per- -
fectly . ploshc ‘material than an eiasflc material,
since the strom rate’ across - the wall of the tube

. lls constanf. : By usmg these - assumphons, the'
stress dlstrlbutlon in the tubular creep specimens

was found to be that shown in Fig. 3.2.12, for

either the Henky-von Mises or the maximum shear

o 3{fi]

o[- e

stress criteria for yielding. The octahedral shear
stress (Henky-von Mises), T _,, or the maximum
shear - stress, T, should occur slmuitaneously
over the wall of the tube. The value of Tocy i

as follows: |

172

The value of T, ., Will depend upon the average

o, relchve to o, und o as follows-

- ©Jov 5
_ (3) _(az > 7, > ar) Tmux = 2 +_4' '

 

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

UNCLASSIFIED
ORNL-LR-DWG 25938

 

 

 

 

 

 

 

)
o P >
: ’ 8
g, 1[
0
o ] F %z
€ A cro/ \/'5'
, |
o
i Ir

 

 

 

 

 

=

b

 

 

Fig. 3.211. Plastic Distribution of Stresses in ¢ Thick-Walled Cylinder Under Internal Pressure.

 

 

 

p b +a
(4) (01>‘7z>0') Tmox =-Z b—a ’
p/ a (az)av
(3 (ot >0, > oz) Tmux =—2(b _a) - ) °

For the specific case in which the inside diameter
of the tube is 0,843 in. and the outside diameter is
0.963 in., the values of T, _, and T, are given
in Table 3.2.2 for the tests reported here, As indi-
cated in Table 3.2,2, a series of tests at 1500°F
and 4000 psi has been completed. The Inconel
specimens were machined from %-in. pipe to have
e 0,843-in. ID end o 0,963-in. OD with a 2.5-in.

gage length, All the specimens were annealed in

188

hydrogen for 2 hr at 1950°F after machining, The
axial creep curves obtained from these tests are
shown in Figs. 3.2.13 and 3.2.14, where the axial
strain, positive or negative, is plotted vs time,
As shewn in Figs. 3.2.13 and 3.2.14, for the tests
in which the stress ratio of g, to o, was greater
than 1/2, the strain was positive, and, for those
in which the stress ratio was less than 1/2, the
sirain was negative, The specimen in which the
stress ratio of o_ to 0, was 1/2 did not experience
strain in the axial direction.” The total strain at
rupture (axial, tangential, and radial) is plotted vs
the stress ratio in Fig. 3.2,15, and the rupture life
of the biaxial creep specimens is plotted vs the
stress ratio in Fig, 3.2.16. All stresses shown in

o>

 
 

 

 

AXIAL STRAIN (%)

 

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL~-LR-DWG 25939

 

 

 

 

Fig. 3.212 Plastic Distribution of Stresses in a Moderately Thick-Walled Cylinder Under Internal Pressure.

UNCLASSIFED

ORNL-LR~-DWG 25940
20

 STRESS (psi), o /oy

05

02 | —

  

04

" TIME (he)

'Fig. 3.2.13. Creep Curves of Inconel Tubes Bilaxially

Stressed and Tested in Argon ot 1500°F.

0 2 5 0 .2 5 0 2 5 w0

UNCLASSIFIED
ORNL-LR-DWG 25941

AXIAL STRAIN ~COMPRESSION (%)

 

- 04
1 2 5 w0 2 5 w02 2 5 10°
- : TIME (hr) )

- Fig; 3.2.14. Creep Curves of Inconel Tubes Biaxiolly

Stressed and Tested In Argon at 1500°F.

189

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

Tan

190

UNCLASSIFIED
ORNL—LR—-DWG 25942

 

20

 

 

 

-
€, _

 

 

 

 

 

 

 

STRAIN, €
/
N\
N
/
\,

r——

 

 

 

 

 

-2

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

—14

—16

—18

_20 *

—4000/4000 —2990/4000 O/acoo  2999%/4000 4990/s000 4990/2000 4990/
STRESS (psi), 0, /0,

Fig. 3.215. Total Stroin at Rupture of Inconel Tubes Creep Tested at 1500°F Under Combined A'x'iul and . |
gential Stresses, ' ‘ LF'

 
 

 

 

 

‘chle. 3.2.2. Octohedral and Maximum Shear Stresses of

Biaxial Creep Specimens Tested at 1500°F in Argon

 

QOctahedral

 

 

PERIOD ENDING SEPTEMBER 30, 1957

the figures were calculated by the elastic ‘“‘thin-
wall’’ formula.

These data show some very interesting aspects
In Fig.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1500°F,

200 -

< , Maximum ‘of high-temperature flow characteristics.
"”’/("_‘ ) Sheor Stress,  Shear Stress, 3.2,13, the creep curves demonstrate that the
o S Toey psl) o T, (psi) axial strain rate is independent of the state of
_ _ stress and dependent only upon the stress in the
4000/0 - - 1890 - 2000 axial direction in those tests in which the stress
4000/1000 - 1400 2030 ratio, 0, fo o, is greater than 1/2, It appears
N » from Fig. 3.2.15 that if the greatest extension
4000/2000 - 1560 £ 2070 -were taken as a measure of deformation, the con-
4000/3000 1790 2100 clusion would be reached that the state of stress
4000/4000 2020 2140 seriously uffects‘the umou'nt of deformation v:'n
, . rupture.  There is no basis, however, for this
5000/4000 1820 2140 selection of o measure of deformation; a better
2000/4000 1750 2140 measuwe is the sum of the three strains without
1000/4000 790 2140 regard .to s.im.' 'l:hus it can be seen tfmt the
' elongation is distributed between the axial and
0/4000 1960 . 2]40 transverse directions; what is lost in one direction
~1000/4000 2210 2500 is gained by the other. If the three strains are
—2000/4000 2520 3000 summed with respect to sign., the quantity is
| reasonably close to zero and indicates the con-
-3000/4000 2880 3500 stancy of volume for high-temperature flow,
- ~4000/4000 3270 4000 A correlation of rupture life as a function of the
stress system has not yet been accomplished.
UNCLASSIFIED
40-00/ ORNL-LR-DWG 25943
0 &
e
4000/, . . —1
2000 . /‘
LA
_ 400%/060 (— —— =/i T
2 2000/4000 prit—d oL "
oo T e Peip o s s
LW e e e e oo
. %4000 [
-1 20004000 / -
 =4000/3000 - — —— = 1. .
e 600~ . 800 . . 4000 4200 1400 . 1600 . - 1800 2000 2200

RUPTURE LIFE {hr)

Fig. 3.2.16. Stress-Rupture Properties of Inconel Tubes .Biuxiully_Sfressed’.and Cr&ep Tested in Argon at

191

 
 

 

 

ANP PROJECT PROGRESS REPORT

At high temperatures where intergranular fracture
is the cause of failure, there are two factors
commonly accepted as contributing to failure:
first, the consolidation of vacancies at boundaries,
and, second, stress concentrations arising at
boundaries because of relative motion of the
crystals, The first factor may well be a function
of the shear stresses, either the maximum shear
stress or the octchedral shear stress (maximum
shear stress deviator). In the case of the second
factor, the greatest principal stress present in the

system appears likely to be a measure of the rate

of propagation of the failure, Neither of these
two -factors alone appears to govern the time to
failure. The correlation is further complicated by
the apparent difference in rupture life in tests in
which the axial stress is greater than the tangential
stress and the rupture life in tests in which the
tangential stress was the greater. This behavior
may be caused by the material being isotropic;
however, the symmetry noted in Fig. 3.2.15 gives
an indication that this is probably not the case.
This same phenomenon has been reported by other
investigators3 in results of tests of tubes biaxially
stressed at low temperature, It is their explanation
that this anomaly is caused by an instability of the
uniform mode of deformation; this same instability
can also gradually develop under biaxial stress
systems in a simple tension test,

RELATIVE TENSILE PROPERTIES OF INCONEL
PLATE AND INCONEL WELD DEPOSITS

R. E. Clausing P. Patriarca

Knowledge of the relative strengths of weld
metal and base metal is required for the accurate
prediction of the behavior of welded assemblies.
In many instances it is necessary only to determine
that the weld is as strong or stronger than the base

material for the conditions of operation. Unless the

weld metal is subjected to unusual cenditions,
such as corrosion, impact loads, or loading which
will result in excessive weld-metal deformation,
failures which occur should be confined to the
base metal or heat-affected zones. If the weld
proves to be weaker than the base material, this
factor must be considered in the design of welded
components, Many times welds may be placed in
low-stress areas or may be mechanically reinforced
to provide the necessary strength.

In certain types of assemblies it is desirable to
match the properties of the weld with the properties

192

of the base material as nearly as possible.. Matched
mechanical and physical properties are important
in structures of uniform cross section which may
be expected to deform during operation. Unequal
strengths in the proximity of the welds will result
in unequal deformation and may result in stress
concentrations that will lead to early failure of
the component, Welds which are more resistant to
deformation than the parent plate may be as un-

- desirable as welds which deform more readily than

the base material,

Although in many instances it is not possible to
test welded assemblies in simulated service con-
ditions, short-time tensile tests provide much
valuable information regarding comparative proper-

ties of welds and parent material, Early tensile

tests of Inconel weldments indicated that, at room
temperature, failures occurred in the weld metal,
whereas, at elevated temperatures, the failures
occurred in the base material. Two specimens
that illustrate these conditions are shown in Fig,
3.2.17; one was tested at room temperature and
the other was tested ot 1500°F, The experimental
work reported here was undertaken in order to
determine the relative strengths and elongations
at failure of Inconel weld metal and base material.

A number of 0,252-in,-dia tensile specimens were
machined from an Inconel weldment, as shown in
Fig. 3.2.18. The weld was deposited as a fillet
by using INCO No. 62 weld wire and PS-1 (ref 6)
welding procedures, Preliminary testing revealed,
as indicated previously, that the specimens were
not likely to fail in the weld metal when tested at
the elevated temperatures of interest, It was
decided, therefore, to reduce the weld portion of
the specimen to a stondard size which would ensure
an all-weld-metal test section. The 0.252-in.-dia
specimens were etched lightly to accurately locate
the weld area, which was subsequently reduced to
a diameter of 0.126 in. with o gage length of 1/2 in.
A drawing of the resulting specimen is shown in

Fig. 3.2.19. Specimens of this type were used to

obtain the data reported here. Base-metal tensile
specimens were also prepared in the some manner.
The base metal had an ASTM grain size of between
6 and 7 and could be classified as fine-grained
material.

 

6ps-1 is the designation of an ORNL. welding specifi-
cation, s

"a

»

 
 

Tested af Room Temperature

 

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
Y-23246

 

 

  

Tested at 1500°F

 

Fig. 3.217, Tensile Test Specimens Machined from Inconel Weldments. Fracture occurred in weld at room

temperature and in base metal at 1500°F,

Y-23310
UNCLASSIFIED
ORNL-LR-DWG 22383R

 

 

 

 

‘fiz-in.-THICK INCONEL PLATE

INCONEL BACKING STRIP

Fig. 3.218, Location of 0. 252.in.-dia’ Tenslle Specl-_'

mens Relative to the Weld Deposh.

- Two weld specimens and one base-metal speci-

" men were tested at each .temperature.. The base
material was tested in the mzll-annealed condmon

and the welded specimens were tested i the as-

welded "condition.. - A cross-head speed of - 005 -
in./min produced a strain rate of 0.1 in./in.smin -

on the l/2-in. gage length. The appearance after
fracture of the 0.126-in. standard specimen . is

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Y-23309
1/2 —in. UNCLASSIFIED
ORNL-LR-DWG 22384R
GAGE LENGTH
04262 0,00t in.

'WELD METAL AREA

Fig. 3.219. Modified, Transverse, Weld-Metal Tensile

 Specimen.

: "illliisfrated in Fig. 3.2.20. Although the test speci-

men geometry used in this investigation was un-
usual, the consistency of the data was quite satis-

factory. The experimental data obtained are listed

in . Table 3.2,3 aond presented graphically in
Fig. 3.2.21. | ,
As may be seen, both the yield and ultimate

- strengths of the weld metal are greater than those
~of  the parent metal at elevated temperatures,
~while the ductility of the weld metal is lower.

If a component is subjected to plastic deformation,

- the 0,2%-offset strength is of paramount importance,

193

 
 

 

 

ANP PROJECT PROGRESS REPORT

 
 
   

. UNCLASSIFIED
: Y-23245

   

 

 

 

Weld Metal at 1600°F

Fig. 3.2220. Typical Reduced-Sectian 0,126-in.-dia. Tensile Specimens.

Table 3.2.3. Tensile Properties of Inconel Weld Metal and Base Metal

 

 

 

T .
ost Specimen Yield Strength at Ultimate Strength Elongation at Failure
Temperature T 0.2% Offset (psi) %
e 0
(°F) ve (psi) ° )
Room Base metal 41,532 99,516 52
Weld metal 50,074 87,982 40
Weld metal 52,670 90,652 40
1400 Base metal 23,790 34,395 52
Weld metal 32,796 44,672 : 45
Weld metal 47,834 40
1600 Base metal 18,870 21,643 ' 68
Weld metal 25,902 27,225 60
Weld metal 26,361 27,624 34
1800 Base metal 9,475 12,500 80
Weld metal 16,617 17,314 56
Weld metal 16,279 16,455 48

 

194
 

 

 

 

 

{l.‘

PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Y-23311
3 UNCLASSIFIED
(x10%) _ ORNL ~LR-DWG 22382R8
80 | | /h'
”
BASE METAL //
O ELONGATION /
A ULTIMATE STRENGTH /0<
20 | O  YIELD STRENGTH //e\;\\ . 70
‘ AT 0.2% OFFSET ot
I‘ _ 7”7
WELD METAL //
B ELONGATION 7
60 |————— A ULTIMATE STRENGTH 7z 60
® YIELD STRENGTH // ‘
AT 0.2% OFFSET 7 »
+ . l
= gLONGX *
O / 0
@ 40 40 e
0 <1
o S
> S
2 Q
g
Ll
-l
5 30 30
=2
)
=
20 20
10 10
oL — — : ‘ — 0
1200 1300 1400 - 1500 1600 - {700 1800 1900 2000

 TEMPERATURE (°F) -

| ng. _3.2;21.- fensllfi Proficrfles of Inconel Weldments.

"lt may ulso be noted rhat at 1400°F the weld meful' :
‘has a yield strength which is approxlmate!y 30%
greater than that ‘of the base material; at 1600°F
- ably. stronger in tension thon the base material at

the yield strength of the weld is 35% greater; and

- at 1800°F it is more than 70% greater, It is readlly' '

apparent, ‘therefore, that the weld should not de-
form appreciably at these temperatures in structures

wuth uniform stress distribution if serwce ‘condi-
tions comparable to the test conditions are assumed.
In summary, Inconel welds appear to be consider-

temperatures - from 1400 to 1800°F, a condition

which may be desirable or undesirable depending
on the stress distribution in service,

195

 
 

 

ANP PROJECT PROGRESS REPORT

BRAZING ALLOYS FOR INCONEL JOINTS

Oxi_ddtion Studies
. G, M. Slaughter P. Patriarca

. The results of static and cyclic oxidation studies,
at 1500°F and at 1700°F, of joints brazed with q
large number of alloys have been reported.” Other

commercial and developmental alloys have since -

become of interest, and the results of similar static

 

7E. E. Hoffman et al., An Evaluation of the Corrosion
and Oxidation Resistance of High-Temperature Brazing
Alloys, ORNL-1934 (Oct. 23, 1956).

studies of these newer alloys are presented in
Table 3.2.4. The static test results are compared
in Table 3.2.5 with the results of cyclic tests.
It is evident that the oxidation resistance of most
of these alloys is excellent.

Corrosion Studies

D. H. Jansen E. E. Hoffman

In an effort to find brfizing alloys that could be
used in fabricating a NaK-to-fuel heat exchanger,
the remainder of a series of high-nickel-content

 brazing alloys have been corrosion tested in the

fuel mixture NaF-ZrF -UF, (53.5-40-6.5 mole %,

Table 3.2.4. Oxidation Resistance of Dry-Hydrogen-Brozed Inconel T-Joints

 

Brazing Alloy Composition (wt %)

Oxidation in
Static Air ot 1700°F*
For 200 hr For 500 hr For 1300 hr For 200 hr  For 500 hr

Oxidation in Static Air at 1500°F*

 

 

 

Handy & Harman No, 93
Handy & Harman No, 91
Handy & Harman No. 82
Handy & Harman No. 72
Rexweld No. 64

Haynes No. 40

Haynes No. 8244

Coast Metals No. 52 with
added boren

Special Nicrobraz No. 50
Coast Metals No, 53 + 1% Li

Nicrobraz No. 130
Colmonoy LM
Commercial GE No. 81
Experimental lron-Base

Experimental Palladium-

Base Self-Fluxing
Nicrobraz No. 150
Nicrobraz No. 60
Nicrobraz No. 10
Nicrobraz No. 45
L. C. Nicrobraz

Cobalt-Modified Coast
Metals No, 52

93,3 Ni—3.5 Si—1.9 B-bal Fe and C Slight Slight Slight Moderate  Moderate
91.3 Ni—4,5 5i—2.9 B-bal Fe and C Slight Slight Slight Slight Slight

82 Ni-4.55i-2.9 B-7 Cr—bal Feond C  Slight Slight Slight Slight Slight
72.5 Ni-5$i-3.5B-16 Cr-bal Fe and C  Slight Slight Slight Slight Slight
72.5Ni-4 5i-3.5 B-15Cr—4 Fe-1C Slight Slight Slight Slight Slight

73 Ni—4 Si—3 B—14 Cr—0.4 C-bal others  Slight Slight Slight Slight Slight
74.5 Ni=3.5 Si—2 B=9.5 Cr—bol others  Slight Slight Shight Slight Slight
91.25 Ni—4.5 $i=3.25+ B~bal others Slight Slight Slight Slight Slight

77 Ni=13 Cr—-10 P + wetting agent Slight Slight Slight Complete Complete
82.1 Ni-4.5 5i-2.9 B~7 Cr-3 Fe-0.5 Moderate  Moderate Severe Severe Severs

others + 1% Li (voids) {voids) (voids) {voids) (voids)
4.5 5i~3.5 B=0.2 C=bal Ni Slight Slight Slight Slight  Slight
55i—3 B—6 Cr=2.5 Fe~0.15C—bal Ni  Slight Slight Slight Slight Slight
10.2 5i-19 Cr=5 Fe-0.25 C~bal Ni Slight Slight Slight Slight Moderate
88.3 Fe-4.8 5i~2.8 B-4.1 Cv Slight Slight Moderate  Moderate  Moderate
Pd-Li Slight Slight Slight Severe 7 Complete
3.4 B-15 Cr-0.15 C-bal Ni Slight Slight Slight Slight Stight
9 Si-15 Mn=bal Ni Slight Slight Slight Slight Slight
11 P=bal Ni Slight Moderate Moderate Corfiplete Complete -
6 P-bal Ni Moderate  Moderate  Moderate ~ Severe  Severs
4.55i-3.5 B=13,5 Cr-4,5 Fe-0.15C-  Slight Slight Slight Moderate  Moderate
bal Ni
Coast Metals No. 52 + 20% Co Slight Slight Slight Moderate  Moderate

 

*Slight, 1 to 2 mils of penetration; moderate, 2 to 5 mils of penetration; severe, greater than 5 mils of penetration,

196

' t‘}

o

 
 

 

b

 

fuel 44) and in NaK (56-44 wt %) in a seesaw
furnace oapparatus at 1500°F for 100 hr. The
specimens used for the tests were brazed Inconel
tube-to-header joints. The alloys, their compo-
sitions, and the results of the individual tests are
listed in Table 3.2.6.

The Nicrobraz No. 130 alloy, which exhibited
the best corrosion resistance to both mediums, is
shown in Fig., 3.2.22, The only bad feature of this
alloy is that numerous voids were formed. This
void formation is attributed to the brazing cycle,
since the voids were present in the as-received
specimens prior fo etching. The composition of
this alloy is nearly the same as the compositions

PERIOD ENDING SEPTEMBER 30, 1957

of two other brazing alloys (Handy and Harman
Nos. 91 and 93) tested previously; the boron con-
tent differed slightly. It appears that with the
alloys containing 91 to 93% Ni, the porosity in
the fillet is related to the amount of boron in the
alloy. No porosity was found in Handy and Harman
alloy No. 91 (1.9% B), very little was observed in
Handy and Harman alloy No. 93 (2.9% B), but, as
shown in Fig. 3.2.22, a considerable amount of
porosity was found in the Nicrobraz No. 130
(3.5% B).

In an effort to lower the melting point of the Coast
Metals No. 52 alloy, 1 wt % boron was added to the
alloy mixture, The additional boron produced a

Table 3.2.5. Comparison of Static and Cyclic Oxidation Resistance of Dry-Hydrogen-Brazed Inconel T-Joints

 

Brazing Alloy Composition (wt %)

Tested at 1500°F for Tested at 1700°F for
500 hr 500 hr

Static Test Cyclic Test Static Test Cyclic Test

 

 

Handy & Harman No. 93
Handy & Harman No. 91
Handy & Harman No. 82
Handy & Harman No, 72
Rexweld No, 64
Haynes No., 40

Haynes No, 8244

Coast Metals No, 52 with
added boron

Special Nicrobraz No. 50

Coast Metals No, 53 + 1% Li

Nicrobraz No. 130
Colmonoy LM
Commercial GE No. 81
Experimental Iron-Base

Experiméntfl Palladium-Base
Seif-Fluxing

7 Nicrobraz No. 150

Nierobraz No. 60
Nicrobraz VN.ca. 0
Nicrobraz No, 45
L.C. Nicrobraz -

Cobalt-Modified Coast
Metals No, 52

93,3 Ni=3.5 $i—1.9 B-bal Fe and C Slight Slight Moderate  Moderate
91,3 Ni=4.5 $i—2.9 Bbal Fe and C Slight Slight Slight Slight
82 Ni~4.5 $i-2.9 B~7 Crbal Fe and C Slight Slight Slight Moderate
72,5 Ni=5 $i-3.5 B_16 Cr—bal Fe and C Slight Slight Slight Stight
72.5 Ni—4 Si-3.5 B=15 Cr—4 Fe—1C Slight Slight Slight Slight
73 Ni—4 $i-3 B=14 Cr—0.4 C—bal others Slight Moderate  Stight Moderate
74.5 Ni~3.5 5i=2 B-9.5 Cr—bal others Slight Shight Slight Slight
91,25 Ni—4.5 $i3.25+ B-bal others Slight Slight Slight Slight
77 Ni-13 Cr—-10 P + wetting agent Slight Slight Complete Complete
82,1 Ni—4.5 $i—2.9 B-7 Cr—3 Fe-0.5 Moderate  Moderate  Severe Severe
others + 1% Li . - (voids) (voids) (voids) (voids)
4.5 $i~3,5 B-0.2 C~bal Ni Slicht Slight Slight Slight
5 $i—3 B=6 Cr—2.5 Fe—0.15 C~bal Ni Slight Slight Slight Moderate
102 $i-19 Cr=5 Fe-0.25 C—bal Ni Slight Slight Moderate  Moderate
' 88.3 Fe—4.85i-2.8 B—4.1 Cu Stight Moderate  Moderate ~ Moderate
Pd-Li Shight Slight Complete Complete
3.4 B15 Cr—0.15 C-bal Ni Stight Slight Slight Stight
98i-15Mabal Ni Stight Siight  Slight Slight
11Pbal Ni Moderate  Moderate  Complete  Complete
6 P-bal Ni B | Moderate  Severe Severe Severe
4.5 5i=3.5 B-13.5 Cr-4.5 Fe-0.15 C=bal Ni  Slight Moderate  Moderate  Moderate
Coast Metals No. 52 + 20% Co Slight Slight Moderate  Moderate

 

197

 
 

 

 

ANP PROJECT PROGRESS REPORT

Table 3.2.6. Resulis of Corrosion Tests in Seesaw Furnaces of Brazed Inconel T-Joints Exposed to
Ne F-ZrF4-UF‘ (53.5-40-6.5 mole %, fuel 44) ond to NaK (56-44 wt %)

Test period: 100 hr
Test temperature: 1500°F in hot zone

 

Tested in Fuel 44 Tested in NaK

 

 

Brozing Alloy Used : i
razing Alloy Use Weight Loss Metallographic Notes Weight Loss Metallographic Notes
(%) (%)
L.ow-Melting Nicrobraz 0.05 Scattered subsurface voids to 0.04 Depleted to o depth of 3 mils;
(83% Ni—6% Cr—5% Si-3% B-3% Fe) a depth of 2 mils; alloy de- subsurface woids to a depth
pleted to same depth of 3 mils »
Rexweld No. 64 0.07 Depleted to a depfh of 3 mils; 0.06 Particle leaching and void
(72.5% Ni-15% Cr—4% Fe—4% Si— subsurface voids to a depth formation to a depth of 2
3.5% B-1% C) of 2 mils mils
General Electric No. 81 0.17 Attack to maximum depth of Uniform attack to a depth of
(65% Ni=19% Cr-10% Si-5% Fe-1% Mn) 7 mils 3 mils
Nicrobraz No. 50 0.08 No attack or depleted alloy Attacked to a depth of 5 mils
(74% Ni-13% Cr-10% P=3% others) found; voids in fillet
Nicrobraz No, 130 0.01 Light, subsurface voids to a 0.03 Depleted to maximum depth of
(92% Ni-4.5% 5i-3.5% B) depth of 1 mil; numercus 3 mils
voids in body of fillet
Coast Metals No, 52 plus boron 0.007 Porous areas in fillet; fillet 0.03 No attack or depletion; fillet
(89% Ni-5% Si—4% B-2% Fe) + 1% B severely cracked severely cracked
Coast Metals No. 53 plus lithium 0.10 Alloy depleted to depth of 0.06 Depleted to o depth of 2 mils
(81% Ni—8% Cr—4% Si—4% B-3% Fe) + 3 mils; voids present 1 mil
1% Li from surfoce; porous areas
in body of fillet

 

UNCLASSIFIED iy o, |’ YRS
NGk ‘9,01

       

BRRINN UNCL ASSIFIED
R Y.22251

 

 

 

- ® -
%(a,*' ‘a ! ”f >
v v o s

Fig. 3.222. Nicrobrax No, 130 Brazing Alloy (92% Ni-4.5% Si=3.5% B) After Exposure to (a) NaF-ZrFiUF‘
(53.5-40-6.5 mole %, fuel 44) and to (b) NaK (56-44 wt %) for 100 hr at 1500°F in Seesaw Furnace Apparatus. Note
the, limited attack and the porous dreas formed throughout the fillet during brazing. Etchant: electrolytic oxalic

acid. 75X. Reduced 16%. {(Seecred-with-cuptiom)
198

 
 

 

 

 

 

 

brittle alloy that cracked during the brazing opera-
tion. The alloy obtained with the boron addition
showed the same corrosion resistance to fuel 44
and to NaK as did the orlgmal Coast Metals No. 52
alloy.

A Coast Metals No. 53 alloy with 1 wt % lithium
added to increase flowability during the brazing
operation also showed essentially the same cor-
rosion resistance as the non-lithium-bearing Coast
Metals No. 53 alloy when tested in fuel 44 and in
NaK in the seesaw furnace apparatus. :

The Handy and Harman alloys Nos. 91 and 93,

which showed good resistance to the fuel 44,8
have been tested in NaF-KF-LiF-UF, (11.2-41-
45.3-2.5 mole %, fuel 107) at a hot-zone tempera-
ture of 1500°F for 100 hr in the seesaw furnace
apparatus. Both alloys showed about the same
resistance to fuel 107 under these conditions.
Corrosion was limited to 3 to 4 mils of depletion
of the minor constituents from the exposed edge

 

8p. H. Jansen and E. E. Hoffman, ANP Quar. Prog.
Rep. June 30, 1957, 0RNL-2340, p 230.

UNCLASSIFIED
Y-23064

,fl “”“

"*(a)g,

 

PERIOD ENDING SEPTEMBER 30, 1957

and the formation of small subsurface voids in the

‘depleted region, as may be seen in Fig. 3.2.23.

‘Coast Metals alloy No. 52 was also tested in
fuel 107 under the same conditions. Subsurface
void formation was more extensive in this alloy
than in the Handy and Harman alloys, as shown in

Fig. 3.2.24,

EFFECT OF GRAIN SIZE ON CORROSION OF
INCONEL BY FUEL 30

J. H. Devan R. S. Crouse

During the examination of the fuel circuit of the
NaK-to-fuel heat exchanger ORNL-1, type IHE-3,
unusually deep intergranular voids were found in
tubes near the tube-to-header joints of the NaK
outlet header.? The Inconel tubing in this area
had relatively large grains because of the heat
freatment required to back-braze the tube-to-
header joints. It was therefore of interest to de-
termine whether the metallurgical changes accom-
panying the grain-coarsening heat treatment of
Inconel could affec_t the corrosion resistance of

 

9G. M. Slaughter, ANP Quar. Prog. Rep. June 10, 1956,
ORNL-2106, Fig. 3.4,36, p 198.

UNCLASSIFIED
Y-22836

T

oy

L5)

Flg. 3.2.23. Bruzlng Alloys .(a) Hufidy & Harman No. 91 and (b) Hondy & Harman No. 93 After Exposure to

- NaF-KF-LiF-UF (11,2-41-45.3-2.5 mole %, fuel 107) at 1500°F for 100 ht in Seesaw Furnace Apparatus, Etchant:

10% oxalic acid, electrolytic. 150X. Reduced 11%. Sosretawith-eeption)

199

 
 

ANP PROJECT PROGRESS REPORT

Y "y
Xml

! i ity SR i

i 5\] l.‘_b\ ) im‘ ST
g J\ \ i Wn
N

 

BEEREN (INCLASSIFIED
B Y.22835

 

Fig, 3.2.24. Cocst Metals Brazing Alloy No. 52 (89% Ni=5% Si-4% B-2% Fe) After Exposure for 100 hr ot
1500°F to NaF-KF-LIF-Ufi (11:2+41045.3+2.5 mole %, fuel 107) In a Seesaw Furnace Apparatus. Etchant: 10%

oxallc acid, electrolytices 150X: (Sacredwrith-esption)

this alloy in molten flucrides, Accordingly, an
Inconel forced-circulation loop, 7641-50, was
prepared with hot-leg sections which had been
annealed at 1940°F for 2 hr in hydrogen. The
annealing treatment was intended to produce grains
similar to those formed during the furnace brazing
of heat exchanger and radiator tubes, The loop
was operated with NaF-ZrF -UF , (50-46-4 mole %,
fuel 30) for 1000 hr with @ maximum fuel-to-metal
interface temperature of 1600°F and a minimum fuel
temperature of 1300°F. Under similar conditions,
a standard Inconel loop (ASTM. grain size 4-8)
would show 5 to 7 mils of general and intergranular
subsurface void formation,

The sizes of the grains of the annealed sections,
measured in both before-test and after-test samples,
were very random, ranging from ASTM 5 to greater
than ASTM 1 within a single sample. After-test
samples tcken from the point of maximum wall
temperature revealed heavy general and inter-
granular void formation to a depth of 5.5 mils. In
addition, a few scattered intergranular voids were
visible to @ depth of 12 mils, as shown in Fig.
3.2.25. Thus preferential grain-boundary attack,

200

. UNCLASSIFIED
- T-12848

 

ERE

¥
mer!a

EEFEEREE

 

 

 

Fig. 3.225. Intergranuvlar Voids Formed During
Operaticn of Inconel Forced-Circulation Loop 7641.50
for 1000 hr with Fuel 30. Heated sections of loop had
been annealed at 1940°F for 2 hr prior to operation.
Etchant: coqua regia, 250X, Reduced 29%. (beerat
with-copticn),

 
 

 

 

4%

such as that found in the exariination of heat
exchanger units, occurred to a limited extent in
the coarse-grained sections of this Inconel locop.

Thermal-convection loop tests were run several
years ago to evaluate grain-size effects, but little
difference in the depth of attack was noted between
as-received and annealed specimens.'® In re-

 

10
G. M. Adamson, ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL 1771, p98- ; br eb

PERIOD ENDING SEPTEMBER 30, 1957

viewing " the previous tests, however, it appears
that subsurface voids were much more concen-
trated in the grain boundaries of coarse-grained
tubing than in the grain boundaries of fine-grained
tubing, While the depth of the intergranular voids
in the coarse-grained tubing did not greatly exceed
the depth of voids in the grain matrix of fine-
grained samples, such tests indicate the possibility
of relatively deep attack under certain conditions
of grain-boundary orientation in coarse-grained
material,

201

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

3.3. WELDING AND BRAZING STUDIES

P. Patriarca

EXAMINATION OF ALL-WELDED INCONEL
IMPELLER THAT CRACKED IN SERVICE

G. M. Slaughter

The sump-type centrifugal pump (model DANA),
which was used for circulating the fuel mixture
NaF-ZrF“-UF‘ (50-46-4 mole %, fuel 30) in small
heat exchanger test stand SHE-C, was found,
during a routine examination aofter termination of
a test, to have a cracked impeller. The all-welded
Inconel impeller had operated for 2350 hr in the
temperature range of 1200 to 1500°F, with the
bulk of the operation at 1300°F. The location
and the extent of the cracks in the welded vane-
to-plate joint are shown in Fig. 3.3.1, and o
photomicrograph of one of the cracks is presented
in Fig. 3.3.2.

The impeller vane-to-plate welds were fabricated
according to the weld joint design shown in

 

UNCLASSIFIED
Y-22570

   

F'g- 3-30]0
Showing Location and Extent of Cracks Found After
Operation for 2350 hr in the Temperature Ronge of

1200 to 1500°F with Fuel 30. (SecreTwith—cupriom

AlleWelded Inconel Pump Impeller

NCLASSIFIED

Fig. 3.3.2. Photomicrograph of Weld Fracture Shown in Figs 3.3.1. 100X. Etchant: electrolytic oxalic acid,

202

 
 

 

¥

Fig. 3.3.3. It is evident that the fabrication
procedure provided only o limited joint length
where the vanes were welded to the top plate.
Furthermore, the joint design and lack of accessi-
bility permitted only very shallow weld pene-
tration. This condition is exhibited in the upper
left of Fig. 3.3.4, which shows a polished and
macroetched section of the impeller. A modified
fabrication procedure is now being used in which
the joints are brazed for additional reinforcement
and to improve reliability.

BRAZED-JOINT CRACKING TESTS
G. M. Slaughter |

Preliminary tests have been conducted to de-
termine the extent of braze cracking during the
fabrication of high-conductivity-fin NaK-to-air
radiators at the York Corp. The cracks have
resulted from the transverse shrinkage and dis-
tortion of the tube sheets during the deposition
of the longitudinal welds of the headers.

As a means of simulating the conditions under
which the cracking occurred, tube-to-header broze

FILLET WELD AS MUCH AS POSSIBLE
FROM OUTSIDE,; WELD LENGTH X
AND Y TO BE SYMMETRICAL

FOR ALL VANES

 

SECTION BB

'SECTION BB IS TYPICAL
‘OF WELDS AT TIPS OF
5 VANES FOR
APPROXIMATE
DISTANCE SHOWN

 

PERIOD ENDING SEPTEMBER 30, 1957

fillets were prepared and then cracked by bending
the headers. The degree of deformation obtained
upon bending is apparent in Fig. 3.3.5, which
shows the samples before and after bending.
Bending of the header to the extent indicated
produced cracks which were considered to be-
visually representative of those found in the
radiators fabricated at the York Corp.

'Metallographic examinations of the cracked
joints revealed the cracks to be in the braze fillet.
In no case did the cracks penetrate into the tubes
or the tube sheet. A typical crack in the braze
fillet is shown in Fig. 3.3.6.

‘Sample joints were then rebrazed with and
without the use of additional Coast Metals No. 52
brazing slurry. The cracks appeared to completely
heal in. all cases. The addition of slurry to the
fillet before rebrazing appeared to be advan-
tageous, however, in that it permitted a more
uniform fillet and provided an additional source

" of alloy for healing the cracks.

UNCLASSIFIED
ORNL—LR—DWG 25944

 

 

   

 

SECTION AA

Fig. 3.3.3. Weld Deiullfi of Impeller Vaneoto-Plate Joints.

203

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

BN UNCLASSIFIED
Y-22571 |

-

Figs 3.3.4. Vane-to-Plate Welds of All-Welded Inconel
Impeller Fabricated According to Design Shown in
Fig. 30303-

FABRICATION OF ART FUEL FILL-AND-DRAIN
TANK

E. A. Franco-Ferreira

G. M. Slaughter

As reported previously,! a study of some of the
problems associated with fabrication of the ART
fuel fill-and-drain tank is under way. Specifically,
tests are being run to establish suitable welding
and brazing procedures for making the tube-to-tube
sheet joints in the tank heads.

As a result of the high degree of restraint
exerted by the 1)y-in.-thick tube sheet, the tube-
to-tube sheet welds, when made with the stendard
fusion procedure used for thinner tube sheets,
have consistently exhibited root cracks of the
type illustrated in Fig. 3.3.7. In order to produce

 

‘E. A. Franco-Ferreira, ANP Quar. Prog. Rep. June
30, 1957, ORNL-2340, p 225.

204

 

‘UNCLASSIFIED -
Y.23346

 

 

Fig. 3.3.5. Simulated Tube-to-Header Joints for
Braze-Cracking Tests.

joints free of such root cracks, a low restraint,
trepanned type of joint was designed. The joint
design and the welding procedure for obtaining
crack-free welds are presented in Fig. 3.3.8. A
machine welding method is preferred for the
fabrication of these joints in order to assure
consistency and reliability. Welds made according
to the prescribed procedure are shown in Fig.
3.3.9. |

Brazing experiments were also conducted in
order to determine the proper amount of preplaced
brazing alloy and the type of preplacement pro-
cedure to be used, as well as to evolve a suitable
brazing cycle. Calculations were made to de-
termine the amount of Coast Metals brazing alloy
No. 52 required to fill a 0.003- to 0.004-in.-wide
onnulus  between the %-in.-OD tube and the
I‘/z-in.-thick tube sheet. The results indicated
that two brazing rings, each weighing approxi-
mately 1.3 g, would be required for the brazing
of each tube. :

The configuration of the specimens, that is,
the thick tube sheet and the comparatively thin-

" walled tubes, posed additional problems.- When

the assembly is being heated to the brazing
temperature (1920°F), the temperature of the
massive tube sheet lags behind the temperature
of the tubes. Several joints may be seen in

 
 

 

i
i
1

 

2=
e

 

 

 

PERIOD ENDING VSEP TEMBER 30, 1957

UNCLASSIFIED

 

 

 

 

Fig. 3.3.:6. Typica!l Crack in Braze Fillets 100X. Ase-polished.

Fig. 3.3.7. Root Crack in o TubestosTube Sheet

Joint Made Under Conditions of High Restraint. Etchant:
electrolytic oxalic ccids 50X.

Fig. 3.3.10 in which the brazing alloy failed to
fillet because the ring melted around the hot tube
ond did not wet the cooler tube sheet. The use
of radiation shielding, which minimized direct
radiation on tubes from the heat source, produced
the results shown in Fig. 3.3.11. Such shielding
definitely improved the reliability of filleting, but
the use of shielding would be impractical for the
fabrication of the full-scale heads. Therefore a
relicble brazing methed has been developed which
depends upon keeping the brazing alloy away

~from the tubes until the tube sheet has become

hot enough .to melt the alloy. In order to ac-

‘complish this, the brazing alloy is contained in
~annular sumps machined in the tube sheet around
each tube. ~ When the tube sheet reaches the

brazing temperature, the brazing alloy flows into

“the ‘space between the tubes ond tube sheet
~through three holes which are drilled from each

sump into the tube hole. A detail of the sump

-arrangement.is shown in Fig. 3.3.8.

"Both dry powder and cast rings were used
successfully for the preplacement of the brazing
alloy on the test specimens. When powder was
used, the sumps were filled level with the tube

205

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 25945

~ JOINT .DESIGN: WELDING SEQUENCE :

t = 0060 in., TREPAN DEPTH = 2/
TUBE, 0.625-in. OD

  
   
 

 

    
    

!
”“I ) MAXIMUM
- _ OVERLAP
- . MINIMUM
£ CURRENT

RS

TAPER

{

Y5 in. EXPAND -]

DRILL AND REAM TO 0.004-
1 0.001-in. CLEARANCE
ON DIAMETER

 

 

 

 

ALLOY FEED HOLE

SUMP (2/ DEEP) (¥/32-in. DRILL)

BRAZING ALLOY SUMP DETAIL

WELDING CONDITIONS :

MACHINE MADE WELD
CUP SIZE — ¥g-in. OPENING

ELECTRODE—- THORIATED TUNGSTEN %3, in., POINTED
ELECTRODE PROJECTION — ¥4 in.

ELECTRODE SWING — 0.640-in. DIAMETER

ARC DISTANCE — 0.050 in.

ARGON FLOW THROUGH TORCH—45 TO 50 cfh
HELIUM BACKUP GAS FLOW—15 cfh

ARC CURRENT—58 TO 60 amp

ELECTRODE TRAVEL SPEED (PERIPHERAL)—5 in./min

Fige 3.3.8. Joint and Welding Procedure Details for Tubé-to-Tube_Sheef Welds of ART Filleand«Drain Tpnk-
(Sectatunithcaption)

206

 
 

 

 

UNCLASSIFIED
Y297

 

Test No. 4
RN

Fig. 3-3.9; Tube-to-Tube Sheet Welds Made by Using
the Optimum Welding Procedure Prescribed for Such
Welds in the Heads of the ART Fue! Fill-and=-Drain

Tank. @eersiauith caption)

UNCLASSIFIED
Y-215%

 

Fig. 3.3.10. Tube-to-Header Joints Made ot o Rate
of Rise to Brozing Temperafure of 300° F/hr Without
Radiation Shielding.

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
Y-21 %1

 

Fige 3.3.11.  TubestoeHeader Joints Showing Good
Fillets Obtained By Using Radiation Shielding.

sheet, and then the powder was wetted with
Nicrobraz cement. The use of rings simplifies
the prepfacement procedure, but it is desirable
to prime the rings with a small amount of powdered
alloy wetted with Nicrobraz cement.

The brazing cycle finally decided upon provides
for a rate of temperature rise of 300°F/hr to the
brazing temperature of 1920°F, 1 hr at that
temperature, and cooling ot a rate of 300°F/hr.
The relatively low rate of temperature rise is
required to avoid brazing alloy liquation before

‘the massive header sheet has reached the brazing
temperature.

CORRELATION OF RADIOGRAPHIC
- AND METALLOGRAPHIC DETERMINATIONS
OF POROSITY IN TUBE-TO-HEADER WELDS

G. M. Sluughter
The determination of porosity in tube-to-header

welds by radiography requires the use of carefully
developed procedures ond skillful interpretation

~ of the radiographic film. - Further the geometry of

tube-to-header joints : containing - small-diameter
thin-walled tubes is so complex that a precise
correlation of the results of radiographic and

207

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

metallographic examinations would be extremely
beneficial. A preliminary study has therefore been
conducted to obtain the basic information needed
for such a correlation.

For this investigation, approximately 60 welds
were made by inert-gas-shielded arc welding of
short lengths of Inconel tubing, 9’“ in. OD with
0.025-in. walls, into holes drilled on the circum-
ference of sections of 3-in. sched-40 Inconel pipe.
Each weld was given o designation before it was
radiographed, and, ofter the films had been
carefully examined, the locations of pores were
noted on the weld specimens. Five welds that
contained defects were then removed and the
pores were located by alternately grinding and
polishing, with a microscope being used frequently
as a visual aid.

Gross porosity was found in every case where
it was revealed in the radiographic examination.
The sizes of the voids detected are summarized
in Table 3.3.1. It may be seen that the pores in
these five welds were relatively large ond thus
were readily detectable by careful radiographic
examination. A similar investigation will be

Table 3.3.1. Sizes of Voids Detected in Tube-tc-Header
Welds by Radicgraphic Examination

 

Void Dimensions®

Weld No.  Defect Designation
(in.)

 

1 P-3-1-Wé 0.013 x 0.011
2 P-3-1-W8 0.012 x 0.008
3 P-9-1.W§ 0.009 x 0.009
4 P-16-2-W12 0.013 x 0.010
5 P-9-2-W12 0.010 x 0.008

 

*Void diameter was measured in two directions with a
Filar eyepiece.

conducted in which finer pores, in the range of
0.003 to 0.005 in., will be studied in order to
determine the limitations of detection by the
radiographic method.

FABRICATION OF SODIUM-TO-NcK HEAT
EXCHANGERS
G. M. Slaughter

The brazing of the two sodium-to-NaK heat
exchangers recently received for installation in

208

the ETU north head was observed at the
Ferrotherm Corp. The brazing equipment, pro-
cedures, and general furnace cycles utilized were
described previously in connection wuth the fabri-
cation of radiators at the York Corp.?2 '

During the brazing of heat exchanger No. 1, the
exit dew point from the retort was approximately
~100°F throughout the brazing cycle. The spread
of the readings of the four thermocouples at the
start of the hold peried was 6°F, from 1914 to
1920°F, and at the end of the hold period was
11°F, from 1921 to 1932°F. All thermocouples
reached 1920°F ot some point in the brazing
cycle, but the moximum temperature attained was
1932°F.. During the brazing of unit No. 2, the
exit dew point from the retort was approximately
~95°F throughout the brazing cycle. The spread
of the four thermocouples at the start of the hold
period was 5°F, from 1918 to 1923°F, and ot the

end of the hold period it was 4°F, from 1920 to

1924°F. The maximum temperature aflumed was
1927°F.

Upon opening the retorts, the units were found
to be clean and bright, oand good filleting was
evident at all tube-to-header joints. Minor brazing
alloy run-off was noted at one header, but it is
thought that this could be minimized in the future

by the use of commercially available stop-off

compounds. The two job samples, which were
brazed at the saome time as the complete units,
were evaluated at ORNL and found to be satis-
factory.

INVESTIGATION OF MATERIAL FOR
RADIATOR HEADERS

G. M. Slaughter

Several NaK-to-air radiator headers have been
completely machined at York Corp. and cre ready
for welding to the tubes. York has reported,
however, that these headers were fabricated from
material which is difficult to weld. The welds
made thus far have cracked and have been found
to be porous. Further, there has been inconsistent
penetration.

Consequently samples of the header material
were transmitted to ORNL for examination and
evaluation.  Preliminary conventional machine

 

26, M. Slaughter, ANP Quar. Prog. Rep. June 30,
1957, ORNL-2340, p 223.
 

 

 

 

 

i
,!

welding tests were conducted, and distinctly
different welding behavior from that of normal
Inconel was evident. The tube-to-header welds
of the York material had a *‘grainy’’ appearance.
Low-power examination and dye-penetrant in-
spection also revealed cracks in the welds.

In an attempt to determine the reason for the
observed differences in welding characteristics,
o metallographic comparison was made between
the York material and some lnconel plates of
known satisfactory welding characteristics. All
samples of the questionable material were found
to contain a quantity -of inclusions (possibly
oxides) distributed randomly as stringers. The
samples of known satisfactory ~characteristics
were relatively free of inclusions.: The distri-
bution and appearance of the stringers of the York
material are shown in Figs. 3.3.12 end 3.3.13,
and a typical area in material which has exhibited
satisfactory welding behavior is shown in Fig.
3.3.14.  The inclusions which appear .as dots
are probably titanium nitrides ond were evident
in all samples examined. Such inclusions are

apparently unobjectionable. Efforts to positively

establish the difference between these materials

 

 

PERIOD ENDING SEPTEMBER 30, 1957

by conventional chemical analysis were unsuc-
cessful. Welding tests are being conducted on
other lots of inspected Inconel pipe in an effort

to ' provide the necessary quantity of suitable

material for York.

FABRICATION OF A SMALL SEMICIRCULAR
HEAT EXCHANGER

E. A. Franco-Ferreira

- The construction of a semicircular heat ex-
changer for experimental tests was undertaken. A
description of the heat exchanger and the test
objectives was presented previously.3 The initial
planning and procedura! experimentation have been
completed, and it is expected that the first of
the two heat exchangers being fabricated will be
available soon.

The initial experiments, which were made in
order to devise suitable welding and brazing
procedures for the tube bundle, were performed
on scrap material. An example of the experimental

 

3J. C. Amos et al., ANP Quar. Prog. Rep. March 31,
1957, ORNL-2274, p 43.

INCHES

0.02

 

 

0.03

 

 

 

Fig. 3.3.12. Stringers in Inconel Header Mnferriulk Available at York Corp. Asepolished. 100X.

209

 
 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fige 3.3.14. Inconel Which Exhibited Satisfactory Welding Characteristics. Note relative freedom from large
stringerss Asepolished. 100X.

210

 

 

 
 

 

 

.n

results is shown in Fig. 3.3.15. Work was not
started on the actual heat exchanger until all
procedures had been carefully planned. The
components of the heat exchanger are shown in
Fig. 3.3.16. All parts were minutely checked
before the assembly sequence was started.

Experimentation had ~ shown that successful

assembly of the tube bundle into the tube sheets,
and subsequently, into the channel would depend

‘primarily upon extreme accuracy in- the bending -

of the tubes. Consequently the jig shown in
Fig. 3.3.17 was built so that ‘the tubes could be

individually checked for conformity to the required
dimensions. - Thus, each tube in the heat ex-

changer was hand-picked for assembly by checkmg

_ it as shown in Fig. 3.3.18.

In thé course of os_sei'nblyk of the tubes into the

tube sheets it was found that small inaccuracies

in bending, even though within the specified
tolerances, gave rise to undue stresses under the
restraint of the comblike tube spacers. As a
result, a decision was made to stress-relieve the
tube bundle in the channe!l before welding and
brazing the tube sheets. The tubes were as-
sembled into the tube sheets, and the tube bundle

 

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
PHOTO 40469

   

Fig. 3.3.15." Prelimincry Welded and Brazed Semi-
circular Heat Exchanger Test Headers As-polished.
100X+ Reduced 31%.

UNCLASSIFIED
PHOTO 40670

Fige 3-3-16. Components for Semicircular Heat Exchanger.

211

 
 

 

 

ANP PROJECT PROGRESS REPORT

Fig. 3.3.18. Tube Being

 

Checked in Jig.

UNCLASSIFIED
PHOTO. 41101-

UNCLASSIFIED
PHOTO. 41100

&

 
 

was placed into the channel with shim spacers
between tubes. The resulting assembly, shown
in Fig. 3.3.19, was sfress-relievse,cvl_. in dry helium
at 1500°F for 1 hr. After the stress-relief treat-
ment, the comb spacers were welded into the tube
bundle, ond the tube sheet welds were made.

Assembly methods dictated that the tube ends
protrude randomly beyond the tube - sheets, as

shown in Fig. 3.3.20. Since it is of prime im-

portance that the distence between the face of
the tube sheet and the first bend in each tube be
known accurately, the distance that each tube end
protruded beyond the tube sheet was measured
accurately with a depth micrometer before the

ends were cut off flush with the tube sheet. This

measurement was correlated with the over-all
measurement of each tube before assembly, which
was obtained by the photo-lofting technique illus-
trated in Fig. 3.3.21. The length of the cut-off
portion was subtracted from the original length
to obtain the required dimension.

The NaK inlet and outlet nozzles were welded

to the tube sheets before the tube sheets were.

brazed. The assembly and the preplaced Coast
Metals brazing alloy No. 52 rings are shown in
Fig. 3.3.22. The tube sheets were then back

brazed, one at a time, in a special, curved retort

PERIOD ENDING SEPTEMBER 30, 1957

which was placed in @ pit furnace, as shown in
Fig. 3.3.23. The remaining components are to
be assembled by welding.

UNCLASSIFIED
PHOTO 41103

 

Fig. 3.3.20. Tube Sheet with Tube Ends Protruding
ot Random Lengths.

UNCLASSIFIED
PHOTO 41102

 

Fige 3.3.19. Assembled Heat Exchanger Jigged for Stress Relieving.

213

 
 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED:
PHOTO 40885 -

 

 

 

 

 

 

 

* .
Ve D

P .%&w;wm

R
3 .m...m ! SR

; z,
i

 

 

 

17
3k

o
51

 

 

 

  

Method Used for Accurately Measuring

Dimensions of Tube Bends.

Fig- 3-3021.

 
       

Header Assembly with Nozzles and

3.22.

3.
Preplaced Coast Metals Brazing Alloy No.

5
-

 

52 Rings.

UNCLASSIFIED

  
  

PHOTO 40889

 

I S Al A

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fige. 3.3.23. Special Curved Retort for Brazing the Tube Sheets of a Semicircular Heat Exchanger.

214

 

 
 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

'3.4. CORROSION AND MASS TRANSFER STUDIES

CORROSION OF YANADIUM IN FL-UORIDE FUEL
AND IN LITHIUM

D. H. Jansen  E. E. Hoffman

The possibility of usmg vanadium as a con-

stituent of a container material for the fuel mixture

NaF-KF-LiF-UF (11.2-41-45.3-2.5 mole %, fuel
107) ond for lithium is being studied. In order
to investigate the effect of fuel 107, o vanadium

capsule clad with type 347 stainless steel was

filled to one-third its- volume with the fuel and
inserted in a seesaw furnace apparatus for 100 hr

with the hot zone of the capsule at 1500°F,

A considerable amount of mass transfer was

found after the test in the cold zone of the

capsule, as shown in Fig. 3.4.1. The crystals
were found by chemical analysis ‘to ‘contain, in -
addition to vanadium, 0.95 wt % Fe, 0, 21 wt % Ni,
and <0.50 wt % Cr.  Thus it appeared that ‘the
vanadium . capsule material was - not- - pure. An’
analysis of the material from which. the cnpsule, -

was fabricated showed traces of the same . im-
purities with the same relative magnitudes. ~The

fuel picked up approximately 700 ppm of voncdlum :
during the test. The surface urec-to-volume ratio

of the capsule was 10 in.2/in.3.

Thickness measurements of the . capsule -
dicated uniform removal - of vanadlum to @ depth“
of 1.5 mils from the hot zone of ,ihe capsule, -

UNCLASSIFIED
TUess

TYPE 347

 

Fig. 3.4.1. Cold VZon;- of ~Vdnd_diufn' Capsule Exposed .
to Fuel 107 for 100 hr in Seesaw Fumace Appurcfus with
e Hot-Zone Temperoture of 1500°F and a Cold-Zone.

Temperature of 1200°F, 5X. Reduced 35%. ('561:?!1'
" rion) :

and there was intergranular attack to a depth of

1 mil, as shown in Fig.‘3.4.2l.

A static vanadium CQpédl‘é.-"-f&i‘ll.éc‘i' with lithium
was  also tested. for' 100 hr at 1500°F. Slight

surface roughenmg on the walls of the bath zone

~ was ' the extent of the attack found after this
test,

| ";.MOLYBDENUM AND NIOBIUM IN CONTACT
WITH LITHIUM

E. E. Hoffman

Mdiybdenum and niobium were tested in contact

with lithium in o seesaw-furnace test for 500 hr
with o hot-zone temperature of 1500°F and a
cold-zone temperature of 950°F. The test capsule

‘,_t:‘q_ns'is‘ted of & molybdenum or a niobium cup
enclosed in a jocket container of type 316 stain-

less steel, as shown in Fig. 3.4.3. The two
types of copsules were loaded with lithium,

placed in seesaw-furnace apparatus, and cycled
" at arate of 1 cpm. The lithium bath was alter-
_ ‘thquf in contact with the refractory metal at
approximately 1500°F and with the stainless-
“steel cold zone at approximately 950°F. Following
- the test, the lithium was removed and the copsules

were examined.

'Mefu'llographfc inspection revealed no attack

. 7 ... on either the molybdenum or the niobium hot-zone
- VANADIUM -

walls, as shown in Figs. 3.4.4 and 3.4.5. Type

| 34 2_5-31'6 stainless-steel specimens from the cold-zone
. STAINLESS - -
STEEL - .

“walls showed less than 0.0005 in. of surface

‘roughemng and less than 0.0005 in. of scattered
- mass-transfer crystals. These “small crystcls
- were similar to crystals seen in previous experi-
- _ments with lithium in all-stainless-steel systems.

Microspark spectrographic examination of the
stainless-steel - wall indicated no detectable in-

~crease of molybdenum or niobium in comparison
_with standard samples of type 316 stainless steel.

The results of these: tests indicate that under
the conditions of these experiments, there was
little tendency for dissimilar-metal mass transfer
or temperature-gradient mass transfer of molyb-
denum or niobium to the stainless-steel wall.

215

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT.

 

 

 

L UNCLASSIFIED |
Y.22939 ¢

' 23 @

Wl

I...

£

<

QU2

G033

_094

005

_.'gwcé

>

O

O—.

W

NN

 

 

Fig. 3.4.2. ' Intergranular ‘Attack In Hot Zene (1500°F) of Vonadiom Seesaw Test Capsule After Exposure for
100 hr to NaF-KF—L!F-UF‘ (11.2-41-45.3-2.5 Mole %, Fuel 107), 500X. Etchant: NH,OH + H202.- Seeretwitr
taptienl

NPT  TUNGSTEN-NICKEL-COPPER ALLOY
ORNL-LR-DWG 25946 |N SOD"JM ’

W. H. Cook E. E. Hoffman

Two specimens of the alloy Mallory 1000
(90% W-6% Ni-4% Cu), manufactured by P. R.
Maliory and Company, Inc., were exposed to-
sodium for 100 hr in the hot zones (1500°F) of
Inconel containers inserted in a seesaw-fumace
apparatus cycled at a rate of 1 cpm. Maliory 1000
is being considered for use as a transition layer
between tungsten - carbide—cobalt cermets and
Inconel in disk and seat brazed joints of sodium
or NaK valves for use at high temperatures. One
of the specimens was tested as-received and the
other was heat treated at 1920°F for ‘é hr in a
furnace with heating and cooling rates of
300% /hr. . This heat treatment simulated a fab-
rication operation. that would  be involved in
construction of the valves. '

AR ' Metailographic examinations indicated that the
Fig. 3.4.3. Seesaw Fumace Test Capsule. specimens were equally attacked. The untested

  
  
      

TYPE 346 STAINLESS STEEL
{1in. SCHEDULE 40)

COLD~ZONE {950°F)
THERMOCOUPLE

216

 
 

 

 

 

 

»

PERIOD ENDING SEPTEMBER 30, 1957

8l UNCLASSIFIED
e, Ya23334. |

o

   
  

 

 

X00§—|

 

|e——— YouI00°0

Fig. 3.4.4, Wall of Molybdenum Capsule Following Exposure for 500 hr to Lithium in the Hot Zone of a Seesaw
Furnace at 1500°F. Etchant: 50% H,0, + 50% NH,OH. 500X. {Seeretsmiticuption)

X008 ——=|

|e—— youig00°0

 

‘Fig. 3'4_5. Wall -of_i Niobium :Cupmle F_o“dwipg Exposure for 500 hr to Lithium in the Hot Zone of a Seesaw
Furnace ot 1500°F. Etchant: HF + H,0 + H,S0, + HNO,. 500X. (Sestetwith-capTion)

- ' 217

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

and tested specimens are shown in Fig. 3.4.6.
‘Sodium had completely penetrated each specimen
{nominal dimensions: :/?2 X ‘4 X '/2 in.). The -

untested specimens had pore-space areas that
constituted approximately 1% of the microscopic
field of view, while the pore-space areas of the
tested specimens constituted approximately 4%
of the field. The attack appeared to have enlarged
existing pores and created new ones. Each speci-
men had a weight loss of 1.3%. Apparently, the
quantity of copper present in Mallory 1000 is
sufficient to cause this material to be of doubtful
value for long exposure to sodium or NaK at

1500°F.

TUNGSTEN CARBIDE- AND TITARIUM
CARBIDE-BASE CERMETS IN SODIUM

- W. H. Cock E. E. Hoffman

Three types of tungsten carbide- and titanium
carbide-base cermets were exposed for 100 hr
to sodium at 1200°F in Inconel capsules inserted

 

 
 
  
 
  
     
  
 
 

  

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_in a seesaw-furnace apparatus, The designations

and compositions of the cermets are given in
Table 3.4.1. These corrosion tests were part

of over-all evaluation tests to - determine the

suitability of these materials as seat components
of valves for sodium or NaK service at 1000°F.
The specimens were tested at 1200°F rather than
at 1000°F in an oftempt to make the corrosion,

if any, severe enough so that the differences

in corrosion resistance would be more apparent;
however, metallographic examinations of the
tested specimens indicated that they were not

attacked. :

It is interesting to note that K162B was attacked
by sodium to a depth of 0.0005 in. in a similar
test in which the temperature' was 1500°F.

 

w. H. Cook, ANP Quar. Prog. Rep. March 31, 1957,
ORNL-2274, p 171.

UNCLASSIFIED
< ¥e23269 0

   

 

    
   

il

fl(C‘_ —
15, T Y
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wa

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KR
43

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3 P A o 5 a2, - i A 1
§ =By C Ty Y AN A £
s o HE RS % . -
_ - o R Ve . ‘. i
= Loy % i g ;4 o ' e . - 4
; St Y . e T A PN K i T o e ‘. E
F St LE o Sk e . i o AN Lo B oY a7 \
v F - v £y T o . i ’ e s e |
s A P o .
{ ; ‘_/ s : 3 z =‘\ 4 c ‘ 9 , A

h

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4,

Fig. 3.4,6. Mallory 1000 (90% W-6% Ni-4% Cu) (c) As-Received and (b) After Exposfire for 100 hr to Sodium at
1500°F in an Inconel Container in Seesaw Furnace Apparatus. The globular particles are tungsten; the black areas

are voids. 100X. Unetched.

218

 
 

 

 

 

@

&

PERIOD ENDING SEPTEMBER 30, 1957

Table 3.4.1. Designations and Compositions of Cermets Tested in Sodium at 1200°F in Seesaw Furnace Apparatus

 

~ Chemical Composition {(wt %)

 

Designation* Primary Component

 

Co Ni Mo W Ti Ta Nb C
K94 - WC 11.75 80.6 = 1.4 0.6 5.4
KM | WTIC, 10.75 685 67 54 17 67
K162B TiC 25 5 521 03 45 13

 

*These cermets cre manurfo-c:'tur.ed by Kennametal,

stated by the manufacturer,

EXPERIMENTAL STUDIES WITH MOLTEN
LITHIUM

E. E. Hoffman  T. Hikido?

Reports have been received from NDA cdver.mg- |
of

corrosion by lithium at high temperatures. The. | .
principal reports ;ssued during the year are listed

the fiscal year 1957 subcontract studies?

below:

1. Determination of Nztrogen in thbmm by N .
Sax, N. Chu, R. W Miles, and R. H Miles, -

NDA-38;

2. Purzfzcatzo-n- of thbwm by Vacuum Dtstzllatzou. _

by W. Arbiter and S. Lazerus,. NDA-39; -

3. Effect of Nitrogen on Corrosion by thb:ém by

J. M. McKee, NDA-40

4., A Method for Determmmg the Solutzon Rate of ‘
Container Metals in Lithium by M. Kolodney

and B. Minushkin, NDA-41;

5. Progress on Lithium Experzmental ‘Program

During June, 1957, NDA Memo 75-18.

A brief summary of the recent work is presented
here to brmg up o date the - progress repofled ’

prevnously.

An analyhcal procedure for the determmatlon s
of nitrogen was developed- that is sahsfactory ‘_
for concentrations ‘as low ds 10 ppm. No com-
pletely satisfactory method has yet been found,
however, -for measuring low concentrations of .
oxygen in lithim despite considerable effort.

 

2Orl _assignment from fhe USAF.

3This Nuclear Development Corporcfion of Amerlcu L

{NDA) subcontract is coordinated at ORNL by E.

" Hoffman and T Hikido, o
4T. Hikido and E. E, Hoffman, ANP Quar Prog Rep

June 30, 1957, ORNL-2340, p 235.

Inc., and the designofions and compositions given are those

Solution rate studies of stainless steel speci-
mens (principally type 304 stainless steel) have
‘continued. The base-line solution rates observed
in this study at 1600°F (1.1 to 0.48 mg/in.2.hr)
- were of the same order of magnitude as the rate
of weight loss of specimens suspended in the
legs of thermal-convection loops (0.83
mg/in.2:hr) in which lithium was circulated for
~approximately 100 hr at a hot-leg temperature
of 1600°F, Additions of nitrogen or oxygen
(opproximately 1000 ppm) were found to increase
~ "the solution rate by factors that varied from
1.5 to 3 in 4-hr tests. The test apparatus is
particularly suitable for the screening of cor-
~ rosion-resistant - materials and for comparing
the effects of additives and lithium impurities
on the solution process. The main limitation of
the apparatus is the lack of an inert container
in which to perform the solution rate studies.
It -is hoped, however, that molybdenum will be
a satisfactory inert container.

- Several - type . 316 stainless steel-lithium
~‘thermal-convection loop tests have been com-
. pleted. . These tests were conducted with a hot-

"leg temperature of 1600°F, a cold-leg temperature
“at 1100°F, and a lithium velocity of 7 to 8 fpm.
~ The results of the tests have indicated that,
- -in" general, the rate of loss of hot-leg weight

- increased with .increased nitrogen ‘contamination
. of the lithium.. The lowest rate of weight loss
~ “in the 11 loop tests was approximately 60 mg/in.?2
" in 100 hr, which represents quite heavy and rapid

attack. - It was concluded . that further efforts
to reduce this type of corrosion by still further
elimination of nitrogen from the lithium were
not justified. Only two of eleven loops com-
pleted the scheduled 500-hr test without plugging.

219

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

The nine loops plugged in times that ranged

from 31 to 457 hr. The weights of the crystals
found in the loops varied from opproxlmutely
1to4g :

Mass-transfer crystals were token from five
locations in each of eight different loops for
analysis.  The compositions of the crystals
varied considerably from sample to sample, but
the compositions are fairly well represented by

the over-all average composition shown in Table

3.4.2. As may be seen from the composition,
the crystals were high in chromium, nickel, and
manganese content in comparison with the base
material. Present plcns call for continuation of
this study with primary. emphasns on refractory
metals and corrosion inhibitors.

INERT ATMOSPHERE CHAMBER FOR
~ THERMAL-CONVECTION LOOP TESTING OF
NONOXIDATION-RESISTANT MATERIALS

E. E. Hoffman L. R. Trotter

A stainless steel inert-gtmosphere chamber in
which thermal-convection loop tests may be
conducted was recently constructed. This device
was built so that nonoxidatien-resistant materials,
such as the refractory metals, could be tested
in the unclad condition, since attempts to clad
molybdenum and niocbium with oxidation-resistant
alloys have been generally unsatisfactory, Almost
invariably, the refractory metals have cracked
either during the cladding operation or during the
corrosion test, and therefore the correding medium
(sodium, lithium, or fused salt) has .contacted both
the refractory metal and the cladding metal during
the test. A typical crack is shown in Fig. 3.4.7.
Other methods of protecting niobium from oxidation
are being studied, but none have been perfected
to date. :

- The essential features of the inert-atmosphere

chamber are shown in Fig. 3.4.8. Thermal-
convection loops may be operated in the chamber

. either under o vacuum of less than 5 u or with

an atmosphere of purified helium or argon. A
stainless steel liner is incorporated in the
chamber that serves as a thermal-radiation shield
for the outer walls. The liner would also simplify
cleaning of the chamber walls if a loop leaked.

Several preliminary tests of the chamber have

been conducted in order to determine the effec-
" tiveness of the system in protecting several

UNCLASSIFIED
¥-23348

 

Fig. 3.4.7. Soddle-Weld Section of a Type 310 Stain-
less-Steel-Clad Niobium Thermal-Convection Loop After
Exposure to Lithium for 500 hr. The stainless steel

cladding of the saddle weld was stripped off after com-

pletion of the test. The crack occurred during startup
or operation of the loop.

Table 3.4.2. Averages of Analyses of Mass-Transfer Crystals Found In Eight Lithium=Type 316 Stainless Steel
Thermal-Convection Loops Compared with Anclysis of As-Received Stainless Steel Pipe

 

Average Content (wt %)

 

Cr Ni - Mn , Mo

 

Fe
Mass-transfer crystals* 55.9 22,5 18.0 33 . 12
 As-received pipe 64.9 17.4 12.1 e 22

 

 

(June 14, 1957).

220

:‘”f*These data were compiled from a report by J. M. McKee, Effect of Nitrogen on Corrosion by Lithium, NDA-40
 

 

 

 

"

   
 
    
       
 

DIFFUSION e
PUMP—

T

   

|

W

 

 

 

AP

 

TO ROUGHING 4 Il
PUMP 2 2 I

W

4o

 

 
 
    

&

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR-DWG 22162b

 
 

L 9, COOLANT OUTLET
H.' I

A~ COGLING COILS

 

 

A
R —

IR

N
COOLANT INLET

    
 

5 10 5 20
INCHES

Fig. 3.4.8. Inert-Atmosphere Chomber for Testing Thermal-Convection Loops Constructed of Nonoxidaotion-

.Resistant Mu!*erlqls.

refractory metals from oxidation - ot elevated

temperatures.  In the first test, specimens of.

molybdenum ond "riiq:::bium ~were suspended in @
furnace inside the chamber- (argon atmosphere)

and heated to 1700°F for 23 hr and 2100°F for

1 hr. The specimens showed weight gains of

0.1 and 0.5 mg/in.2, respectively. In a second,
more comprehensive experiment, specimens of four
refractory metals were heated to 1700°F for 24 hr

in evacuated Inconel capsules and in the inert-

atmosphere chamber filled with argon at a pressure .

of 14,7 psi. Hardness measurements of these
specimens are compared with the hardness of
the as-received material in. Table 3.4.3. The

hardness values indicate that in all cases the
specimens tested in the chamber were softer
than as-received material but not substantially

-different from the specimens tested in the

evacuated capsules. The weight-change datq,

which were obtained only for the specimens

tested in argon, indicate that the argon atmosphere
was very pure. e ,
Since these results were encouraging, two
attempts  were made to operate unclad niobium
thermal-convection - loops in the chamber. In
both tests, leaks developed in the saddle-weld
areas of the loops. These welds had been made
in a dry box in an atmosphere of high-purity argon.

221

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 3.4.3. Hardness of Various Refractory Metals As Received and After Being Heated for 24 hr at 1700°F
in an Evacuated Inconel Capsule or in an Inert-Atmosphere Chamber

 

Diamond Pyramid Hardness

Weight Change (mg/in.z)

 

(500-g load)

Niobium

As received 212

Heated in&ocuurfi ' . 152

Heated in argfin 160 0
Molybdenum

As received . 215

Heated in vacuum 207

Heated in argon 210 -0.1
Titanium

As received _ 246

Heated in vacuum 253

Heated in argon 227 +35.4
Zirconium

As received 185

Heated in vacuum 19

Heated in argon 122 +2.5

 

Further tests of niobium will be attempted with
loops fabricated by Fansteel Metallurgical
Corporation. The saddle welds of these loops
are fabricated according to the unique design
shown in Fig. 3.4.9, and it is hoped they will
not crack.

FORCED-CIRCULATION LOOP TESTS OF
SODIUM IN HASTELLOYS, INCONEL,
AND STAINLESS STEEL
J.H.DeVan R. S. Crouse
Two forced-circulation loops, one fabricated of
Inconel (loop 7426-26) and the other of Hastelloy B
(loop 7642-3), were operated with sodium for
2000 hr in order to evaluate the comparative
aemounts of mass transfer that would occur in
these systems over relatively long periods of
time.”The maximum sodium temperature in both
loops was 1500°F, and a 300°F temperoture drop
was maintained between hot- and cold-leg sections.
The loops also contained by-pass cold traps to

222

UNCLASSIFIED
¥-23350

 

Fig. 3.4.9. Niobium Tubing Saddle Weld.

 
 

 

 

¥

—-in - Table 3.4.4.

 

establish and maintain a low level of oxide
contamination in the sodium. ‘

The crystalline deposits found in both loops
after shutdown were weighed, and the amounts
were practically the same, being 21 g for the
Hastelloy B loop and 20 g for the Inconel loop.
The value of 21 g for the Hastelloy B loop may
be compared with a value of 17 g for a similar
Hastelioy B .loop operated for - 1000 hr under
similar conditions.> It is evident that mass
transfer deposits form in a Haostelloy B system
in which sodium is circulated at a substantially
faster rate during an initial 1000-hr operating
period than during a subsequent 1000-hr period.
A similar situation exists in the case of Inconel
systems, although the decrease in the rate of
mass transfer during the second 1000-hr period
is somewhat less for Inconel than for Hastelloy B.
A plot showing the variation in mass transfer
buildup as a function of time for both sysfems
is presented in Fig. 3.4.10. Ll

UNCLASSIFIED
ORNL—LR-DWG 214094

 

 

 

 

 

 

 

 

 

 

 

25
R
— 20 hsTELLO‘L? e .
- /’ A"
3 s r WCON®
Gi —
0
% /
o 10
g |~
w / . . _ .
= _____7_/ LOOPS OPERATED AT {500°F WITH 300°F A7 ___|
5
// COLD TRAPS MAINTAINED AT 30Q°F .
7 '
0

0 500 1000 500 2000
AT OPERATION (hr) o

Fig. 3.4.10. Mass Transfe_r_Byfldup as a Fukngfion_of_-;‘ 7
Time in Foi-ced-Clrcfildtion Loops '_Operut_ed with iSod‘lu'm. _;

The resuhs of chemlcal analyses of the cold-
leg deposits formed in the 2000-hr tests are given -
In. both loops,  the cold-leg
deposits contained greater percentages of nickel -
than were present in the base metals. ~The inner
surfaces -of the ‘tubing comprising the hot legs
of both loops revealed  intergranular attack to
a depth of 2 mils. The deposit in the cold-leg
secfion of the Haste_lloy B loop reached a maximUm

 

5J. H. DeVan, R. S. Crouse, and D. A. Stoneburner,
Alggs Quar. Prog. Rep. June 30, 1957, ORNL-2340,
p

conditions;

PERIOD ENDING SEPTEMBER 30, 1957

Table 3.4.4 Analyses of Cold-Leg Deposits from
Hastelloy B and Incone! Forced-Circulation Loops
“Operated for 2000 hr with Sodium

 

Elements Present in Cold-Leg

Loop Moferi-al Deposits (wt %)

 

 

Ni Cr Fe Mo
~ Hastelloy B 97.4 0.53 0.50
Inconel 85.6 12.8 0.70

 

thickness of 24 mils, whereas the deposit in the
Inconel loop had a maximum thickness of 30 mils.

A Hastelloy W forced-circulation loop, 7642-4,
which was also operated with sodium, was ex-
amined following 1000 hr of operation at a hot-
leg temperature of 1500°F. The mass transfer
deposits found in this loop were approximately
17% greater by weight than the deposits found
in the Hastelloy B loop operated under identical
5 namely, 20 g compared with 17 g.
A metallographic examination revealed a spongy,
heavily attacked area approximately 1.5 mils
deep along the hot-leg surfaces exposed to
sodium. = These corroded areas, which are
illustrated in Fig. 3.4.11, seemed to be quite
brittle, and in many cases had pulled away from
the metal underneath them. The cold-leg deposits
in this loop were found by analysis to contain

97% Ni and 3% Cr.

SR E]

  

|NC|:'IE5

 

 

 

 

" Fig. 3.4.11. Hot-Leg Surface of 'Ha.sféllo'y- W Forced-

Circulation Loop Which Operated 1000 hr with Sodium at
1500°F. Etchant: H2Cr04-H Cl. 250X, Reduced 31%.

223

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

A type 347 stainless steel forced-circulation
loop also completed 1000 hr of operation with
sodium at @ maximum temperature of 1500°F and
a temperature drop of 300°F. The loop (7426-27)
also .contained an oxide cold trap, which was
maintained at 300°F. Visual examination of this
loop revealed only slight traces of mass transfer
in the cooled portions of the loop. The deposits
were similar in extent ond appearance to those
which have been observed in types 316 and 304
stainless steel loops operated under similar
canditions, and were much less in amount than
the deposits found in Inconel loops operated
under similar conditions. A typical cold-leg

section from the type 347 stainless steel loop

is shown in Fig. 3.4.12. The hot leg of the loop
showed light surface pits and wvoid formation
to a depth of 2 mils.

 

 

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Lrwoc v Wi . v gt
il e @, o My P e E e R

Fig. 3.4.12. Mass Transfer Deposits in Cold-Leg
Section of Type 347 Stainless Steel Forced-Circulation
Loop Operated with Sodium for 1000 hr ot ¢ Maximum
Temperature of 1500°F, Etchant: glycerol regia. 750X.
Reduced 32%.

DIFFUSION OF NICKEL IN LIQUID LEAD
J. L. Scott  H. N. Leavenworth, Jr.%

Studies of the effects of temperature and con-
centration on the diffusion of nickel in liquid
lead were continued as a part of an over-all
fundamental investigation of the mass-transfer
process. The initial work and the experimental
method were described previously, ond the dif-

 

60n assignment from Pratt & Whitney Aircraft.

224

fusion coefficients for saturated solutions of

various compositions were presented.” Additional
studies have now been conducted in order to
determine the effect of temperature on the diffusion
coefficient at constant composition. A constant
composition for a series of tests. was obtained
by saturating the lead, which filled the capillary,
with nickel at a given temperature, and then
making the diffusion measurements at a higher
temperature.  Saturation temperatures of 363°C
and 481°C were used for these tests. The results
are shown in Fig. 3.4.13, which is a plot of the

logarithm of the diffusion coefficient, D, vs the

reciprocal of the absolute temperature. The
curve for the saturation temperature of 481°C
is slightly above that for the saturation tem-
perature of 363°C, but the two curves are roughly
parallel. In order to determine whether the spread
in these curves was a result of random variation,

a statistical anclysis of the data was made. These

calculations showed that the curve for 481°C
differed from the curve for saturated solutions
at the 50% confidence level, but not at the 95%
level. The data for the curve for 363°C were
found to be significantly different from the satu-
ration data at the 95% confidence level. Thus,
it appears that the diffusion coefficient for nickel
in lead is a function of composition, even though
the saturation composition is less than 0.5 wt %

 

7J. L. Scott and H. W. Leavenworth, Jr., ANP Quar.
Prog. Rep. June 30, 1957, ORNL-2340, p 236.

5 UNCLASSIFIED
x10 ORNL—LR—DWG 25947
200

100

50
0O SATURATED AT TEMPERATURE OF RUN

A SATURATED AT 481 °C

20 C SATURATED AT 363 °C

== DIFFUSION OF NICKEL IN SATURATED LiQUID
10 =~==DIFFUSICN OF NICKEL iN UNSATURATED LIQUID

SELF-DIFFUSION OF LEAD

DIFFUSION COEFFICIENT, Otem™/sec)

 

M L5 12 125 13 435 14 145 15 455 46
+ x 10° oK)

Fig. 3.4.13, Diffusion in Liquid Lead,

 
 

 

 

 

 

 

 

 
   

at the highest temperature studied. It also appears
that a change in composition affects the entropy
of activation for diffusion but not the enthalpy
of activation. These results are only tentative
because the range of temperatures studied and
the amount of data obtained at each temperature
Jnsufficient for a complete analysis. Higher
Tatures will be- attained through the use of
quartz diffus cells, “which.. will _replace the
less expensive pyrex vells used in these tests,
At the same time, a secohd\method of chemical
analysis for nickel in lead, namely,  activation
analysis, will be utilized in addition fo the
polarographic method, so that results from in-
dependent anolyhcol methods €an be compored

PURIFICATION OF somum '
J. L. Scott  H. N. Leovenworth Jr.

The vacuum shll which was desngned for
purifying small batches of sodlum for use in

fundamental studies,  has now - ‘been. fobncated

and installed. A picture of the - stlll ‘before the
insulation was installed is shown in Flg. 3.4.14,
The basic components of the ‘still are ‘the “sodium
pot, the receiver for the . dlshlled mefol ‘and @
storage tank for undlstllled meml “The pot and
receiver are each 8 in. in dlometer ond about

12 in. high. Sodium is. introduced into the system

and withdrawn through the tubes below the
receiver. It is pumped from contomer to contomer

with the use of puflfied orgon un&er pressure.

An odvanfoge of this sysfem is thof sodlum can
be transferred from the receiver to the pot “and
redistilled any number of times without opening
the system. The heat sources for distillation

and  for momtommg the sodium in the molten
state are Calrod units, as shown, Heatmg tape
will be utilized -to maintain the tronsfer tubes
above the meltmg poinf of sodmm. Cooling for
_condensation  of - sodium vapors  is- provided by
“coils through wbich efhyiene glycol is pumped. -
These “coils “are |n3|de the tube between -the
pot und the “receiver. - “The level of the sodium
- in the recelver is determmed by spark-plug probes' L
~that pass through ‘the receiver flange. S
- Safety is provuded through the use of the: regu-
- lating system which .opens the volve shown in
~ Fig. 3.4.14. to introduce purlfled argon when the
pressure ‘in the system rises to above 1 p.- This

device, ‘which operates off a thermocouple gage,

also turns off the power to the still and the

vacuum components. The vacuum is provided by a
Megavac-25 pump and a Consolidated Vacuum MC

'500 dlffuslon pump.
.seporoted from the stiil by the valve and a cold

from supersoturoted ‘solutions.
ocud is shghtly soluble in water and has a high
temperature coeff:cnent of solubility. As the

PERIOD ENDING SEPTEMBER 30, 1957
These components are

trap, which is not shown. The pressure in the

system is measured with a National Research

type of ion gage. Preliminary tests of the system
have shown that a vacuum of 5 x 10~5 4 is readily

‘.oftomoble Operation of this unit will be started
.as _soon as the electrical wiring and instrumen-

tation are completed.

“MASS TRANSFER IN AQUEOUS SOLUTIONS
J L. Scott P. Y. Jackson®

Thermol-convechon loop tests in which water

is “circulated through a cylinder of solid benzoic
_ocud locufed in the hot zone have been initiated

in the hope that this “system will simulate a
hqmd-metol system for the analysis of nucleation
Solid benzoic

benzoic acid dlssolves, it s transferred by
essentially free convection to the cold zone,
which is_ maintained ot a temperature approxi-
mately 20°C lower than the temperature of the
hot zone. The process may be monitored by
taking somples from the loop for titration and
visually studying the nuclecmon ‘and growth of
crystals in the cold zone. * Although insufficient

- 'data for complete analysis have been obtained,
some ‘trends have been ‘observed. Titration of

samples of solution show ‘that the solid dissolves
at a rate which is nearly constaont, since saturation
is not approached in the hot zone. A considerable
degree of supersaturation is attained in the cold

. zone before crystallization ‘becomes visible. The
 condition of the cold-zone sutface, whether clean

or dirty, smooth or etched, or straight or formed

_in a_series of bulbs, affects the rate of formation
~of crystal nuclei ond the rate of attachment of

these nuclei to the wall.  When the cold zone

is provided with a collar of,crysfols upon which
_new melecules can deposit, the concentration of
 the acid in the water levels off, as shown in
.run 3 of Fig. 3.4,15, with little tendency toward
supersaturation. In the absence of crystals, the

concentration rises to a value correspondmg to

‘some degree of supersaturation in the cold zone,
~ but after recrystallization has begun, the con-
"centration falls rapidly to a value corresponding

 

8 Summer participant, St. Peter’s College, Jersey City,

225

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

RECEIVER FOR[

STILLED SODIUM}

 

PRESSURE AND POWER™™

 

REGULATING SYSTEM |

226

Fig. 3.4.14. Sodium Still.

UNCL ASSI

Y2338

FIED
3

 

a

 
 

 

to the saturation level, run 8 of Fig. 3.4.15. The
hot-zone saturation curve was calculated, from
the following equation, to show the solution rate
at the hot-zone temperature of 60°C for the given
surface area A and loop volume V:

(-

where

C = C (1 - ~AT/V)

C = concentration at time T,

C, = saturation concentration,

a = the solution rate constant, which is
evaluated from the solution rate at times
close to zero. '

Further tests will be made to determine the effects
of changes in the hot-zone temperature, cold-zone
geometry, and the surface conditions on the mass-
transfer process.

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR-DWG 25948

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

T
g9 SATURATION
= P CONCENTRATION
« 8 12 g/liter AT 60°C
p
g7 o PLUGGED]
z J [ —FIRST CRYSTALS VISIBLE
Q6
g 7 '%fi: PLUGGED | |
o5 £ e =] “SATURATION CONCENTRATION _|
2 / = CORRESPONDING TO MIXED MEAN
S ) COLD ZONE TEMPERATURE OF 40°C
w4 T 1
@ / \SATURATION CONGENTRATION CORRESPONDING TO
5 3 / CONDENSER WALL TEMPERATURE OF ~30°C____|
g, |
g ® COMPUTED SATURATION CURVE FOR HOT ZONE |
£ ]_© NO CRYSTALS IN COLD ZONE AT BEGINNING OF TEST-RUN 8 _|
& A EXCESS OF CRYSTALS IN COLD ZONE AT BEGINNING OF TEST-
g o RUN 3 R D

0 50 100 150 200 250

TIME (min)

Fig. 3.4.15. Concentration vs Time Curve for a
Benzoic Acid-Water System in a Thermal-Convection
Loop.

227

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

< . : T

YTTRIUM METAL PRODUCTION

T. Hikido! R. E. McDonald?
Jo Ho CQObS

Experimental studies of the production of high-
purity ytirium metal were continued. In the process
being developed, lithium is used to reduce a mix-
ture of YF ;-MgF ,-LiF to yield an yttrium-magnesium
alloy. The fluoride mixtures used are prepared by
the Chemistry Division with the use of techniques
developed for fluoride-fuel processing (see Chap.
2.4, **Production of Purified Fluoride Mixtures'').
The results of the experiments completed thus far
are summarized in Table 3.5.1.

The low yields obtained in runs L-2 and L-3 have
been attributed to segregation in the crucible; that
is, the lithium is separated from the YF ,-MgF ,-L.iF
layer by the LiF slag, which has a density inter-
mediate between the densities of the two reactants,
In run L4, the reaction retort was tilted at a 45-deg
angle to increase the areas of the interface layers
and also to promote thermal-convection mixing.
The yield was increased from 67 to 83% by this
procedure.

 

10n assignment from USAF.
20n assignment from Pratt & Whitney Aircraft.

3.5. MATERIALS FABRICATION RESEARCH

The lithium was added in the first three runs in
the form of sticks cut into short lengths. In run
L4, the lithium was transferred into the reaction
retort in the molten state by differential argon-gas
pressure. The molten lithium was filtered through a
porous stainless steel filter in this transfer process.

Experiments have been conducted on vacuum
heat treatment of the ytiriumemagnesium alloy to
distill off the magnesium and excess lithium. A
20-g sample of the yttrium sponge produced in this
manner from run L-1 was arc melted with the use
of a tungsten electrode under an argon atmosphere.
The oxygen content of the arc-melted button was

determined to be 1200 to 1300 ppm by a. vacuum- -

fusion analysis performed by the Analytical
Chemistry Division, A 30-g button from run L-2
has been arc melted and is now being analyzed.

HYDROGEN DIFFUSION THROUGH METALS

R. E. McDonald T. Hikido
J. H. Coobs

Equipment with which to measure the rate of
diffusion of hydrogen through potential cladding
materials for hydride moderators is being built,
The apparatus is shown schematically in Fig. 3.5.1.
The test specimen for the basic diffusion measure-
ments will be a heavy walled tube of the material

Table 3.5.1. Summary of Results cf Ytirium Reduction Experiments

 

 

 

Quantity Yield of Y-Mg Alloy
Run of Lithium (9) Yield
Ne. Charge Reductant Added (% of theoretical)
Theoretical Actual
(9)
L-1 591 g of YF3, 151 . 450 375 83
(10% excess)
232 g of MgF,,
174 g of LiF
L-2 1000 g of YF3-M9F2- 137 411 246 60
LiF mixture (53.9- (10% excess)
21.1-25.0 wt %)
L-3 200 g of mixture used 290 822 551 67
in run L~2 (15% excess)
L-4 1000 g of mixture used 150 411 340 : 83
in run L=2 (15% excess)

 

228

4

 
i

 

 

 

 

o

FURNACE MUFFLE
WATER QUT

WATER JACKET

HOT ZONE

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR-DWG 25962

DIFFUSION WINDOW { THINNED TO 0.010in.}
QUTER GAS TUBE

TEST SPECIMEN

 

2-THERMOCOUPLE WELLS

 

~ INSULATOR

 

 

 

/WATER JACKET

 

 

END PLUG SF

= ~—=-— WATER IN

 

 

 

 

 

 

 

 

PALLADIUM VALVE
.~ WITH HEATER

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

 

 

FURNACE CONTROLLER

\, -
~ailf—
"O'LRING

- SEALS

   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

THERMAL-CONDUCTIVITY CELL ARGON FLOWMETER
SCALE ~—mg OF H, FURNACE PROGRAMMER CONSTANT-PRESSURE
§ - ‘ HYDROGEN REGULATOR
\ e
ARGON + HYDROGEN GAS

 

" BROWN RECORDER

Fig. 3.5.1. Diagram of Apparatus for Measuring the Rate of Hydrogen Diffusion Through Metals,

being studied, with a section reduced in thickness.
This thin **diffusion-window’’ section will be held
in the hot zone of the tube furnace. Hydrogen gas
will be maintained at a constant pressure inside the
tube, while the outside of the tube will be purged
continuously with high-purity argon gas. The
purge ‘gas will be passed through a' thermal con-

duetivity ‘cell which will detect changes in con-
ductivity of the gas as a result of small additions =
The factor of 10 dif-
ference between the thermal conductivity of hydrogen
(0.416 x 10~3 cal/cm+sec:°C at 0°C) and that of

of hydrogen to the argon.

argon (0.039 x 10~3 at 0°C) wili make possible the
detection of quite small quantities of hydrogen.

The impulse from the cell will be transmitted to a

strip-type Brown recorder with a scale marked to

present the hydrogen content.

“The same eqmpment WI" be used to measure fhe_ o
rate of loss of hydrogen from clad hydrlded-metalJ
assemblies. It will also be useful for monitoring .

hydrogen losses during thermal cycling and hydrogen
migration tests,

VOLATILITY OF BeO IN STEAM
AT MODERATOR TEMPERATURES

A. G, Tharp

The major reported work on the volatility of BeO
in steom is found in publications by Grossweiner
and Seifert® and by Elliott.# The paper by Elliott,
which was a doctorate thesis, reviews all the work

-on volatile oxides in water vapor.

" As is usual in these studies it is very difficult to

“prove ‘an identification of the volatile molecule.

The best summation of the present data indicates
that the major reaction is

-~ xBeO(s) + HZO(g)——> xBeO *H,0(g)

There is evidence that one or more volah le species

~exist which are dependent on oxygen pressure. If

 

3L. Grossweiner and R. L. Selferf, ] Am. Cbem. Soc.
74, 2701 (1952) and The Reaction of Beryllium Oxide

- with Water Vapor, AECU-1573 (July 30, 1951).

4. R. B. Elliott, Gaseous Hydrated Oxides, Hy-
droxides, and Other Hydrated Molecules, UCRL-1831
(June 1952).

229

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

the reaction given above is assumed to be essen-
tially the correct one, the data summarized in
Table 3.5.2 are applicable.

The dota of Table 3.5.2 are for equnhbrlum con-
ditions or as near to equilibrium conditions as
can be attained in an experiment of this nature.
The samples of BeO used were in powder form, and
Grossweinerand Seifert used water vapor in helium,
Elliott used H,0.0, and H,0-H,. It should be
pointed out that inactual practice it is not possible
to obtain a greater volatility than that indicated by
the equilibrium constant; that is, no concentration
or pressure can exceed the equilibrium. . The
dynamic conditions of the BeO(s) + H,O(g) re-
action would determine the amount of BeO carried
by the steam. Another important factor would be
the nature of the BeQ, that is, density, crystal
size, external surface area, etc. Another important
factor could be that another molecular species of
hydrated BeO would become the controlling species
under different experimental conditions.

The available data, which are probably quite
accurate, indicate that under some conditions it
would be entirely possible to corrode and transport
large quontities of BeO in steam in relatively
short times.

FABRICATICON OF HIGH-DENSITY BeO
R. L. Hamner

It has been found in the process of fabricating
BeO by hot pressing that the “Luckey S.P.”’-

grade oxide produced by Brush Beryllium Company
is equivalent to the ‘‘flucrescent’’ grade when
densities of 95% or greater are desired. Such
densities -are obtained by pressing at 1900 to
1950°C with a pressure of 2000 .to 2500 psi. A
quantity of high-purity BeSO, has been received
for experimental attempts to lower the temperature
of hot pressing and to obtain high-density material
by cold pressing and/or extrusion followed by
sintering ot about 1500°C. This experimental
approach is based on the production at Battelie
Memorial Institute® of a highly sinterable material
by carefully controlling the calcination temper-
ature to obtain an extremely fine oxide powder.

The hot-pressed BeQ specimens described below

~ were fabricated for physical tests:
1. five BeO cylinders, 2 x 2 x 1/2 in. ID; density,

97%; two cylinders damaged in machining;
2. four BeO blocks, 2 x 4 x 1 in., with four J '4-in.-
dia holes running Iongltudlnally,
one BeO block, 1 x 1 x 6% in.; density, 80%,
one BeO block, 1 x ‘?’ x 6% in.; density, 90%;
one BeO block, 1 x ‘/ X 67, m., density, 96%,
one BeQ cylinder, 2/2 in. OD, / in. 1D, 2/2 in,
long; density, 80%;
7. one BeO-UO, (50-50 mole %) block, /16 X /16
9in.; densny, 98%.

oA w

 

S.S‘ummary of Progress Report on an Invéstzgatzon of
the Manufacture of High Density Beryllium Oxide.
f;g;ls) by the Brush Beryllium Co., PWAC-167 {April 15,

Table 3.5.2. Data on the Reaction of BeO with Steam

 

Ratio of BeO-H20 {9)

 

Flow Rate Pressure to
Temperature H,0 BeO (liters/min) Steam Hydrogen  Oxygen Pressure of H,0 (g)
(°C) Passed Collected at Room Pressure Pressure Pressure —— =
(moles) (moles) Temperature (atm) (atm) (atm) Grossweiner
P Y . EHiott and
Seifert
x 10—4 x 1073 x 103
1270 1.95 0.3 1.3 0.87 0.13 1.0 6.9
1285 1.83 0.84 1.7 0.74 0.26 8.4 7.6
1310 3.44 1.52 1.5 0.95 0.05 4.4 9.8
1310 2.27 1.12 19 0.91 0.09 4.9 9.8
1310 3.9 3.40 1.7 0.74 0.26 8.7 9.8
1254 3.88 1.56 1.4 0.88 0.12 4.0 5.9

 

230
 

 

 

P

The physical properties to be investigated include
modulus of rupture, elastic modulus, internal
friction, thermal expansion, thermal conductivity,
total emissivity, and erosion at temperatures up to
1650°C. In addition, UO o-BeO diffusion studies

are to be made.

BeO THERMOCOUPLE INSERTS AND
IRRADIATION TEST SPECIMENS

R. L. Hamner

Special BeO thermocouple inserts and shells for
determining volume temperature fluctuations in the
ART haoif-scale core model are being fabricated.
Hot-pressed BeO blocks are to be ground to cyl-
inders /8 }/2 in. for the 12 inserts and to cylinders
'4 X l/2 in. for the accompcmymg shells, Holes
will then be made in the ]/4 X / -in, cylinders by
the activation technique to form the shells to
accommodate the inserts, and parallel grooves
180-deg apart will be made in the same manner along
the periphery of the inserts for holding the thermo-
couple wires. For good thermal contact the base
of the shell and the mating surface of the insert are
to be platinized.

Twenty-seven l-in.-long high-density-BeQ cyl-
indrical specimens with diameters of from 0.44 to
0.94 in. are being fabricated for irradiation studies.
These specimens are to be canned in Inconel and

irradiated in the ETR.

THERMAL STRESS RESISTANCE
OF CERAMIC CYLINDERS

R. A. Potter

An apparatus is being designed for the deter-
mination of the heat throughput required to cause

initial failure in ceramic cylinders.  This in-.

formation will aid in determining the thermal stress

resistance of various ceramic materials. * It will -

also provide a means of sfudymg the effects of

varying the physical properties of the muterlals on

their. ability. to withstand thermol shocks. Con-

sideration is being  given to various methods of
heating hollow cylmders mternolly and coolmg.

them exiernaliy.

FABRICATlON OF METAL HYDRIDES
' R A. Poh‘er | '

Nlne groups of dense zirconium hydrlde samples

have been prepared in the hydriding apparatus.®

 

SR. A. Potter, R. E. McDonald, and T. Hikido, ANP
Quar. Prog. Rep. March 31, 1957, ORNL-2274, p 193.

PERIOD ENDING SEPTEMBER 30, 1957

The samples are rods approximately 3 in. in length
and '/2. in. in diameter, and they contain various
amounts of hydrogen, as shown in Table 3.5.3.
These pieces were made as part of an effort to
establish procedures for hydriding massive yttrium
metal,

Modifications have been made in the equipment,
which is now ready for production operation as
yttrium metal becomes available for hydriding. The
procedure to be used is based on a process,
developed at the General Electric Company, in
which a mixture of hydrogen and helium is passed
over the metal.

Toble 3,5.3. Hydrogen Contents of Derise
Zirconium Hydride Samples

 

 

Sample 5 | Hydrogen
Group ?mP e_ Content*
No. Designation (% of total weight)
| a 0.19
b 0.21
il a 0.36
b 0.41
it a 0.54
b 0.56
v a 0.83
b 0.83
Y a 0.86
b 0.92
vi a 0.95
b 0.95
Vi a 05
Vi a 1.10
AX - o 1.15
- b 1.15

 

*Determined on the basis of weight gain.

231

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

3.6, METALLOGRAPHIC EXAMINATIONS OF ENGINEERING TEST COMPONENTS AFTER SERVICE

ART PROTOTYPE TEST RADIATOR
R. J. Gray J. E. Yan Cleve, Jr.

A metallographic examination was made of the
York Corp. ART prototype test radiator No. |,
which operated a total of 870 hr under the following
conditions:

|sothermal operation at 1200°F 390 hr

Operation with varicus temperature dif- 47 hr
ferentials and a meximum NaK inlet
temperature of 1200°F

Operction with varicus temperature dif- 360 hr
ferentials and a maximum NaK inlet
temperature of 1500°F

Operation at the design condition of a 73 hr
NeaK inlet temperature of 1500°F and
an outlet temperature of 1070°F

Thermal cycles, total 9]/2
Slow cycles at a maximum NaK tem- 7
percture of 1200°F
Slow cycles ot a maximum NaK tem- 2
perature of 1500°F
Fast cycle (50°F/min) to a NaK tem- I/i_,

perature of 1525°F

The radiator failed upon reaching full power in the
first controlled thermal cycle. The NaK inlet
temperature had been increased from 1350°F to

1525°F at a rate of 50°F/min and the outlet tem-

perature had remained constant at about 1070°F.

After the radiator had been cleaned it was
sectioned for examination, The section containing
the tube that failed, which was located by pres-
surizing the tube with wafer and observing the
leak, is shown in Fig. 3.6.1. The arrow paints
toward the tube that leaked, A close view of the
tube that failed is shown in Fig. 3.6.2. The
numbered arrows indicate the readily visible
holes, which appear as spots because of internal
lighting during photography. The metaliographic
examination did not reveal evidence of complete
penetration of the tube wall at either the circum-
ferential or the longitudinal fissures.

The some tube is shown in Fig. 3.6.3 rotated
approximately 90 deg. The dark area in the tube
is hole No. 1 in Fig. 3.6,2., The curvature of the
tube as it enters the header is evident. The good
condition of the tube may be attributed to the
promptness with which the fire that occurred as a
result of the leak was extinguished,

Hole No. 1, which was the largest of the three
holes, is shown in Fig. 3.6.4. The opening in the
tube wall in this plane measures 0.030 in. More
of the tube wall that appears at the extreme right
edge of Fig. 3.6.4 may be seen in Fig. 3.6.5.

Hole No. 2 is shown at a magnification of 100
in Fig. 3.6.6 and at a magnification of 500 in

UNCLASSIFIED
Y-13342

 

Figs 3.6.1. ART Prototype Test Radiator No. 1. (SeTfet with-cuptign)

232

&

 
 

 

 

- PERIOD ENDING SEPTEMBER 30, 1957

 

Fige 3.6¢3. Tube Shown in Fig. 3.6.2 After Being fiofuted 90 deg.

UNCLASSIFIED
O Ye23343

UNCLASSIFIED
3341

233

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

Figs 3:6¢4. Hole No. 1 of the Tube That Feileds The opening in the tube wall is 0.030 In. Etchant: 10%

oxalic acids 100X.

 

INCHES

0.02

 

 

 

0.02

 

 

 

 

Figs 3.6:5. Tube Wall Shown at the Right Edge of Fig: 3.6:4. Note that tfie internal pressure in the tube produced

an outer llp at the point of fallura_- 100X.

234

W
 

 

 

Fige 346¢6. Hole Nos 2 of Tube That Failed.
Etchant: 10% oxalic acids 100X.

Fig. 3.6.7. As may be seen the shape of the metal
bordering the hole is simitar to that of the metal
bordering hole No. 1. Figure 3.6.7 shows that
there was .no grain-boundary void formation, and
there is no evidence of incipient failure, even
very near the area that failed,

The microstructures of the three tubes surroundmg
the tube that foiled are shown in Figs. 3.6.8,
3.6.9, and 3.6.10. Again, there is no evidence of
incipient failure in any respect.

The inside diameter, outside diameter, and wall o

thickness of the fube that failed were measured at

various times ‘during polishing ot the specimens, -
Variations in the ocutside diameter and in the wall

thickness were noted in several instances, but all

done by the fire. There was little or no variation -
in - the inside diameter whlch would indicate the -

sresence of a stress high enough to produce plastic
flow in the tube,
YORK CQ,RP.,RADIATOR NO, 16
R, J. Gray J. E. Yan Cleve, Jr.

York Corp. radiator No. 16 failed with a NaK-to-air
leak after 1438 hr of operation, including 1044 hr

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
Y-23354

 

 

Note that the shape of the metal is the same as in Figs 3.645,

under thermal gradient conditions. The NaK inlet
temperature during thermal-gradient operation was
1500°F or above for 541 hr, and the unit was
thermally cycled 21 times.

The fire damage was extensive, and therefore no
attempt was made to evaluate the integrity of the
brazed fin-to-tube joints. Of the 72 tubes attached

to the NaK inlet header, 29 were burned through
" and 5 more were found to be plugged. Four tubes

contained partial plugs, and one was completely
plugged. The plug was drilled out of the tube that
was completely plugged, and the plug material was

. found to be green in color, The partial plugs were

removed by scraping the tube walls and were found

~to -be brown to black in color,  Chemical analysis

the variations could be dlrectly related to damage . showed the green muterlcl to contain 3.51 wt %

‘wranivm and 4.19 wt % zirconium, -

The darker
moterial - contained 7 wt % U, 47 wt % Ni, 7 wt %
Cr,and -3 wt % Fe. The difference between the
iwo plugs therefore was that the dark-colored
naterial contained considerable mass-transfer
leposits. The presence of uranium indicates that
there was a leak of fuel into the NaK in the fuel-
to-NaK heat exchanger included in the circuit in
which this NaK-to-air radiator was operated. The

235

 
 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
. Y-23353

 

 

 

 

 

 

 

 

Figs 3.6.7. Higher Magnification of Fig. 3.6.6. No evidence of incipient follbrg may be seen. Etchant: 10%
oxalic acide 500X, '

-

'INCHES
1 1

o
Q
-

 

 

 

 

 

 

 

 

 

T s
P

 

 

 

Gy, @Oy E
- - . w

 

Fige 3.6.8. Microstructure of Tube Adjacent to the Tube That Failed. No evidence of incipient failure was
found. Etchant: 10% oxalic acids 200X, :

236

 

 

6
 

 

 

 

 

£
i
i
1
i
i
1
1

 

 

 

TR .  PERIOD ENDING SEPTEMBER 30, 1957

CLASSIFIEOH [~
¥.23339 - |

 

 

 

 

 

 

 

 

 

 

 

 

 

. Fige 3.6.9. | Microstruclure of Tube Adlccentto the ':Tl;hfl V‘I_'hcf ‘Flc_:_lled. | 'No.evidcrice of inciplent failure was

founds Etchant: _-l_(_).%_fq“'xo|ic acide 200X, . - 3 D

 

 

 

 

 

 

 

 

 

 

 

 

Figs 3.76.10‘. Mierostructure of Tube Adiccenf to the Tube That Falleds No evidence of incipient failure was
founds Etchant: 10% oxalic acids 200X. '

237

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

examination of the heot exchanger is: descrlbed in

the following: sectlon.

PROCESS"ENGINEERING CORP. HEAT
EXCHANGER NO. 3, TYPE SHE-7

R. J. Gray J. E. Van Cleve, Jr.

A small heat exchanger of type SHE-7, which
was fabricated by the Process Engineering Corp.
and is designated No. 3, was ferminated after
1438 hr of operation with the fuel mixture NaF-
ZrF UF ;. (56-39-5 mole %, fuel 70). The unit
operated under thermal-gradient conditions for
1044 hr and was thermally cycled 21 times. Opera-
tion of this unit was terminated because of failure
of the radiator in the system (see preceding section

of fhi'_s chapter).

Thirty-one samples were removed from the tube

‘bundle and header areas and mounted for metallo-

graphic examination to determine the depth of
corrosion on both the fuel and the NoK sides.
The fuel-side corrosion was found to range from a
minimum of 0.001 in. to a maximum of complete
penetration. The NaK-side corrosion was found to
vary from general roughening of the surface to
penetration to a maximum depth of 0.004 in.

An area from the hot-header section is shown in
Fig. 3.6.11, As may be seen, the voids completely
penetrate the tube wall, but there is little or no
general corrosion. Adjacent areas in which the
same conditions exist are shown in Figs. 3.6.12
and 3.6.13. A metallic layer found on the fuel side
of a tube from the hot-header area is shown in
Fig. 3.6.14, Spectrographic analysis showed the
metallic layer to be 6.3 wt % Cr, 8,5 wt % Fe, and
76 wt % Ni. This metallic layer was not similar to
the gold-colored deposit found in the cool section
of a previously examined heat exchanger. The
gold-colcred deposit was found to have resulted
from corrosion of the header region.

The examination. of the heat exchanger did not
show any open cracks through the tube walls. The
presence of the uranium- and zirconium-bearing
plugs in York radiator No. 16 proved, however,
that a leck was present, Fuel probably leaked
through several of the tubes which showed com-
plete grain-boundary penetration, such as those

shown in Figs. 3.6.11, 3.6.12, and 3.6.13,

238

FEE]

BN 131N CLASSIFIED
RN Y-23431

T
1

t
CHES
1 -

r:

&

S
s

    
      
 

o
ol

EEEEE

B

e
>

2
&

BEEE

00X
1

i f [

 

 

P’ ' : .
P B . - T by g
.022 § Cow s At °
) .oa . i . G & .
2025 | [ T A . )

Fig. 3.6.11. Area from o Tube Near the Hot-Header
Section of @ NaK-to-Fuel Heat Exchanger Operated at
High Temperatures. Note complete grain-boundary
penetration and the wide fissures in both surfaces, with
little or no general corrosion. Etchant: 10% oxalic
acid. 200X, Reduced 24.5%. (Lomfidemttot—with

vhrory) ,

PROCESS ENGINEERING CORP. HEAT
EXCHANGER NO, 2, TYPE SHE-7

R. J. Gray J. E. Van Cleve, Jr,

Process Engineering Corp. heat exchanger No. 2,
type SHE-7, which operated for a total time of 1845
hr and for 1645 hr with NaF-ZrF ,-UF , (50-46-4
mole %, fuel 30) was also exammed Thls small
heat exchanger was operated under thermal-

. "gradient conditions for 1456 hr and was thermally
cycled 29 times. The fuel-inlet temperature was

above 1600°F for 552 hr. When operation was
terminated because of o failure of the economizer
in the circulating cold trap that was operating in

"

o

 
 

 

 

-

 

  
      

EEE]

Y
1

INCHES

£

 

 

Fig. 3.6.12. Tube Adjacent to the Tube Shown in
Fig. 3.6.11. Complete grain-boundary penetration, with

little general corrosion, may be seen.
oxalic acid. 200X. Reduced 24.5%.

Etchant: 10%

the NaK system, the fuel circuit would not dumpr"‘

and the removal of samples was dlfflculf :
Thirty-four samples were removed from vthe tube

bundle and header areas.and mounted for metallo-
graphic examination. - The fuel-side corrosion was

found to range from a minimum of 0,001 in, to a
maximum of complete: penefrahon. “The NaK-side

corrosion varied- from ‘general ‘roughening of the
‘exposed - surface to a maximum depth of 0,010 in.

The general corrosion and gram-boundclry voids

found " in “the ‘wall’ exposed to fuel ore shown in

Fig. 3.6.15, - The general corrosion in Fig. 3.6.15
reaches a- depth of 0.005 in., -and beneath this

 

‘extends 0. 005 in, of gram-boundary voids, fThus“:”

the depth - of corrosion on “the fuel . srde was -

0,010 in., or approximately two- fifths of the tube
wall.

 

PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

 

 

 

 

Tube Adjacent to Tube Shown In Fig.
Complete grain-boundary penetration, with
"Etchant: 10%

- Fig. 3,613,
3.6.12,
“lttle general corrosion, may be seen.

oxalic acid, 200X,

The NaK-side of the tube -'wull'_shdwn in Fig.

- 3.6.‘]_5—is shown in Fig. 3.6,16, The corrosion,

“which ¢an, in this i'insicnée,"be' 'atfiibu'ted to the

'NaK, extends to a depth of 0,004 in. This depth

of attack added to ‘that by the fuel yieids 0,014 in.
of corrosion attack; that is, slightly more than
one-half the tube wall was attacked.

239

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

LIL LB

o
n
(=]

Bl 1]

1
ISCP(

 

 

m‘u’pfim‘,o

  
  

B

} UNCLASSIFIED
. Y-22558

 

 

 

 

 

 

 

Fig. 3.6.15. Fuel Side of Tube from Hot-Hecder Reglon of Process Engineering Corp. NaKeto-Fuel Heat Ex-
changer No. 2, SHE=7. General corrosion may be seen to a depth of 0.005 in. plus 0.005 in. of grain-boundary voids.
Estchant: 10% oxalic acide 200X.

240

")

O
 

 

 

 

 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

  
 
  
 
 
  

UNCLASSIFIED -
Y-22557

' {
;
i
)\ | ;
:’
f Sy ' 9
¢

Figs 3.6.16. NaK-Side Surface of Tfii:e‘f:oivn in Figs 3.6:15. Etchant: 10% oxalic acid. 200X.

Grain-boundary * voids that extended through a
tube that was in the hot-header region are shown
in Fig. 3.6.17, The wvoids progressed to such an
extent that they cannot be definitely attributed
to either fuel or NaK, even though the general
corrosion is 0,008 in, in depth, The usual gold-
colored metal deposit was found along the surface
exposed to the fuel in the cool section.

INCONEL CENTRIFUGAL PUMP THAT
CIRCULATED NaK :

J. H. Dchm R.S. Crouse

- An endurance test of an Inconel centnfugal pump
that circulated NoK was _terminated following
upproxumately 4400 hr of operation primarily at
1200°F,  -For some time before the ¢lose .of the -
test, there were indications that lubncatlng ol
had been leaking past the lower shafi sea| and into
the NaK stream. ‘Upon shutdown, extensive earbon
deposits were found lmmedm:ely below the lower -
pump seal; “and all wetted parts of the pump were

dark ond tornished. Metallographic examination

was therefore made of some of the wetted parts

to determine the nature of metallurgical changes
which might have occurred during the test, The

- examination revealed a heavy carbide precipitate
" to a depth of approximately 5 mils along all Inconel

surfaces exposed to NaK. The extent of carburi-
zation, which is illustrated in Fig. 3.6.18, was not
sufficient to have measurably affected the me-

chanical properties of heavier Inconel sections,
although it was necessary to replace all thinner
- sections, such as bellows and diaphragms.

| WELDS OF TEST COMPONENTS
- G, M. Slaughter
| Thermul-Convechon Loop Welds

- P, Patriarca

A metallographlc ‘study of the fused-salt corro-

snon resistance of saddle welds removed from a

Iclrge number of thermol-convecflon loops has been
madé. Brief operating histories of these loops are

- presented in Table 3.6,1.. No evidence of inter-

granulorf corrosion of improperly oriented grains of
the type found in the welds of Black, Sivalls &

Bryson heot exchanger No. 1 (type - IHE- 8) was
- 'observed

 

E. A. Franco-Ferreira, ANP Quar. Prog. Rep. June 30,
1957, ORNL-2340, p 247.

241

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
L Y-23494

 

 

 

 

 

 

 

 

 

. - .
L W - .
T gy . gt
L . v

 

 

. .
el g -

 

 

 

 

 

 

 

 

 

Fig. 3.6.17. °Tube Wall from Hot-Header Region.
Complete penetration of the tube wall was found.
Etchant: 10% oxalic acid. 200X.

Forced-Circulation Loop Welds

Segments of an electric-resistance-heater lug
removed from a nickel-molybdenum alloy forced-
circulation loop were obtained and examined metal-
lographicatly for weld-metal corrosion. The alloy
(17% Mo—7% Fe—bal Ni) tubing had been welded to
a Hastelloy B lug adapter with Hastelloy W filler
wire. The loop operated for 1000 hr with fuel 107.

242

 

RN UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

- Fig. 3.6.18, Curb-uriiufion Which Occurred on Inconel
Surfaces Exposed to NaK in Centrifugal Pfimp Which
Circulated NaK for About 4400 hr at Approximately
1200°F. The NaK was contaminated with oil following
a leak in the pump seal. Etchant: aqua regia. 250X.

Welds from both the upstream and downstream
ends of the loop were mounted and examined. The
samples were sectioned along the longitudinal axis
of the tubing to permit a composite evaluation of

the joints. No measurable corrosion of the welds

was evident in any of the samples examined,

Heat Exchan ger Welds

The results of metallogrophic examination of
fused-salt corrosion of welds and brazes,and the
fused salt and NaK corrosion of tubes in the Black,
Sivalls & Bryson heat exchangers types IHE-8 were
reported previously.2 Tube-to-header welds from
both the NaK inlet and outlet header have now
been examined to determine the extent of NaK
corrosion. Attack to a depth of approximately
0.003 in. was prevalent in the NaK inlet header,
and in some cases attack to a depth of 0.005 in.

 

R. J. Gray, ANP Quar. Prog. Rep. June 30, 1957,
ORNL-2340, p 244. & el

o
 

 

 

was observed. Typical weld-metal corrosion is

shown in Fig. 3.6.19, and the extent of the corro-
sion in an adjocent tube wall is shown in Fig.
3.6.20. Similar specimens taken from the NaK cut-

PERIOD ENDING SEPTEMBER 30, 1957

let header are shown in Figs. 3.6.21 and 3.6.22.
The extent of corrosion is, of course, much less
in the cooler regions. The amount of mass transfer
was not determined in this study.

Table 3.6.1. bpémflng Histories of Thermal-Convection Loops Examined for Weld-Metal Corrosion

 

 

oo Loop Moteri Fusedaltt Operoting Temperatre T
765 Inconel | 30 1500 1500
779 Inconel 30 1400-1600 2000

' (cycle)
876 Inconel + Inconel 30 1500 500
cast insert
1010 Inconel 30 1500 1000
792 lnconel | 130 + 0.2% ZiH, 1500 500
788 Inconel 304 0.5%ZrH, 1500 500
789 Inconel S 30 + 0.6% ZrH, - 1500 500
790 Inconel - 30 + 0.7% ZrH, 1500 500
791 Inconel 30 + 0.8% ZrH, 1500 500
802 Inconel : * 30 under N, atm 1500 500
786 Inconel | 730 + 2.96% addition 1500 1500
785 Inconel 30 + 2.23% addition 1500 500
830 Inconel 4 1500 500
831 Inconel A 1500 500
734 Inconel 82 1500 500
735 Inconel 82 1500 1500
841 Inconel 82 1500 500
842 Inconel - 82 1500 500

843 lnconel o9 1500 500

844 lInconel o 1500 500
847  lnconel . 93 1500 500
845 lnconel 94 1500 500

846 ' lnconel o4 1500 500

839 Inconel 95 1500 500
840 Inconet 95 | 1500 500
798 Inconel 107 + 0.94% addition 1500 500

243

 

 
 

 

 

ANP PROJECT PROGRESS REPORT

Table 3.6.1 {continued)

 

* Fused-Salt*

Time

 

LoopNo. Loop Material Crotaned R
797 inconel 107 + 1.5%.alddition 1500 500
793 lnconel 107 + 1.83% addition 1500 500
799 Inconel 107 + 1.83% addition 1500 - 500

- 796 Inconel 107 + 3.22% addition 1500 500
806 Monel 107 1500 500
808 Monel 107 1500 1000

N Hastelloy B 30 1500 1500
866 Hastelloy B 30 1625 1000
814 Hastelloy B 107 1500 500
816 Hastelloy B 107 1500 2000
826 Hastelloy B 107 + 2% addition 1500 1000
873 Hastelloy W 30 1500 1000
857 Hasteloy X 30 1500 1000

1014 76% Ni-24% Mo 30 1500 831

1015 76% Ni-24% Mo 30 1500 1000

1004 74% Ni-26% Mo 30 1500 791

1007 75% Ni-20% Mo—5% Cr 30 1500 1000

1008 75% Ni-20% Mo 5% Cr 30 1500 240

1006 77% Ni—20% Mo~3% Cr 30 1500 1000

1013 85% Ni-15% Mo 30 1500 1000

 

 

®*Fused-salt composition:
No. 30 - Na F--ZrF[UF4 (50-46-4 mole %).
No. 41 - NaF-ZrF ,-UF , (63-25-12 mole %).
No. 82 - NaF-LiF-ZrF4-UF4 (20-55-21-4 mole %).
No. 91 - NaF-LiF-ZrF‘-UFA (53-35-8-4 mole %).
No. 93 — LiF-ZrF ,-UF , (50-46-4 mole %).
No. 94 ~ KF-ZrF",-UF4 (50-46-4 mole %).
No. 95 — Rl:F-ZrF4-UI=4 (50-45-4 mole %).
No. 107 = NaF-KF-LiF-U F 4 (11.2-41-45.3-2.5 mole %).

244
 

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED

 

mcm;s,’

 

0.02

 

.

jc.03

 

 

 

 

 

Fig. 3.6.19. - NeK. Corroslon in Tube-tc-Heuder Weld of NoK Inlet Header of Black, Sivalls & Bryson NaK-to-Fuyel

 

 

 

 

 

* Heat Exchcnger Type lHE-& ‘Etchants eleciro!yfic oxclic ccid. IOOX- Confidential with-coptiony
o
[ 23]
e
x
O
z
0.02
»
./ Figs 3.6.20. NcK Corrosion in Tube Wall Adjacent to Weld Shown in Fig. 3.6.17. 100X« Etchant: electrolytic
oxalic acid.
| 245

 

 
 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

S

i
Py

 

 

 

 

Figs 3.6:21s NaK Corrosion in TubestorHeader Weld of NaK Outlet Header of Bleck, Sivalls & Bryson NaKeto=Fuel
Heat Exchanger Type IHE-8. Etchant: electrolytic oxalic acids 100X, (Cenfidentietwithcaption)

 

 

       

 

 

UNCLASSIFIED
"
= 10§
T
T
=
0.02
’ Tl
; C : v . e aner ok wi oL St R S A S AT
lo.o3
|
!
1
*
e ©
o
, e *5si : A T e
’ S e e S DRI s AT R N .&-.?s‘\i\f.qfi_‘? A

Fige 3.6:22. NaK Corrosion in Tube Woll Adjacent to Weld Shown in Fig. 3.6.19. Etchant: electrolytic oxalic
acids 100X,

246

 

 
 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

3.7. NONDESTRUCTIVE TESTING
R. B. Oliver

EDDY-CURRENT MEASUREMENTS OF METAL
THICKNESS '

R. B. Oliver | J. W. Allen
R. A. Nance

The Metal Identification Meter (MIM-1) developed'

for the sorting of Inconel, austenitic stainless
steels, and the common Hastelloys was success-
fully adapted to the thickness measurement of thin
sections of these alloys. A simple circuit change

allows adjustment of the scale range so that finite

thicknesses of these alloys, up to 0040-|n., can
be measured with a maximum error of £0,001 in.

The instrument was used to monitor the wall thick-
ness of a type 347 stainless steel cylinder while -
the wall was being lathe-turned from the original

‘/4 in. to 0, 020 in. The stainless steel cylinder,

which was 3/2 ft long and 6 ‘in. in diameter, was
fitted onto a wooden mandrel. - Both the Metal
Identification Meter and a resonance  ultrasound
thickness gage showed the finished wall thickness
to vary from 0.013 to 0,40 in. The readings made
with the two instruments agreed to within 0.001 in.
at all times. . The Metal ldentification Meter and

the resonance ultrasound instrument are both ex- -

cellent instruments for measuring metal thickness

when only one side of the material is accessible.
When applied to metal sections thinner than

0.020 in., the eddy- t method embodied in .
" o eddy-current methor emboclec. in:. "“make a similar.analysis of the impedance of a probe

“coil ‘in proximity with a flat metal plate have not,
~ in the past, been successful, but by making several -

the Metal ldenhflcahon Meter is more accurate

than the resonance ultrasound method, and, when
applied to section flucknesses greater than 0,060
or. 0 070 - ine, the resonance ultrasound method is.

 

‘has been completed.
-stable than its predecessor,?

more accurate. The meter is compact, portable,
and simple to operate, and it requires no coupling
between the probe and the work surface. The
values of several components in the circuitry of

the eddy-current instrument can be changed so that

measurements can be made of thicknesses of metals
having conductivities beyond the range of detection
of the present model of the Metal ldentification
Meter.

‘A new instrument, illustrated in Fig. 3.7.1, for
the measurement of sheet and cladding thicknesses
This instrument is more
being nearly free
from drift, and is more sensitive and versatile in

that it will accommodate a wider range of probe-

coil types and of test frequencies., The new in-

strument is being used in studies of the optimum
test parameters for measuring various combinations
of cladding and core materials.

MATHEMATICAL ANALYSIS OF EDDY-CURRENT
PROBE COILS

J. W, Allen M. O. Chester®

Numerous mathematical studies of the impedance
of a coil surrounding a cylindrical red or a tubular
object have been made that form a rational basis
for the selection of test parameters. Attempts to

drastic '_assumptions regqrding- _the field of the

 

2. w, Allen and R. A. Nance, ANP Quar. Prog. Rep.

R. B. Oliver, J. W. Allen, dnd R. A. Nance, ANP March 31, 1957, ORNL-2274, p 227.

- Quar. -Prog,: Rep. ]ur_:e’i _30,.' 1957, ‘_ORNL-2340,'-_p '25§.’ o 3St.lmmer employee from Comell University.

247

 
 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

L W T A
IS TV L) . )

L R T ETY

. - EDDY-GURRENY =
B 1103 LYY
Marx & B Y

N D.RNL METALLURGY DIVIBION

 

 

 

Fig. 3.7.1. Eddy-Current Thickness Gage, Mark 1l, Serial 1.

248

UNCLASSIFIED -
Y-:23439

 

 

 

 
 

 

 

"

probe coil a semi-empirical mathematical expres-
sion has been developed for a single-turn coil
near an infinitely thick metal plate. The equation

 

 

PERIOD ENDING SEPTEMBER 30, 1957

The curves of Fig. 3.7.2 were plotted from Eq. 1
by using the terms k222 and v as separate param-
eters even though they are related. |t should

 

 

 

 

 

 

 

R = a-c res:stanceof conl

© = 2r times the frequency of the coul

L = inductance of the co:l in the presence of B

the metal

oL = the reactance of the coil in the presence
of the metal, '

wL, = the reactance of the coil in air,
j= J—-_]-l‘ | o

k2 = Jopryo,

po = 4mx 10~ -7, = permeoblhty of nonrnagnehc’

materials in mks units, .

o = the
mhos/meter,

a = the radws of fhe equwalent smgie-turn -

- eoil,
v.= I/a: , L o
"1 = the “lift-off'* or spoc:ng between equw-
- alent single-turn co:l und fhe metul
rsurfc:ce. : : :

| The term K is determmed by the relohonshlp of‘ |

the followmg equahon' '

 

@ fo(Kd>.=[.

A2
” r
v+ 4

in which Jq is the Bessel function of zerd‘erder.

and of the first kind,

conductivity of the l'\:léfql,"ini,n;».

is: also be noted that these curves are universal,
R+jolL 1 | L
(1) L "7 1+ — =(
@iy (ka)2 : /2 |
| (Ka)? o
1 "1 va [ g 2 v+ 2+ 1172
S ll=—=+— = [1+— + In
| L | _4 .vz_‘ o v2 v L2\2 | 1/2
1 | 1 1o . 2 3 +
teom g — - q - " .
[ wka)? ]2 | 1 1
— - —{—+ In —
(Ka)? 2 2
o - J
~in whlch ~since . the parameters are nondimensional and may

-be calculated for any given test condition. For
practical testing purposes, the equivalent “‘single-
turn radii’’ and the “‘single-turn lift-off’’ values

‘may be calculated by a graphical technique on the

curves shown in Fig, 3.7.2. In general, very good

agreement has been obtained between a con-
- siderable amount of experimental data and the
curves. More experimental data will be obtained

and compared in order to increase the confidence

- level in applying these calculations.

The .problems involved in obtacining a mathe-

 matical expression for the impedance of probe

- coils in the presence of plates haying finite thick-
ness and in the presence of a cladding-core couple
_are much more difficuit and more complex than

those presented by this case of an infinitely thick
- plate.  There is some possibility, however, that
~ these cases can also be solved. The results of

- the  theoretical probe-coil studies should be
vuluoble in the future application of eddy-current
_probe coils to the measurement of metal thickness

r_cnd of the thickness of cladding over a core,

INSPECTlON OF TUB!NG

R. W, McClung - 'R. B. Oliver
" Both the encircling-coil eddy-current method and
the immersed-ultrasound pulse-echo method were
used to inspect approximately 18,500 ft of Inconel
tubing during the quarter, The tubing ranged in

249

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED

ORNL-LR-DWG 23097
1.00

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60 19 09 |
2 33 kY Juper
0.90 =4¢ X0 (PERMEABILITY) _]
. #o
47 OF NONMAGNETIC MATERIAL)
w= ANGULAR APPLIED
7 FREQUENCY

- 080 hO J ' o= CONDUCTIVITY {mhos/meter] |
g Vo 25 2= "EQUIVALENT" COIL RADIUS
3 O | {meters)
o O - _
z £ 14.2 4

070 Y A= "EQUIVALENT“COIL LIFT- —]
? 0?‘0 190 OFF (meters)
5 28.5
= %
o« o, Vara

060

& 105.2-—1k2 02|
o |
364.5
¥
050 15
0.40 ‘
o 0. 0.2 03 04 05 06

RESISTIVE COMPONENT

Fig. 3.7.2. Calcu’lofgd Reldiive Impedance of a Probe
Coil Above a Conductor with Infinite Thickness.

size from ;16 in. OD with a 0.025-in. wall to
'/2 in, OD with 0.065-in. wall. The rejection rates
were dependent on both the quality required and
on the minimum usable length, The average re-
jection rate was about 5%, with variations for
individual batches running from as low as 2%
to more than 10%. Factors that confribute to the
lower average rejection rate in comparison with
the rates previously reported were re-polishing
and re-inspection of ultrasonic rejects and the
removal of magnetic particles by a phosphoric
acid freatment,

INSPECTION OF PIPE
J. K. White R. W. McClung

Approximately 650 ft of Inconel pipe was in-
spected by the immersed ultrasound method during
the quarter with less than 1% rejection.  About
500 ft of the pipe, both 2}, and 3% in. sched 40,
was inspected with resonance ultrasound methods
in order to determine the wall thickness. For a
specified tolerance of +5% of the nominal wall
thickness, the rejection rate was approximately
90%; but for a standard tolerance of +10%, the
rejection rate decreased to approximately 10% of

" the total quantity of pipe.

250

. PLATE INSPECTION
R. W. McClung

The immersed ultrasound method was used for
the inspection of more than 625 ft2 of CX-900
Inconel plate in thicknesses ranging from % to
1 in. The rejection rate was less than 1%. Since
the inspected plate was, in general, larger than
the required size and rejected areas could be left
in the trim, it is difficult to accurately estimate
the true rejection as a result of defects,

CORE SHELL INSPECTION
R. W. McClung

Immersed ultrasound methods were developed for
the inspection of the Inconel reactor core shells.
The core shells are inspected both for the presence
of laminations and for the presence of radial
cracks. For the detection of laminations, 5 mega-
cycle ultrasound is directed perpendicularly to the
surface and along the radius of a shell, and
*‘ringing’’ within the thin wall is established,
Any damping of the ‘’ringing’’ pattern is indicative
either of a change of wall thickness or of the
presence of a lamination, For the detection of
cracks, the 5 megacycle ultrasound is introduced
into the wall of the shell with an incident angle
of approximately 17 deg; any crack lying approxi-
mately perpendicular to the path of the ultrasound
as it travels through the metal annulus will reflect
the signal and indicate the presence of that crack.

Areas on a core shell that indicate damping of
the “’ringing’’ of the ultrasound are also examined
with resonance ultrasound techniques to assure
measurement of the shell thickness with an accu-
racy of approximately 0.001 in, and to differentiote
between possible laminations in the wall and
sections of wall thinner than specified. The
instaliation for the continucus medsurement of
the wall thickness of a core shell is illustrated in
Fig. 3.7.3, and the facility for ultrasonic examina-
tion for the detection of laminations and cracks is

shown in Fig. 3.7.4.

DETECTION OF UNBONDED AREAS IN PLATES
WITH RESONANCE ULTRASOUND

‘R. B. Oliver

Several of the boron-containing plates for the
ART shielding, which were fabricated by Allegheny-
Ludlum Steel Corporation, blistered when they were

]
 

 

 

 

 

 

 

»

 

PERIOD ENDING SEPTEMBER 30, 1957

= UNCLASSIFIED
Y-23440

Fig. 3.7.3. Facility for Wall Thickness Measurements of Core Shellis by the Ultrasonic-Resonance Method.

(Socres-sith-capion)

heated for formmg. The boron carblde-copper core

was separated from the type 430 stainless steel
cladding- by @ copper diffusion barrler,‘and the
blisters indicated large . unbonded areas at the

-mterface ‘between the cladding .and the diffusion’

bacrier,  Since only. large unbonded areas formed

 

blnsters, a nondesfruchve inspection method was',
needed in order to detect all unbonded areas before
further processmg., The resononce-ultrasound e
method was selected as the method most suited
for measurement of these thin sections, In order to -

use the ulirasomc ‘method, the #ansducer ‘is sup= -
plied wnth a frequency-modulcted sngnal and
resonance is established for each case that the

instantanecus frequency satisfies the equation:

nV-
Tz
where
- f = frequency (106 cyc Ies/sec),' '
V = velocity of sound (105 in /sec),
t = fhe ihlckness of the ob|ect (m.),
n ‘= asmall, whole integer.

" Either the fundumenfal frequéncy n = l) or -the
dlfference between any two successive harmonic
frequencies fixes the thickness of the test object,

251

 
 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
Y¥-23476

 

 

Fig. 3.7.4. Facility for Ultrasonic Inspection of Core Shells for Laminations and Cracks. (-Gemvm-fl-rfl'phan')

252

 

 

 

 
 

 

 

L]

and any change in thickness will cause a shift in
these resonance frequencies.. A lamination or an
unbonded area will appear as a change in thickness
and may be readily detected if the defect area is
at least one-fourth the transducer area. The in-
strument presents the amplitude of the resonance
pecks on the vertical sweep of a cathode-ray os-
cilloscope, and the frequency values, or thickness
dimensions, are displayed on the horizontal sweep
of the oscilloscope.

A modulation range of 2 to 4 megacycles was
chosen, since this range would produce two
harmonic resonances in the full 0.100-in. plate
thickness but would be too low to produce a reso-
nance in the 0.010-in. cladding layer. Also, in
this frequency range, on unbonded area in the
interface near the transducer results in a complete
loss of signal; while an unbonded area in the
interface on the opposite side from the transducer
results in a pronounced shift in the location and
spacing of the resonance peaks. This shift in

PERIOD ENDING SEPTEMBER 30, 1957

the resonance frequencies is equivalent to a
change in thickness equal to that of the cladding
layer on the far side of the plate. If the unbonded
area approaches the area of the transducer, or is
greater, the set of rescnance lines representing
the full plate thickness will disappear oand a new
set of lines will appear; if the unbonded area is

less than the area of the transducer, both pairs of

resonance lines will be on the screen, A signal
gate is available on the instrument to aliow the
incorporation of visible and audible signals to
indicate the occurrence of an unbonded area,

‘Several of the larger unbonded areas detected
with resonance ultrasound and confirmed by
stripping the cladding layer from the plate are
shown in Fig. 3.7.5. The correlation observed
between the nondestructive ond destructive tests
indicates that the resonance-uvltrasound method is
a simple ond reliable way to locate unbonded
areas in this type of plate.

 

Fig. 3.7.5. Unbonded Areas in Stainless-Steel-Clad B ,C-Cu Plates.

253

 
 

 

 

 

 

 
 

 

 

 

 

Part 4
RADIATION DAMAGE
G. W. Keilholtz

254

 
 

 

 

 

 
 

 

 

4.1. RADIATION DAMAGE
- G.W. Keilholtz

EXAMINATIONS OF IRRADIATED -
COMPONENTS AND MATERIALS
A. E. Richt

N. A, Carter W. B. Parsley
C. Ellis E. D. Sims
E. J. Manthos R. M. Wallace

MTR In-Pile Loops

Disassembly of MTR in-pile loop No. 6, described

previously,! has been completed except for re-
moval of the pump impeller from the impeller
housing.  Thirty metallographic specimens were
obtained from the fuel circuit, and 27 of the speci-
mens have been examined. |t appears from the
results obtained thus far that the maximum depth of
corrosion is 3 mils. Specimens from the outlet
side of the nose coil showed the greatest pene-
tration. Specimen 672, which was taken from the
outlet side of the nose coil approximately 1 in.

downstream from the location of thermocouples 7

and 8, is shown in Figs. 4.1.1 and 4,1.2,  Speci-
men 651, which was taken at the location of thermo-
couples 9, 10, and 22, on the nose outlet side, is
shown in Figs. 4.1.3.and 4.1.4. Specimens 672 and
651 both show the maximum depth of attack of
3 mils... A typical specimen which showed little or

no corrosive attack is shown in Flgs. 4,1 5 and
4.1.6. This specimen, No. 641, was taken approxi-

mately 13 in. from the face of the pump on the nose
inlet side,

A remotely operated lathe will be used to' open-

the impeller housings of the pumps removed from

in-pile loops Nos. 4, 5, and 6 so that the impellers

can be examined. ' The impeller housing of the

pump from foop No. 5 will be opened first, since its
activity level is much lower than the activity

levels of the other two smpeller housmgs. .

Moderal‘or Materiul s |

The three Cc:psules__:rrudtated in the MTR high-" |

temperature moderatorematerials testing assembly
were examined,? ‘The test assembly, after removal

 

C. C. Bolta et al,, ANP uar. Prog. Rep. Sept. 10,
1956 ORNL-2157, p 81. ¢ P P

2E0r details of test conditions see J. A. Conlin, ANP
Quar. Prog. Rep. June 30, 1957, ORNL-2340, p 107.

from the container, but with the capsules in place,
is shown in Fig. 4.1.7. The largest capsule con-
sists of a graphite slug enclosed in a nickel can,
and the smallest capsule consists of ZrH jacketed
in molybdenum. The remaining capsule consnsts of
three BeO slugs in an Inconel can. The thermo-
couples on the BeQO capsule were all intact, but
some of the thermocouples on the graphite and

/ ZrH capsules were broken off or damaged. Di-

‘fl‘i‘énsions of the three capsules, taken before and
after irradiation, are compored in Table 4.1.1.

The three BeO slugs were removed from the
capsule by cutting the welds at the end caps and
then pushing the slugs out of the can. Dimensions
of the BeO slugs before and after irradiation are
compared in Table 4,1.2. No evidences of swelling

Table 4. 1.1, Dimensions of Moderator«Material
Capsules Before and After Irradiation
in the MTR

 

Diameter of Can*
(in.)
Before After
Irradiation irradiation

Capsule

 

 

- Z?H" in a molybdenum can Not available 0.5366

0.5367
0.5360
0.5363

1.0412-1,0425 1.0417
1.0404
1.0402
1.0404

BeO in an Inconel can

Graphite in o nickel can  1.289-1.290  1.2914

1.2905
1.2900
1.2897
1.2894
1.2885

 

*Measurements were tocken at evenly spaced points
along the length of the capsule,

- &y 257

s

{‘\o

 
 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

Fig. 4.1:2. Specimen 672 After

 

Etching. 250X.

: UNCLASSIFIED
il RMG 1815

 

UNCLASSIFIED .
RMG 1816
 

PERIOD ENDING SEPTEMBER 30, 1957

 

  

UNCLASSIFIED
-~ RMG 1817 .

    

 

[ ]

i - . F . - ; :: . : ?a- ! . f
* & 'z' [T ‘e . *
. #* B 95 . > . - ‘
. " . » & . e " S

] - . &
; Flg- 4.1.3. §
| x

UNCLASSIFIED
' RMG 1818
»

-~
|
!

 

Fig. 4.1.4. Specimen 651 After Etching. 250X,

259

 

 
 

ANP PROJECT PROGRESS REPORT

 

 

Fig. 4.1.6. Specimen 641 After Etching. 250X,

260

 
   

 

8 UNCLASSIFIED. | -
3 PHOTO 41095

 

 

o Fig. 4.1.7. Copsules Containing Moderator Materials Shown During Disassembly After Irradiation in the MTR.
O~

LS6L ‘08 Y39WIL4IS ONIAGNI A0id3 d

 
 

 

 

ANP PROJECT PROGRESS REPORT

or cracking were noted on the BeO slugs, which
are shown in Figs. 4.1.8 and 4.1.9. Slug No. 1 was
at the top of the capsule and thus was farthest from
the reactor.

An attempt was made to remove the graphite slug
from its can in a similar manner, but it could not
be pushed out easily. Therefore, the can on the
graphite slug will be slit open when the ZrH,-
containing capsule is sectioned on the remote
milling machine.

Table 4.1.2. Dimensions of Three BeO Slugs Before
and After lrradiation in the MTR

 

Length - Diameter
Slug (in) (in.)
Number Before  After Before After

Irradiation liradiation Irradiation  Irradiation

 

 

 

 

1 0.8325 0.8297 0.9964 0.9953
0.8298 0.9957
2 0.8318 0.8315 0.9950 0.9938
0.8316 0.9940
3 0.832 0.8317 0.997 0.9967
0.8317 0.9964

UNCLASSIFIED

RMG 1821

   

(a) (&)

  

ARE Specimens

Six specimens from the ARE have been examined.
Specimen R8, a section from the pressure shell
wall, including a weld, is shown in Fig. 4.1.10. A
crack may be seen at the weld junction, and there
are several voids in the weld area.

A section from a serpentine bend in a fuel tube in
the center of the reactor, specimen R7, is shown
in Figs. 4.1.11 and 4,1.12. There was subsurface
void formation on the interior wall to a maximum
depth of 3.5 mils, but the penetration was not

~uniform. Some parts of the wall showed deeper and

more . dense penetration than others. The outer
wall of the serpentine bend, which was in contact
with sodium, showed what appeared to be a mass
transfer deposit (Fig. 4.1.13), in addition to some
subsurface void formation, The deposit had plated
on the wall to a maximum thickness of 1 mil.

Specimen R3, which was also taken from a ser-

~pentine bend ina fuel tube in the center of the core,

also showed some subsurface void formation;
however, the density and depth of penetration were
not so great as for specimen R7. The areas of
attack were localized, and some areas of the wall
showed no attack, as may be seen in Figs. 4.1.14
and 4.1.15. No deposit similar to that found on
specimen R7 was noted on the exterior wall,

Specimen F5, which was taken from the inlet of a
fuel-to-helium heat exchanger, was cast in epoxy
resin so that the fins on the tube would not be

UNCLASSIFIED
RMG 1823

UNCLASSIFIED
RMG 1822

 

{c)

Fige 4.1.8. BeO Slugs After lrradiation in MTR. (a) Slug No. 1, which was farthest from reactore (b) Slug No. 2,
{c¢) Slug No. 3, which was closest to reactors 2X. Reduced 5.5%.

262
 

i
i
1
i

 

 

 

i
!

 

 

2

g

 

PERIOD ENDING SEPTEMBER 30, 1957

3 . UNCLASSIFIED

 

 

Fig. 4, 1.9. Opposite Ends of BeO Slu§$ 1, 2, and 3 After Irradiation in the MTR. 2X. Reduced 32%.

UNCLASSIFIED
- RMG-1795 .

 

Fig. 4.1.10. Section from ARE Pressure Shell, Including o Weld. 5X. {Secretwithcoptiory

' 263

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

UNCLASSIFIED

 

- RMG-1828 . - ©

(Eorrat with-captiond

UNCLASSIFIED

i RMG-1827 -

 

Fig. 4.1.12, ‘Section Shown In Fig. 4.1.11 After Etching. 250X. (Secretwithreeption)

264

 

 
 

PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

Fige 4.1.13. Mass Transfer Deposit on Outer Wall of Fuel Tube at Serpentine Bend. This surface was in contact
with sodium during ARE operation. As polisheds 500X. (Sectet with.seption)

 

 

 

 

 

 

Fige 4.1.14. Section of Specimen R3 Taken from Serpentine Bend In Fuel Tube in ARE Core. Etcheds 250X.
{Sesret-wiirtuption)y—

265

 
 

 

 

ANP PROJECT PROGRESS REPORT

 

B UNCLASSIFIED |
"RMG-1830

Fige 4.1.15. Another Section of Specimen Shown in Fig. 4.1.14. Etched. 250X. (Secarwmmrrasusny

damaged during cutting. The fin-to-tube wall
junction is shown in Fig. 4,1.16. (In the ARE fuel-
to-helium heat exchangers the fins were helical
strips placed in grooves on the tubes. The strips
were not brazed to the tubes.) The interior wall of

the tube showed subsurface voids to a depth of

4 mils, as shown in Figs. 4.1.17 and 4.1.18.
Specimen F9, which was taken from the middle of
a bend in a fuel-to-helium heat exchanger, also
showed subsurface voids to o depth of 4 mils, as
shown in Figs. 4,1.19 and 4,1.20. Specimen S6,
the bottom bellows from sodium valve U-23, was
also cast in epoxy resin. No cracks in the bellows
folds were found, and only a slight roughening of
the inside surface was noted, as shown in Fig.

4.1.21.

CREEP AND STRESS-RUPTURE
TESTS OF INCONEL

_ J. C. Wilson
W. E. Brundage N. E. Hinkle
W. W, Davis J. C. Zukas

MTR Experiments

An eight-specimen tube-burst creep test at 1500°F
in the MTR, as reported previously,’ resulted in

266

times to rupture that were much shorter than ex-

pected, and overheating of the specimens under
stress during reactor startup was thought to be a
possible explanation for the shortened rupture life.
Subsequent out-of-pile tests condu,c‘:ted‘-'by Douglas
and East of the Metallurgy Division in other
apparatus but under conditions that duplicated the
time-stress-temperature  history of the in-pile
specimens have indicated that the time to rupture
was not appreciably affected by the overheating
cycle. _

Another out-of-pile test in which the MTR dppd-.

ratus is duplicated is currently being conducted by
the Solid State Division. The time-temperature-
stress history of the in-pile specimens is again
being duplicated exactly. The apparatus used for
the Solid State Division test differs from the appa-
ratus used by the Metallurgy Division in that the
heaters and thermocouples are in the same chamber
as the specimens and contamination of the atmos-
phere is a greater problem. '

The creep apparatus irradiated in the MTR has

been cut from the irradiation plug and returned to

 

J. C. Wilson et al., ANP Quar. Prog. Rep. June 30,
1957, ORNL-2340, » 268, 2" g. Rep. ]
 

 

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
RMG-1832

 

 

o
f’

Lttt

L

X

fare ) M

 

 

 

Section of Specimen F5 Showing Fin-to-Tube Wall Junction in ARE Fuel-to-Helium Heat Exchanger.

Fig. 4.1.16.
Fins were wound helically in grooves and were not brazed to the tubes.

. 100X.

Etched

D

A
.
g2
<9
J
NR
u,

 

 

 

 

Interior Wall of Tube Shown in Fig. 4.1.16. As polished. 250X. (mflnupflmfi—

L.17.

4,

Figo

267
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

Fig. 4.1.18. Section Shown in Fig. 4.1,17 After Etching. 250X. (fzTret~with-<cuption)-

B UNCLASSIFIED |
© RMG-1835

 

 

" Fig. 4.1.19. Section of Specimen F9 Taken from Middle of a Bend in an ARE Fuel-to-Helium Heat Exchanger.
As polished. 250X. (Sscrat~withcaption) ' '

268

 
 

 

PERIOD ENDING SEPTEMBER 30, 1957

UNCL ASSIFIED
RMG- 1836

 

250X.

19 After Etching.

1.

Section Shown in Fi'g. 4,

. 20.

4

Fig.

>S-ect.I6tr| ofrs.peclmen. Sé 'I-'a.ken frorfi the Bottom Bellows of Sodium Valve U-23,

 

 

Etch’do 250 x.

4.1.21.

F‘gn

 

 

 

)

269
 

 

 

 

ANP PROJECT PROGRESS REPORT

ORNL. Examination of the specimens. for fracture

location and surface - contomlnoflon is scheduled-

in the hot cells. |

Temperature control in a recent MTR |rrad|cmon'3
of the elgh_t-spec_:lmen apparatus was . found ‘to ‘be -

difficult because of the steep gamma.ray heating

gradient in specimens oriented with their long -

axes along the direction of the gradient (specimens

parallel to beam hole uxis),' An out-of-pile mockup

of an apparatus with specimens mounted perpen-
dicular to the beam hole axis is being tested.

Only four specimens (instead of eight) can be.

accommodated in this apparatus, but improved
temperature  control should result during startup
and shutdown of the reactor.

LITR Experiments

Leaks of undetermined origin caused three stress-
corrosion experiments in the LITR to be discon-

tinved. Inconel specimens exposed to NaF-ZrF o

UF, (62.5-12.5-25 mole %, fuel 46) were being
tested at 1500°F and a stress of 2000 psi. One
specimen _had been in the reactor for 100 hr after
startup, but the other two were irradiated for less
than 24 hr. Since all the apparatus was leaktight
(as determined with a helium leck detector) before
irradiation, thermal stresses that occurred during
irradiation are believed to have caused opening of
brazed joints or tubing connections. The appa-
ratus now being constructed is being instrumented
to check the point of failure. The specimens that
failed are to be examined in the hot cells. The
apparatus that failed showed that improved furnace
construction had cllowed control of the temper-
ature at 1500°F for a wider variety of fuel loadings
and neutron fluxes than was possible earlier.

An experiment was carried out in beam hole
HB-3 of the LITR to investigate the effect of
irradiation on several kinds of thermocouple in-
sulation. Glass-insulated thermocouple pairs;
sheathed, swaged, MgO-insulated pairs; ceramic
insulators; and metal-sheathed, unimpregnated,
gloss-insulated pairs were tested. The specimens
were tested in air and in helium at temperatures up
to 1500°F. Measurements of electrical leakage
between the wires of a pair were made in each
case as a functlon of time, flux, mtegrated flux,
and occasnonal variations in temperature during
irradiation and during shutdown periods. The data
have not yet been correlated completely, but it
has been established that the insulating materials

270

.and practices used in past (and projected) LITR

and “MTR cpporctus “have not ‘caused errors .in
- temperature measurements. The only insulation
- that failed during the tests wasona metal-sheathed,
" unimpregnated, glass-insulated pair. The electrical
_ potential used during the tests was about 20 v. A
- metal-sheathed, swaged, MgO-insulated pair of the

type intended for use in the ART was included in

- these tests, buf a more definitive experiment that
~ will include a study of the effects of irradiation on

the resistivity of Inconel is to be conducted.

LITR YERTICAL IN-PILE LOOP
W, E. Browning

J. E. Lee, Jr. H. E. Robertson
M. F. Osborne R. P. Shields

The forced-circulation loop which was operated
in a vertical hole in the LITR for 235 hr, as de-
scribed previously,4 is being disassembled for ex-
amination and analysis of the fluoride fuel mixture
circulated therein. The loop has been removed from
the outer container, and the major subassemblies
have been detached. Examination of these sub-
assemblies is in progress. The supposition that
the pump became inoperative because of beormg
seizure was substantiated.

The two pump-shaft bearings were removed by
cutting laterally through the pump and shaft above
and below each bearing. The shoft in the bearing
nearest to the fuel could not be turned by the manip-
ulator, A black porous deposit was found on the
wall near the bearing that was, presumably, radio-
lytic decomposition products from the grease, The
shaft in the bearing just below the motor could be
tumed, but only a small amount at a time, and the
bearing above the motor turned freely. -All three
bearings had been lubricated with the same radiation-
resistant lubricant. The difference between the
behavior of the bearings is attributed to_the differ-
ence in dosages from the decay -of short-lived
fission gases as they diffused upward through fhe
pump.

Of the 55 thermocouples in the Ioop, 22 fctled
Seven of the failures were due to an open. circuit
in the platinum-rhodium leads. One thermocouple
failed because the bead came loose - from the
Inconel fuel tube, and 14 failed because both the
platinum and the platinum-rhodium lead wires had

 

‘w. E. Browning et al., ANP Quar. Prog. Rep. June 30,
1957, ORNL-2340, p 275.

i

 
 

 

 

 

 

 

 

broken. |llustrations of thermocouples that failed
are presented in Figs. 4.1,22, 4.1.23, ond 4.1.24.
These thermocouples were located on the fuel
tubing in the high-temperature region of the loop.
Three beads were used for each thermocouple in-
stallation, as shown in Figs. 4.1.22 and 4.1.24.
Both leads had pulled away from one bead shown
in Fig. 4.1.22, and only one lead remains attached
to another of the beads. The pair of wires that
pulled away from the bead during operation are
shown in Fig. 4.1.23. The tapered ends are typical
of ductile failure in tension. Wires that broke about
Y to 1/2 in. from the bead are shown in Fig. 4.1.24.
The ends of these wires also show the tapered
effect illustrated in Fig. 4.1.23. The mechanical
damage of the lead wires probably occurred during
thermal expansion of the loop.

- The platinum resistance heaters and their lead
wires were examined closely to determine the
cause of failure of 4 of the 22 heater circuits.
The heating elements were in good condition, and

 

UNCL ASSIFIED
RMG-1877a

  

 

PERIOD ENDING SEPTEMBER 30, 1957

. UNCLASSIFIED
RMG-1877b

Fig. 4.1.23, WIres"Tfiut Pulled Aweay _fror.n. Thermo=
couple Bead Shown in Fig. 4,1.22,

the failures appecred to have been in the lead
wires,’
" Particular ottention was given to the d:sfrlbuhon

of fuel and fuel vapor deposits in the pump. The

pump was sectioned longitudinally along the shaft,

~_ond -one-half is shown in Fig. 4.1.25. No vapor
. deposits were found in the region above the Graph-
itar vapor baffle,
- splashing of the fuel in the sump chamber. It is
. .inferesting ‘to compare the vapor deposfls with
- those observed in the out-of-pile pump,’ in which
,much Inrger deposns of ZrF, were found, The
,dnfference may be. attributed to the’ longer operating
~life of the out-of-plle test pump of over 3500 hr.
" Microscopic examination - of the - Graphitar vapor

There- was no evidence of

' baffle- showed. fhat the shaft had not. fubbed against

the baffle; the original tool marks on:the Graphitar

. "-::'-’r_..:-we'f.e undisturbed. The general appearance of the

 

Fig. 4.1.22, Thermocouple That Failed During Op-
eration of a Vertical In-Pile Loop in the LITR.

 

5W E. Browning et al., ANP Quar. Prog. Rep. March
31d 19]5;.4 ORNL-2274, p 255, esp Figs. 4.1.22, 4.1.23,
and 4.1,24,

271

 
 

 

ANP PROJECT PROGRESS REPORT

 

 

Fig. 4.1.24. Thermocouple That Failed During Op-
eration of a Vertical In-Pile Loop in the LITR Showing
Broken Lead Wires.

loop, as observed during disassembly, was goed.
The quartz tape that was used to insulate and bind
the wires was not discolored or damaged. One qir
annulus was slightly bowed. Some of the structure
of the loop may be seen in Fig. 4.1.26. The heater
section shown in the figure is in good condition.
No discoloration was found on any of the ceramic
heater forms.

Tests are being prepared of bearings designed to
withstand fission-gas radiation for longer operating
times and thermocouple installations designed to
prevent damage of the thermocouple leads during
thermal expansion of the fuel tubes. Samples of
minute deposits found at various locations in the
pump are being analyzed chemically and radio-
chemically in order to determine the distribution of
fission products in the fuel. Specimens of the
Inconel tubing are being taken from the fuel circuit
for metallographic examination to determine the
extent of corrosion in the loop.

272

MTR STATIC CORROSION TESTS
"H. L. Hemphill

“Two Inconel capsules containing the fuel mixture

W. E. Browning

- NaF-ZrF -UF, (53.5-40-6.5 mole %, fuel 44) were

withdrawn from the MTR after 720 hr of irradiation
at 1500°F under static conditions. These capsules
will be returned to ORNL for examination. Two
Hastelloy B capsules were inserted in the MTR, but

“operational difficulties made it necessary to delay

temperature cycling,

FLUX MONIfORING OF MTR STATIC
CORROSION CAPSULES

D. E. Guss

The average thermaleneutron-flux exposures of
several static corrosion capsules irradiated in the
MTR have been calculated from the Cof? disin-
tegration rate of the Inconel container material,
The average fluxes for four of the capsules were
also determined by the yields of Zr?5 and Cs'37
from U235 fission in the fuel contained in the
capsules. The values are presented in Table 4,.1.3.
Revised flux values based on corrected irradiation
times are given for three capsules.

RELATIVE CONTRIBUTION FROM THE
N16%(n,p)Co%® REACTION TO THE Co%®
DISINTEGRATION RATE IN
IRRADIATED INCONEL

D. E. Guss

The contribution to the Co%? disintegration rate
in irradiated Inconel arising from the nickel con-
stituent through the Ni%%(n,)Co%0 reaction was

investigated further,® In order to obtain a measure

of this contribution the Co®° disintegration rate
was determined for several pieces of vacuum-

melted carbonyl nickei (~0,002% Co) which had

- received @ 75-min irradiation in the lattice (hole

.C-46) of the LITR in September 1955. The Co%0

disintegration rate per gram of nickel was equiva-
lent to that which would have been produced by

1 x 103 g of cobalt. Even if all this activity .
were attributed to the Ni®%(n,p)Co80 reaction, that

is, if there were assumed to be absolutely no

 

6p, .E. Guss, ANP ar. Prog. Rep. June 30, 1957,
ORNL-2340, p 3760 2" !

 

 

 

 
 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

   
  

 

 

   
 

  
 
 

 

 

        
 

 

 
 

PHOTO 41397
- | " REGIONS WHERE VAPOR
“4E DEPOSITS COULD FORM
| . GRAPHITAR.
| VAPOR BAFFLE
it
FUEL SURFACE
g IMPELLER
’ FUEL OUT ~ FUEL IN
fi Fig. 4,1.25, Sectioned Pump from Vertical In-Pile Leop Operated in the LITR.

 

273

 

 

 
 

vLe

 

 

e

L6
AIR TUBES TO
LEGS OF LOOP

s -v"(j
%‘5?’ i I%%%’&%g

% o i
o
5 é}&" B
. {k;m .

A

P T i

G oy

 

 

P R
2 T g

 

 

Fig. 4.1,26. View of Partially Disassembled L.ITR Vertical In-Pile Loop.

 

L30d3Y $SIN90¥d L1D3Irodd dNV

 
 

 

 

"

PERIOD ENDING SEPTEMBER 30, 1957

Table 4,1.3. Thermal-Neutron Flux Yalues Obtained from Flux Monitoring
. of MTR Static Corrosion Capsules

 

Average Flux Average Flux

 

Capsule Discharge Irrod.iation ' Cal.cu.rlated from Co%0 Cfll:t.:lufed from
Number Date Time Disintegration Rate Fission-Product
_ (Mwd) in Inconel Capsule Yields
(n/cmz-sec) (n/cmzcsec)
x 1014 x 1014
227 12-29-53 367 1.39
232 11-18-54 1282 1.44
237 3-28-55 C 990 1.42
245  6-20-55 | 939 0.80*
252 7-7-55 535 2.36
253 3-28-55 990 1.60
274 6-20-55 | 939 0.78*
275 | 6-20-55 939 1.10%
316 - 7-5:56 1980 177 1.97
317 7-5-56 1980. 1.12 1.19
333 9-24-56 510 0.78 0.76

345 9-24-56 - s10

0.59 0.67

 

*Revised values based on corrected irradiation times.

cobalt present in the nickel, it may be seen that

the contribution to the Co®® activity in irradiated
Inconel arising from the Ni%%(n,p)Co%? reaction
would be negligible.” The Inconel used in fabricating
corrosion-test capsules contains about 1,5 x 10-3 g

of cobalt per gram of Inconel, and thus the con- -
tribution to the total Co%9 activity from the nickel
would amount to less than 0.5%. Further, the tests’
are run under conditions in which the ratio ofthe
fast-neutron flux to the fhermal-neutron flux |s- '

clbout unity,

" ETR.'IRRAD'IATION' OF MODERATOR MATERIALS

W E. Brownmg G. Somuels
- R. P. Shields

Preparqhons have continued for urudlahon mi

the ETR of moderator materials for use in high-

~ temperature reactors. * The irradiation request was

approved by the AEC and design details are being
worked out with ETR personnel. Fabrication of the

BeO test specimens has been started, and arrange-

ments have been made for fabrication of yttrium
hydride test specimens at GE-ANP. (For details

of design of equipment for the irradiations see
Chap. 1.4, “Engineering Design Studies.”’)

... EFFECT OF IRRADIATION ON
~ THERMAL-NEUTRON SHIELD MATERIALS

J. G. Morgan
~ M. T. Morgan P. E. Reagan
| ‘Boron Nitride

Boron nitride is similar to graphite in many re-

- spects, but it is white in color and much more re-

sistant to oxidation at high temperatures. Favor-
able results have ‘been obtained from radiation

damage tests conducted on boron nitride as a
_neutron shielding material. The tests were made,

primarily, to explore the general integrity of this
relatively new ceramic under irradiation and,

275

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

specifically, to determine weight, density, and
dimensional changes, Boron nitride was obtained
from two manufacturers (Norton Company and The
Carborundum Company) in powder form and 3/1 gmine~
oD, l'/z-in.-long test specimens were hot pressed
from the powder. The Norton samples are designated
6002, and the Carborundum samples are designated
Cl.

The samples were irradiated in an inert atmosphere
either in quartz ampoules or in Inconel capsules.
High-temperature irradiations were made in the
Inconel capsules, and the quartz ampoules were
irradiated at the temperature of the test reactor
water, The high-temperature tests were monitored
by thermocouples pressed against the outer surface
of the sample with leads extending through a
ceramic seal in the capsule top. The thermal-
neutron dosage was calculated from the activity
of a cobalt foil located against the sample.

Surface darkening of the samples duringirradiation
is thought to have been caused by a metallic
deposit. The coating was found to be primarily
magnesium, with some silicon and traces of calcium
and copper. The irradiated samples retained good
crystallinity, The results of x-ray diffraction ex-
ominations of samples irradiated in the MTR are
presented in Table 4.1.4. The conditions of the
tests and the physical property changes resulting
from irradiation are described in Table 4.1.5, and
irradiated and unirradiated samples are compared
in Fig. 4.1.27. The dimensicnal changes that
occurred during the tests were small, and the
changes that were detected could not be definitely
attributed to radiation damage, since boron nitride
is a soft material’ and the micrometers used scraped
material from the surface during measurements of
the samples. Both the weight and the density
changes were less than 1%. The burnup values
shown in Table 4.1.5 are average values calcu-
lated on the basis of neutron flux and irradiation
time. Boron isotopic analyses are being made.

These tests indicate that boron nitride is un-
damaged by irradiation at a temperature of 1800°F
to a thermal-neutron dosage of 1.3 x 1020
neutrons/cm?.

Physical property measurements are now being
made on the four samples irradiated at 70°F. The
gas evolved during irradiation will alsobe analyzed.

 

7The samples had a Mohs' scale hardness of 2 before

lrradluhon, which increased to 2.5 during irradiation.

276

Table 4, 1,4, Results of X-Ray Diffraction
' Examinations of Irradiated
Boron Nitride

 

Lattice Spqcing Dimension

(A)

Unirradiated

Plane |

 

Indices Original 6002-1 6002-4

 

Powder Hot-Pressed Sample
(002)  3.32 3.33 3.37 3.33 -
(100) 217 2,17 2.17 2.17
(101) 2.06 2.06  2.06 .
(102)  1.81 .82 1.82  1.82
(004) 1.66 1.68  1.66
(104) 1.32 1.32
(110) 1.25 1.25 1.25
(12) 117 137 1.7

 

Hexagonal Structure Axes

 

A 2.50 : 2.50

C 6.64 6.74* =

 

*The expansion in C is 1.6%.

UNCLASSIFIED
RMG- 1842

 

(a)

 

o)y »

Fig. 4.1.27. Hot<Pressed Boron Nitride (2) Unir-
radiated ond (b) Irradiated in the MTR at 1780°F to
an Average g0 Burnup of 4%. '
 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

- Tahle 4.1,5. Test Conditions and Physical Property Changes Resulting from
Irradiation of HotPressed Boron Nitride

 

.Sample Number

 

6002-1 - 60022 6002-3

6002-4 6002-8 cl-n Cl-12

 

Test facility  MTRA-28 LITRC-39 LITRC-39 MTR A-28 Out-of-pile LITR C-39 LITR C-39
. ‘ test

Test temperature, °F 1600 70 70 1780 1470 70 70

Irradiation time, hr 710 67 T4 606 600 1148 1148

Thermaleneutran flux, 5x 1013 1.7x10"  1.7x10'® 4 x 1013 1.7x10" 17 x10'3
neutrons/cmZssec | o ' '

Doss, neutrons/em? 1.3 x 1020 1,02 x10% 1.02x 1020 .72 x 1019 7.02 x10'? 702 x 107

Weight change, % 0.9 o e 0.26 - o

Density change, % None *# e 0.13 0.77 o B

Dimensional change, % flon’e e L w 0.1 None >k **

Average B'0 burnup, % '5.0." 48 48 4.0 None 3.8 3.8

 

*Weight-change measurement not made because chips were |osi when sample broke.

**Physical properties yet to be determined.

" Boren Carbide

Irradiation tests of two hotspressed boron carbide

tiles that were prototype samples of ART tiles

were completed. A sample designated J-1 was
irradiated .in the LITR C-39 facility for 1148 hr at

the reactor water temperature to a cajculated

average burnup of 1,4%. A sample designated J-3 . _
was irradiated in the LITR C-42 facility for 910 hr .. .
at 1500°F to an average bumup of 2.2%.. Post- -

urud:uhon exammahons of these scmples are in,,. ‘
_ -+ than the CaB,-Fe cermet, and there was some

~_ separation - of the core from the cladding. No
~ significant changes in structure of the core or the
cladding ‘were found. There was no noticeable
‘increase in fragmentation or porosity. . Longitudinal

progre ss.

. An_ ou?-of-plle test ‘was conducted in’ whlch a-
 third sample (J-2) was cycled 18 times: from 200 to
- 1600°F, "~ The . sample was ‘then "examined under
@ microscope af magnifications of 20 and 40, and '_
- no cracks were observed

Ce rmeis

in evacuated quartz ampoules cooled by the LITR
reactor water and examined after two different

bumups. The boron in both cermets was natural
boron. There were no structural changes as a re-

sult of irradiation to an average B'0 burnup of

. 19%, as shown in Table 4.1,6, but both materials

were damaged by irradiation to an average burnup

of 38%.

" The CaB,-Fe cefmet showed slight cracks

"throughouf the core matrix after the irradiation to
~'an average bumup of 38%, as shown in Fig. 4.1.28.

The BN-Ni cermet was more severely cracked

cracks in the sample irradiated to an average burn-

‘up of 38% are shown in Fig. 4.1.29, and the two

irradiated samples are compared at hngher magmfi-

e e . = . cation in Fig. 4,1,30.
Plates of BNoNl (103 wt % BN) and CaB -Fe-
(7.6 wt % CaB,) cermets clad with type 304 stum-fi 3
less steel have been irradiated at less than 345°F

Stainless-steel-clad B C-Cu cermets ‘have also
been studied. The first three specimens irradiated

“were fabricated by the ORNL Metallurgy Division.

They were 0.5 in. long, 0.1875 in. wide, and

- 0.102 in. thick. The cores of these specimens

277

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Takle 4.1.6. Effect of Irradiation on Dimensions and Hardness of BN<Ni and CaBéFe _
Cermets Clad with Type 304 Stainléss Steel |

 

 

 

 

 

 

 

Change After Average 10 Change After Average B_m .
. Burnup of 19% o " Burnup of 38%
Measurement (%) : @)
CaB,-Fe® BN-Ni% CaBy-Fe? BN-Ni®
Length o <+0.1 , <+0.1 +0.8 - +0.4
Thickness <+0.3 <+0.3 - +2.1 +5.4 .
Density <+0.2 - <40.2 ‘ -1.2 - =1.0
Hardness _ o :
Core +1.0° +100 | +36€ | 4204
Cladding : 425 . +25 +55 , - +47.6
2Typical sample. _
bAverage of three samples,
“Based on Dicmond pyramid hurdness values (4-kg load); all other hordness values are based on Knoop hardness
{1-kg load).
UNCLASSIFIED N " UNCLASSIFIED
RMG- 1447 B ’ o i ,V..,.:.R‘__‘G.ITIz
&*

Fig. 4.1.28. lrradiated CaB -Fe Cermets Clad with Type 304 Stainless Steel. (a) Average plo0 burnup of 19%.
(6) Average B0 burnup of 38%. 95X, Reduced 20. 5%

278
 

PERIOD ENDING SEPTEMBER 30, 1957

- UNCLASSIFIED
ORNL—LR-DWG 21074

  
   

 

1

  

        

| SIDE VIEW (4X) END VIEW (8X)
Flg; 4.1.29. BN-Ni Cermet Clad with Type 304 Stainless Steel Irradiated to an Average B0 Bumup of 38%.
Reduced 25%.
'UNCLASSIFIED S - UNCLASSIFIED -
RMG-U95 | 4o £ s oo TREE  RMG-1708
|
l

 

Fig. 4.1.30. Iradiated BN-Ni Cermets Clad with Type 304 Stainless Steel. (a) Average B' burnup of 19%.
(b) Average 810 burnup of 38%. 75X,

279

 
 

 

 

ANP PROJECT PROGRESS REPORT

were 6.6 wt % B,C dispersed in copper, and they
were 0.082 in. thick. The cores were clad on two
faces with 0,003 in. of copper followed by 0.007 in.
of type 430 stainless steel. The cores were ex-

posed on the four’edges of the specimens. The.
fourth specimen irradiated was a similar sample

fabricated by the Allegheny-Ludlum Stee! Corp.;
it was 1.5 in. long. The irradiation histories of

“these samples are given in Table 4.1.7.

Hardness measurements made on the stainless
steel cladding, the copper layer, and the core after
irradiation .are summarized in Table 4.1.8, except
that post-irradiation core hdrdness measurements

 are not-répartéd for samples whi‘ch‘crock'ed severly,

The results “of dimensional. and density measure-
ments dre presented in Table 4.1.9. While damage
to these samples appears to be a function of burn-
up ond temperature of irradiation, . the - thermal

~ Table 417 Results of LITR leradiation of Stainless-Steel-Clad B,C-Copper Cormet Samples .

 

 

Avérage Temperature - Average g0 : « S
lrradiation Capsule s Temperature Radiation Damage
Nomb of Irradiation Burnup Crel Eound
umber ©c) %) ycles = - Foun
227 (3 samples) 200 6 3 None; no cracks or
separation of core
from clodding
327 (3 samples) 300 19 7 None; no cracks or
separation of core
from cladding
127 (3 samples) 420 18* 19 Longitudinal core cracks
and separation of
core from cladding
527 (1 sample) 870 12.5 8 Longitudinal core cracks

and separation of

core from cladding

 

*Determined by isotopic analysis; other values were calculated,

Table 4,.1.8. Hardness Measurements of StainlesssSteel-Clad B4C-Cu Cermets Before and After lrradiation

 

Irradiation Capsule Location of

Average Knopp Hardness

Pefcenfage Change

(1-kg Load)

 

 

in Hardness
Number ' Measurement Unirradiated irradiated ofter frradiation
Sample Sample :
227 - Cladding 139 166 +19
Copper layer 64 ‘ 76 +19
Core 76 102 +34
327 Cladding 139 _ 215 - ~+55.
Copper layer 64 88 +38
Core 76 - 85 +12
127 Cladding 139 186 +34
Copper layer 64 65 -2

 

280
 

 

i
i
i
;

Table 4,1,9, Dimensiona! and Density Changes
as a Result of Irradiation of Stainless-
Steel-Clcd B 4C.'.-(.'.’u Cermets

 

 

 

"

 

s

Irradiation - _Dimens‘ional Ch‘i'-f‘.gé*l -Diensit-y
Capsule S (%) Change
Number | ength Width Thickness (%)
227** +1 +1 +1 +1 7
327 +1 +1.8 +2.0 =3.9 (butk
' density)-
127 +0.5 +0.8 +2.3 -2.3
527 +2.5 +3.5 +121 116

 

 

’
"

 

 

*Results are averages of values for three sump‘es, ex-
cept for capsule No. 527 which contained only one
sample. '

**The samples in éupsule No. 227 were sligfifly oxi~
dized, but there was essentially nc change as a result
of irradiation. o :

UNCLASSIFIED |
RMG-1832

 

PERIOD ENDING SEPTEMBER 30, 1957

cycling history is also important, In a composite
plate of this type, which is essentially black to
neutrons, most of the (n,0) reactions occur in the
first few thousandths of aninch of sample thickness.
The core becomes harder and more brittle and the
stainless steel cladding acts as a bimetallic strip
that causes uneven stresses. The uneven stresses
are believed to be the reason for a large part of
the splitting and warping. An unirradiated stainless-
steel-clad B,C-Cu sample is compared in Fig.
4.1.31 with a sample irradiated at 200°C to a 6%
average B! burnup. No cracks are evident in the
irradiated sample. The separation of the core from
the cladding and the cracks which occurred at a
higher irradiation temperature and a higher bumup
are illustrated in Fig. 4.1.32.

In  out-of-pile tests it was demonstrated that
thermal stress domoge can occur in the absence of
irradiation. More cycles were required to cause
damage in the out-of-pile experiment, but the

P GNCLASSIFIED |
RMG-1533

Fig. 4.1.31. Stainless-Steel-Clad B4C-Cu Cermets (a) Unirradiated and (5) lrradiated at 200°C to an Average

Burnup of 6%. 500X. Reduced 19%.

281

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

   

  
  
  
 

      
 
  

T8l N

. ! . ,a'pi. ::.. "\ i

B VB L e
.o e g

’ .‘) '.i

-~

UNCLASSIFIED
ORNL—-LR—DWG 21028A

Fig. 4.1.32. Stainless-Steel-Clad B4C-Cu Cermet lrradiated at 420°C to an Average B10 Burnup of 18%. (a) A
typical core crack. (b) Separation at interface of core and cladding. 250X.

irradiated samples had experienced a different
thermal-gradient history that may account for the
damage with fewer cycles, In the out-of-pile tests
the samples were cycled under vacuum by in-
serting and withdrawing them from a furnace. Two
temperature ranges were used for the tests: 75 to
650°C and 100 to 870°C. In the 75 to 600°C range
the heating rate was 300°C/min for the first minute
and 175°C/min for the second minute; the cooling
rate was 250°C/min for the first minute and
200°C/min for the second minute. In the 100 to
870°C/min range the heating rate was 960°C/min
for the first '/2 min, 400°C/min for the second

282

!& min, ond 75°C/min for the second minute; the
cooling  rate was 600°C/min for the first Y min,
200°C/min for the second I/2 min, and 150°C/min
for the second minute. One cycle was a temper-
ature excursion from minimum to maximum and retum
to minimum. Descriptions of the samples and the
specific conditions of each test are given in Table

4.1.10.

Examinations of the specimens (at a magnifie
cation of 500X) after thermal cycling showed that
the Norton Company sample had not cracked or
chipped. Sample No. 2, fabricated at ORNL, is

 
 

 

 

shown in Fig. 4.1,33. This view of the core-to-
cladding interface shows no evidence of cracks.
The scratches were made during hond polishing.
Sample No. 3 was found to have one small crack
about 0,006 in. long ot the interface between the
copper and stainless steel layers. In sample No. 4
there was complete separation of the stainless steel
cladding from the copper layer over about one-
third the area on one face, as shown in Fig. 4,1.34.
Inaddition cracks were found in the core ends about
midway between the clad faces, as shown in Fig.
4.1.35. Sample No. 5 was similar to sample No. 4,
except that there was slightly more separation at
the interface and there were more cracks. in
general, sample No. 6 did not show as severe
interface separation or as many cracks as found in
sample No. 5. It was noted that small cracks
initially present in the material were enlarged.
Some cracks were found that started at the core-
to-cladding interface and propagated into the core.
The sample was not constrained durmg cycllng,
but after the test it was bowed about /32 n. in
the middle. Apparently unequal stresses were
created by the bimetallic cladding.

Fig. 4.1.33.
less-Steel-Clad B

PERIOD ENDING SEPTEMBER 30, 1957

4

UNCLASSIFIED
RMG- 1868

 

Thermally Cyecled, Unirradiated, Stain-
C-Cu Cermet Sample No. 2.

Table 4.1, 10. Condmons of Thermal Cycling Tests of Composites Made Upof a B C-Cu Cermet Core,
e Layer of Copper, and a Layer of Stainless Steel

 

 

Tempe;'afure
Source Dimensions Sample Number of Range
{in.) Ne. Cycles ©C)
Nerton Company* 0.25 x0.25 x 1.0 1 i8 100-870
ORNL 10.1.x0.1875 X 0.5 2 20 75-600
| 3 20 75-600
20 100-870
4 20 75-600
38 100-870
5 20 75-600
e 58 100-870
© Allegheny-Ludium 0.1 x0.189 X 1.5 6 58 100-870
- Steel Corp. '

 

*ART prototype samples.

283

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

SR UNCLASS!FIED
BN RMG-1869

 

Fig. 4.1.34. Thermally Cycled, Unirradiated, Stain-
less-Steel-Clad B4C-Cu Cermet Sample No. 4, Showing
Interface Separation.

EFFECTS OF RADIATION ON
ELECTRONIC COMPONENTS

J. C. Pigg
C. C. Robinson 0. E. Schow

The general results that have been obtained in
the investigations ofthe effects of nuclear radiation
on semiconductor devices indicate that the barrier
phenomenon should be studied under very closely
controlled conditions. The equipment necessary
for such a study is being developed. The experi-
mental results obtained during the quarter are

described below.

Experimental Resvlts

The results of gamma irradiation of IN 38-A
point-contact diodes (greased and ungreased),
described previously,® have been analyzed with

 

85. c. Piga, C. C. Robinson, and O. E. Schow, ANP
Quar. Prog. Rep. June 30, 1957, ORNL-2340, p 268.

284

respect to annealing behavior, ond curves de-
scribing the behavior have been prepared. The
first effect studied was the temperature change of
the samples, as shown in Fig. 4.1.36. The initial
drop of the temperature may be of significance,
and no doubt the rise in temperature during the
first hour had some effect on the current through
the samples. In any event, the drop, or annealing,
in the current of the sample cannot be attributed
to temperature changes, since a rise in temper-
ature would be expected to increase the current
through the sample, and, in Figs. 4.1.37 and
4,1,38, it may be seen that the reverse currents
{(at 1-v bias) through the samples decreased with
time. The greased sample (Fig. 4.1.38) under-
went the most severe change, and it would be
expected that this change would be reflected in
the slope of the reverse saturation current, which
is indicative of the surface conditions of the
sample. In Fig. 4.1.39, however, it may be seen
that the slope of the reverse saturation current
does not show such a change. On the contrary,
the greased unit seems to have suffered less sur-
face change than the ungreased unit. At the end
of 37.5 hr of annealing, both surfaces were appar-
ently the same. An effect of surface coating on
the bulk characteristics is indicated in Fig. 4.1.40,
which shows the reverse current ‘Y’ intercept, or
the saturation current, of these samples. The
gamma-ray dosage the samples received was of the
order of 9 x 10'6 photons/cm?, which was sufficient
to cause bulk damage in germanium. Bulk changes
known to take place with a dosage of this magnitude
are all of such a nature as to decrease the bulk
conductivity of the sample. Speculation has led to
the possibility that a revision of the picture of
radiation effects in electronic barriers is in order,
but as of now there is insufficient evidence to
conclusively justify this revision. Experiments
are being planned to clarify this problem, however,
as described in a following section.

In view of the annealing observed in semicon-
ductor devices by moany workers, @ number of
questions have been raised as to the actual amount
of damage that may be introduced into a sample by
a given amount of radiation at a given flux level.
Such information is desirable in order to correlate
device data with data resulting from “studies of
bulk material. Many experiments have been per-
formed on bulk material in which the samples were
maintained at some low temperature, usually that of

 
 

 

 

 

»

3]

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED |
“_RMG-1870 -

 

Fig. 4.1.35. Thermally Cycled, Unirradiated, Stainless-Steel-Clad B4C-Cu Cermet Sample No. 4, Showing
Cracks in Core. :

UNCLASSIFIED
ORNL-LR—-DWG 22953

 

 

 

 

 

 

e

 

20
19 -
- P- :
§ 18
Ll
o«
=
7 o *—9 o ®
&
il
o
=
i
-

o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o . L2 3 _ a 8 6 7 8 37 38
o - ANNEALING TIME { hr) :

Fig. 4.1.36. Temperature Changes Durlng‘Anngullfig of Point-Contact Diodes (1N-38-A).

285

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
s i ORNL-LR-DWG 22950
10
<’<,
5 5 "-...\.‘_.
~
e
107¢
o { 2 3 4 5 6 7 8 37 38

TIME (hr)

Fig. 4.1.37. Reverse Current (1-v Bias) Through Ungreased Sample (IN-38-A) as a Function of A'nfieuling Time.

UNCLASSIFIED
ORNL-LR-DWG 22949

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

TIME (hr)
s ' 2 3 4 5 6 7 8 9 10
10 ]
%
o
I\ -
3 "~ |
s ~e -
~ —— L
o« 0o T ° 0-10 HOUR SCALE
D~40 HOUR SCALE
' -
107® ¥
0 5 10 15 20 25 30 35 40
TIME (hr)
Fig. 4.1.38. Reverse Current (1-v Bias) Through Greased Sample (IN 38-A) as o Function of Annealing Time. k?"f

286
 

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL-LR-DWG 2295¢

 

 

2x10°% \

I lamp)

 

/

\(’“"GREASED SAMPLE TT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\
-_——_..
o
\—.___‘ 0 L GREASED SAMPLE ) ® /0
o \’
107°
0 1 2 3 4 5 6 7 8 37 38

ANNEALING TIME (br)

Fig. 4.1.39. Reverse Saturation Current (1-v Bias) of Samples (1N 38-A) as ¢ Function of Annealing Time.

UNCLASSIFIED
ORNL-LR-DWG 22952

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\\__;Q/GREASFD SAMPLE
39  © —
o \\ oV
UNGREASED SAMPLE o
| o
»
2 - -
o t 2 3 4 5 6 7 8 37 38

ANNEALING TIME (hr)

 

Fig. 4.1.40. Reverse Current *'Y’’ Intercept vs Afinecling Time.

287

 
 

 

 

ANP PROJECT PROGRESS REPORT

liquid nitrogen, in order to ‘‘freeze in’’ the damage.
Application of the freezing principle to devices has
not yet been tried, but it could possibly provide
much worthwhile information needed in the solution
of some of the radiation-damage problems. It thus
seems to be worthy of preliminary investigation,
One such preliminary investigation has been
undertaken with fransistors.  Tronsistors were
selected because the information obtained would,
in all probability, have immediate value and because
the collector current, being a function of the emitter
or base current, would present any domage in a
magnified form; that is, any changes noted would
be indicative not only of the damage to the collector-
base diode but also of the damage to the emitter-
base diode. Although not presented in detail here,
the data resulting from this experiment indicate that
damage to devices can be frozen into the sample to
some degree. Whether this is all the damage or
even a significant portion has not yet been de-
termined, since it is possible that some annealing
occurs even at liquid-nitrogen temperature. In this
experiment, two samples were bombarded at liquid-
nitrogen temperature while two other samples were
bombarded at the facility ambient temperature. The
facility temperature was not recorded, but it was
approximately 20 to 25°C, After the irradiation the
cold samples were kept cold, and the warm samples
were allowed to anneal at room temperature.
Approximately 120 hr later the samples kept at
liquid-nitrogen temperature were allowed to warm
up, and the characteristics were observed.
Comparison of the characteristic curves of the
“warm’’ sample immediately after irradiation with
those of the *“cold’’ sample immediately after being
warmed to room temperature shows some difference
which might be attributable to annealing of the
warm sample while being irradiated. Comparison of
characteristic curves after the cold sample was
warmed shows, of course, significant annealing in
the warm sample. Refinement of this type of experi-
ment is indicated in order to determine not only the
amount of annealing after irradiation but to verify
possible annealing during irradiation. For instance,
it may be better to cool the samples bombarded at
room temperafure tfo liquid-nitrogen temperature
after the bombardment for comparison with samples
bombarded at liquid-nitrogen temperature., Longer
exposure would also provide more quantitative
information, since this would permit more annealing
to take place in the samples bombarded at room

288

temperature and would introduce more damage into
both the wam and the cold samples. Further
information from this experiment may be available
when characteristic curves taken on an **X.Y"
recorder are analyzed.

Results of the gamma-ray bombardment of the
coated and uncoated point-contact diocdes indicated
that it would be advisable to determine whether
there was any difference in the effect of radiation
on devices of different but known material con-
ductivity, Samples of point-contact diodes espe-
cially constructed from material of known re-
sistivity were therefore obtained and bombarded in
the Solid State Division gamma-ray source. Several
curves showing the results of the bombardment
have been prepared. 7

The diodes studies were CK-708's made by the
Raytheon Company from crystals of known re-
sistivity. There was no coating material on the
crystals. The four samples selected were 4, 4.1,
11, and 12 ohm-centimeter material, respectively.
Two of the samples were bombarded in such a
manner as to permit data to be taken while they
were in the gamma-ray source. The other two
were irradiated at the same time, but were not
wired for data-taking. 4

In view of the temperature dependence of the
reverse current of the samples, a normalization of
currents with the temperature of the gamma-ray
facility was attempted. An approximate normali-
zation of the current was taken from the curve of
Fig. 4.1.41. This normalization is considered
sufficiently accurate when compared with the
accuracy of the instrumentation. When normadli-
zations of this type are made in the future, more
effort will be expended on the normalization curve.
In order to eliminate temperature effects, the curves
of the results of this experiment were prepared to
show only those points which were taken at the
ambient temperature of the irradiating facility,
that is, about 25°C.

It is immediately noticeable from Fig. 4.1.42 that
the conductivity of the 4 and 12 ohm«cm diodes,
which are the ones from which data were taken
while they were in the source, are similar and
that the conductivities of the 4,1 and 11 ohm-cm
diodes are similar. The observed curmrent is the
product of the current density and the area of the
barrier. Since the area of the barrier depends upon
the contact pressure, as well as on the conditions
of forming, it is possible for diodes made from

(‘ :

 
 

 

 

 

f

PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

—6 UNCLASSIFIED
(X;O ) ORNL-LR-DWG 22957
Q
o .
)
ol
6 /NO. 4
NO. 12—
5 §‘-‘___4.9x10 € amp
"‘\ \
— e —
a \
5
~y
4 [T NO.1
® NO. 4.1
3
INORMAUZED NOS. 11 AND 4.4
2 | - l I
32 . 30 28 26 . 24 ; 22 20

TEMPERATURE (°C})

Fig. 4.1.41. Revérse Cbryent (lev Blus)j"f_hrougl:u CK-TOSV Point-Contact Diode Samples Selected from 4, 4.1,

11, and 12 ohm-ceptimef_er Material as a Function of Temperut_ure._

dlfferent buik materlcxl to hove the same current

under a given set of conditions. Under nrradlatlon,
however, the current densuty changes with changes .
- Thus, .the similarities in’
_?'sample currents are believed to be comcudences.

in- the bufk maferlol

It is- parhculcrly sugmflcunt as - shown in’ Fig.

4.1.42, that the reverse current at one volt of both
_dnodes approached a psuedo plateau that was of the -
same type as observed in other gamma: irradiations-

of semiconductor devices. Such a plateau, or
saturation level, has been observed in gamma

irradiation of bulk germanium,® but for low-
) reSiStivity germanium the decrease in the electron
concentrohon is approxnmately linear out to an
exposure of 5 x 1018 photons/cm , whereas this
‘apparent plateau occurs at an exposure of 1,40 x

> 104 photons/cm?2 for the 4 ohm-cm material diode
“and approximately 7 x 1014 photons/cm? for the

12 ohm-cm rna_terlal_dlode. Further, carrier re-
moval is ‘manifested as a reduction of current,

 

9).'W. Cleland, J. H. Crawford, Jr., and D. K. Holmes,
Phys. Rev. 102, 722 (1956).

289

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

;6 UNCLASSIFIED = -
(;10 ) ¢ HNL-LR-D\VIG'ZEQSG
"|CK=-708~-12 ¢
CK-708-4 ¢
° ® /— 3
/ e
AL
V2E° ~108
AR 40“‘/
} | NO. 4179
-~ \ NO.11 ®
a &,
s pROBABLE CURVE OF NOS, WARDHe — —
: — - )
— — T
4 —
/NORMAL!ZED NOS. 41 AND 4.4
2 )
0 i00 200 300 400 500 600 700 800 2900 3000

IRRADIATION TIME {min)

Fig. 4.1.42. Reverse Current (1-v Bias) Through CK-708 Diodes vs Gamma Irradiation Time.

while in these samples the current increases, It is
evident that other factors are involved. There
would be, of course, a point at which the bulk
effects would begin to show, but higher dosages
then those given to these samples would be re-
quired. The dosage required can be estimated by
referring to previous work done on gamma-ray
bombardment of diodes in which the reverse current
at 1 v continved to rise until a dosage of 1,75 x
107 r was reached and fell thereafter.!0

Two components of the current which composed
the total current through the CK-708 diodes — the
bulk current, represented by the saturation current,
and the surface leakage, represented by the satu-
ration slope -~ are shown in Figs. 4.1.43 aond
4.1.44. In the present model of a diode, these
currents are in parallel, and the sum of the currents
represents the total current through the diode.
Such is not the case, however, and this discrepancy
is part of the reason for the speculation regarding
the need for a new model.

Annealing of the samples is showrin Fig. 4.1.45,
but since there is an instance of a high-resistivity.
sample annealing more rapidly thana low-resistivity

 

10, c. Pigg, Solid State Semiann, Prog. Rep, Feb, 29,
1956, ORNL-2051, p 59.

290

sample and an instance of a low-resistivity sample
annealing more rapidly than o high-resistivity
sample, no conclusion can be drawn from these

- curves, Annealing saturation currents and satu-

ration slopes have not yet been analyzed and are
not presented here. '

The study of the germanium junction as a function
of both temperature and radiation is being hampered
by instrumentation difficulties. The study is to
be as precise as possible, and the sample currents
are to be measured at voltage biases as low as
10 mv. Difficulty hos been encountered with a-c
electrical pickup in the servo-controlled power
supplies and in providing high-impedance inputs to
a Brown recorder, which has a normal impedance of
500 ohms. The servo power supply has been com-
pletely redesigned and now incorporates a number
of advanced features not ordinarily found in re-
cording systems of this type. An input impedance
of 180,000 ohms has been obtained. A block
diagram of the system is shown in Fig. 4.1.46.
The recorder is an advonced model of the recorder
described in ORNL-1199 (ref 11). The more recent

 

", c Pigg, An Automatic Multira}zge Recording De-
vice for Measuring Varying Potentials, ORNL-1199
(April 18, 1952).

 
 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

-6 UNCLASSIFIED
(’;'O } ORNL-LR-DWG 22982
3
I
E2 CK-708-4 o
™ . °
CK-T708-12 .
' o / \ /
had
0
0 100 200 300 400 500 600 700 800

IRRADIATION TIME (min)

Fig. 4.1.43. Bulk Current, Represented by Reverse Saturation Current, Through CK-708 Diodes vs Gomma

Irradiation Time.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

(’é' 0% ORNL-LR-OWG 22083

5

CK-708—4 o

I

—5 ° CK—708—12 A

Lol ~
3
2

0 . 100 200 300 400 500 600 700 800

IRRADIATION TIME (min)

Fig. 4.1,44. Surface Leakage, Represented by Reverse Saturation Slope, of CK-708 Diodes vs Irradiation Time.

291

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

_ UNCLASSIFIED
(“g ) ORNL~LR-DWG 22953
®
A \
\. : -
7 N ~ et
\ v =
/"
NO.4 L "
i \___-.. ’/ et
=
§ 5
~ S\
NN
RN A
a %"’-\
~ -. e
R S .~NO. 4.4 . =T
\ '-_. / -
"'N..,___. . —"_.-______
AV @
3 NO.§
2
o 500 {000 ’ 1500 4800 2000

ANNEALING TIME (min)

Fig. 4.1.45. Reverse Current (1-v Bias) Through CK-708 Diodes vs Annealing Time.

modifications of the servo power supply for the
system are the high-impedance input, the variable-
gain amplifier, the velocity-damping network, and
the variable-phase chopper power supply. Recent
modifications of the Brown recorder servo circuit
are the high-impedance input, the variable-phase
chopper power supply, the positive-current feedback-
damping network, and jitter filter. Other improve-
ments in the system have included replacement of
certain relays to reduce the current through selectro
switches, and replacements of certain relays
which were shown to be incapable of carrying the
necessary currents. The samples for this experi-
ment are IN-91 alloy junction diodes. For com-
parison, a germanium bulk sample is bombarded
with the diodes. The lifetimes and the barrier
heights of the diodes are measured following
irradiations. Changes in the resistance of the bulk
sample provide a check on the neutron dosage to
which the samples are subjected.

Both the dicdes .and the bulk sample have been
bombarded at ~78°C in a facility that has a fast-

292

neutron flux component of 1.66 x 10% neu-
trons/cmZ.sec., After irradiation, the characteristics
curves of the diodes were recorded at -~78°C, and
the diodes are to be warmed to 0°C. Sample
characteristics will be recorded while the samples
are being warmed. Bombardments up to and in-
cluding a dosage of 10! neutrons/cm? have caused
little change in the diode samples. After a bom-
bardment to a dosage of 102 neutrons/cm?, a
flattening of the forward curve at —78°C has been
observed. The reverse current is not observable
at this temperature. The samples are being held at
~78°C pending replacement of faulty relays so
that characteristics may be recorded during the
wormup period.

Experimental Refinements

It has been noted that the changes observed in
the irradiations performed both in this program ond
in other programs cre, for the most part, opposite
to those expected from the results of bulk or sur-
face studies. Therefore efforts have been made to

 
 

 

 

B

 

 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
ORNL~LR—-DWG 22954

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

_ SERVO HIGH IMPEDANCE |
\ AMPLIFIER INPUT
SAMPLE - '
REFERENCE l
VOLTAGE
SWITCH |
Aflfiamfiyc HIGH IMPEDANCE - BROWN
SHUNT INPUT RECORDER

 

 

 

 

 

 

THERMOCOUPLE

 

 

 

 

COLD JUNCTION

Fig. 4.1.46. Block Diagrom of Multipoint Recorder.

compile quantitative information concerning the
behavior of barriers or rectlfyang devices in the |
presence of radiation fields so that, by comparison
with the results of bulk and surface studies, those
chcrccter:shcs unique to the barrier can be defined.
Further refinements of these studies are being
made. Several samples of barriers of known charac-
teristics have been obtained and will be studied.
The equipment necessary for fhese stud|es is bemg ,
designed, built, or purchcsed '
‘A probe mechanism is bemg constructed that
consists basically ofa fine screw-driven base which
will carry a scmple under a probe for determination
of its resistivity as a function of sample displace- .
ment and to determine location and ‘change of the
junction in a sample. '
As mentioned before, the surface conditions or
surrounding environment of a device can play a

more significant part in the behavior of the device
than would be expected from the results of bulk or

- surface studies. In order to eliminate or at least

minimize surface effects it has been decided to

" prepare  samples for irradiation in a stondard

atmosphere. A vacuum has been proposed and may
be tried, but, for the present,.helium will be used,

- since it will provide a heat transfer medium in the

event gamma-ray heating becomes a problem, A
dry box has been obtained, and ‘it will be meodified

‘to provide a helium atmosphere, To further assist
_in sample preparation, a small chemical hood has

been designed and is now being built. This will
permit the safe use of etches in the preparation of
the samples and in changing the surfaces after the
samples have been bombarded. :
In order to increase the range of voltages which
can be applied to a sample and thereby obtain

293

 
 

 

 

 

information about break-through characteristics
and to enhance the accuracy of saturation slope

data, an *“X-Y'’ recorder with increased ranges is

being obtained. A sample power supply that will

“be compatible with the characteristics of the

recorder is to be designed. The recorder can
measure applied voltages up to 100 v on either
scale, and thus the range is beyond the maximum

voltage at which most diodes operate. With proper
. shunts it will be possible to extend considerably

the range and the types of units which can be
studied.

Another multipoint recorder has become available
with which to constantly follow changes in a sample

294

over a period of days. A controlled-voltage power
supply is also available, and it is thought that
with moderate modifications it will be suitable for
use with the recorder. A recorder of this type willi
do much to clarify such changes as the initial
drop in current of the 12 ohm-cm dicde, as shown
in Fig. 4.1.42, Clarification of such changes is
necessary for a conclusive understanding of the
mechanism of changes in semiconductor devices,

The design of the fast-flux irradiation facility
discussed previously is now being detailed. No
date for completion of the design has been set,
but the werk is proceeding satisfactorily.

n
 

 

 

Part 5
REACTOR SHIELDING

E. P. Blizard

 
 

 

 

 
 

 

5.1. LID TANK SHIELDING FACILITY
W. Zobel

RADIATION ATTENUATION MEASUREMENTS
IN PLAIN WATER, BORATED WATER,
AND OIL

D. W. Cady! E. A. Warman?

A comparison of the attenuation of fast neutrons,
thermal neutrons, ond gomma rays in various
liquids has been made through the use of the Lid
Tank Shielding Facility {LTSF) at ORNL. The
liquids investigated were - plain water, borated
water, and transformer oil. The borated water
used in these tests contained 1.34 wt % natural
boron and had a density of 1.05 g/cm3. The
transformer oil was composed of 86.7 wt % carbon
and 12.7 wt % hydrogen cnd had a denslty of
0.87 g/ecm® at 20°C.

In order to avoid mlxmg of the borated wuter or
oil with the plain water in the lid tank, these
materials were contained in the usuval *‘configura-
tion'’ tank which was positioned as close to the
source as was physically possible. The con-
figuration tank is constructed of mild steel and
has dimensions of 654 x 72 x 7]'/ in. The pro-
duction of capture gamma rays in the tenk wall
near the source was reduced by the insertion of a
%-in.-thick aluminum window in the side of the
tank adjacent to the source. A 2-cm gap between
the source and the window was filled with air
contained in a thin plastic bag.

Thermal-neutron, fast-neutron, ond gamma-ray
measurements taken in the three media contained
in the configuration tank are shown in Figs. 5.1.1,
5.1.2, and 5.1.3, respectively. Before the data for
plain water can be. compared with previously
published measurements,3:4 which were taken in
the lid tank rather than the configuration tank, the
new curves must be shifted 3 cm to the left. The

 

On usmgnment from ‘Wright Alf Development Center,
Dayton, Ohio. .

20n osslgnment from Pratt & Whitney Alrcrah, East
Hartford, Connecticut.

3D. R. Otis and J. R. Smolen, Prelimina LTSF Data
Report on Experiments with Advanced Shielding Ma-
terials (LiH, Zr, Be, Stainless Steel, W, Hevzmet, Pb,
and v (Expt 69), ORNL CF-57-2-8 (Feb . 1957).

4R, W. Peelle et al., ANP Quar. Prog. Rep. Dec. 31,
1956, ORNL-2221 (Part 1-5), p 332, Fig. 5.1.1.

agreement between a thermal-neutron traverse in
the lid tank and a traverse in the configuration
tank, with the latter data shifted on the graph a
distance of 3 cm to the left, is illustrated in
Fig. 5.1.4.

The previously reported curves for thermal-
neutron fluxes in oil and inborated water represent
data taken with the instrument response normal-
ized to gold-foil fluxes in plain water. This
normalizing procedure is not particularly valid,
as is indicated by a comparison of counts-to-flux
conversion factors used over a period of time.
For example, the current conversion factor used
with a ]2'/2-in. BF, counter in plain water, ob-
tained by normalizing to the data taken with gold
foils in plain water, is 1.66 x 10=3. Previously,
this same factor would have been used for this
counter for all media. The newly determined con-
version factor for borated water is 9.78 x 10=4, a
change of about 41%. Similarly, the conversion

UNCLASSIFIED
2-01-057-0-383%

%Y 0%

    

- -t - - -
‘thoN' ofl Uob U% GC’O

o

' THERMAL-NEUTRON FLUX (neutrons/cmE-sec-wott)
o o>

O 20 40 60 80 100 120 140 160
2, DISTANCE FROM SOURCE PLATE {cm)

Fig. 5.1.1. Thermal-Neutron Fluxes in Water, Borated
Water, and Oil in the Configuration Tank.

297 j/ 297

 
 

 

ANP PROJECT PROGRESS REPORT

UNGLASSIFIED
2-01-057-0-366

10*

ad uqa uq

FAST-NEUTRON TiSSUE DOSE RATE (ergs/g-hr-watt)
-

 

0 20 40 60 B8O 100 120 a0 160
Z,, DISTANCE FROM SOURGE PLATE (cm)

Fig. 5.1.2. Fast-Neutron Tissue Dose Rates in Woter,
Borated Water, and Oil in the Configuration Tank.

UNCLASSIFED
2+01-057-0-367

GAMMA-RAY TISSUE DOSE RATE (ergs/q-hr-wolt)

 

o 20 40 60 80 100 120 140 160
Z,, DISTANCE FROM SOURCE PLATE {cm)

Fig. 5.1.3. Gamma-Ray Tissue Dose Rates in Water,
Borated Water, and Oil in the Configuration Tank.

factors for the 8-in. BF, counter, the 3-in. fission
chamber, and the I;g-in. fission chamber have been
reduced by 26, 37, and 16%, respectively. Cor-
responding changes were made in the conversion
factors for oil.

The data presented here are opproximately 5%
higher than the earlier data because of the re-
placement of the old !fi-in.-thick boral shutter on

298

UNCLASSIFIED
2-01-057-0—-368

           

0 20 40 60 80 100 120 140 160
#o. DISTANCE FROM SOURCE PLATE {cm)

Fige 5.1.4. Thermal-Neutron Fluxes in Water in the
Lid Tank and in the Configuration Tank.

the source plate by a I4'-in.-thick shutter. Lid
tank data are obtained by determining the dif-
ference between the shutter-open aond shutter-
closed measurements. The old Y%-in. shutter did
not completely shield the source plate from the
reactor thermal flux; therefore, even with the
shutter closed, there was fissioning in the source
plate. When the shutter-closed readings are sub-
tracted from the shutter-open readings, a small
percentage of the source-plate power is eliminated
as background. Thus, an increase in the shutter
thickness essentially increased the effective
source-plate power. This increase is small enough
to fall within the uncertainty in source-plate
power (15%), and therefore a more exact deter-
mination of the new effective source-plate power
will be made.

Thermal-neutron measurements in wdter are-now
being corrected for the flux depression that re-
sults from the presence of the gold foils and for
self-cbsorption and *‘self-shielding®’. in the foils.
 

 

 

These corrections were not made on previously
reported curves.3%

Although it is commonly assumed that the flux
depression for *‘thin'’ foils is negligible, calcula-
tion of the flux depression caused by 2-mil-thick,
1-em2 gold foils in water, oil, and borated water
were made. Several methods of calculation were
attempted, 78 and each produced a different re-

sult, The figures reported here are based ona

method suggested by W. R. Burrus.® The prelimi-
nary values obtained were 8.5% for water, 5.9%
for oil, and 5.0% for borated water. Little flux
depression is normally expected in borated water,
but the amount of boration in the LTSF water is

~so slight, 1,34 wt %, that the 5.0% value is

reasonable. it should be noted here that cadmium-
difference flux measurements in borated water are
of dubious validity. The cadmium ratio is so
small, about 1.5:1, that the errors inherent in
taking the difference between two large, neorly
equal, values are considerable.

The self-absorption correction is a correction to
the measured activity of the foil after exposure.
Since the foil cannot be infinitely thin, there will

be scattering and absorption of the radiation in:
the activated foil itself. Hence the true activity
of the foil, or counting rate, will be reduced by
some factor which is dependent on the foil thick-

ness. The dependence of the self-absorption on

the foil thickness can be calculated roughly by

assuming an exponential absorption and an ab-

" sorption coefficient _independent of the depth in
which the particles are em:fled Based on the
assumption that one-half the thickness of the foil

contributed to the self-absorpnon, the calculat:on

yields a self-absorptlon factor for a 2—m|l-thnck :

gold foil of approx:mately 1%.

 

5W VZOBei“et ali ANP Qufir Prog Rep ]une 30 S

1957 ORNL-2340 (Part 1-5), p 285

' T. Chapman et al., ANP Quar Prog. Rep ]une -
: 10 1955, 0RNL-]896, P 194 : _

’p, 4. Hughes, * Pile Neutron Researcb Adduon-

- Wesley, Cambridge, 1953.

8W R.- Burrus, USAF, Wright Air Developrnent Center,
Dayton, Ohio, pnvate communication.

R, Aronson et als, Penetration of Neutrons

Point Isotropic Fission Source in Water, NY 6267

(Sept. 22, 1954).

" THERMAL~NEUTRON FLUX {neutrons/cm?- sec -watt)

a -

if a foil is quite thick, enough neutrons will be
absorbed so that the nuclei in the center of the
foil will **see’’ a lower thermal-neutron flux than

can be seen by those at the surface. if the ratio
of the absorption to the scattering for the foil is
large, the ‘‘self-protection,’’ or *‘self-shielding,"
can be calculated because the incident flux suf-
fers an exponential decrease in the foil.” For a
2-mil-thick gold foil this caleulation yields a
‘‘self-protection’’ factor of 1.5%.

The effect on the configuration tank water curve

of considering the 11% correction for flux depres-

sion, self-absorption, and self-shielding is shown
in Fig. 5.1.5. Both curves shown in Fig. 5.1.5
already contain the 5% correction for the change
in thickness of the boral shutter. Therefore, the

UNCLASSIFIED
2-01-057-0-369

- 08 0% 0l 08 0l 0 od, o %0

oL

5,
N -n

 

- -
10
0O 20 40 € B8O 100 120 440 160
Z, DISTANCE FROM SOURCE PLATE {cm)

. Fig. 5.1, 5. Thermal-Neutron Fluxes In Water in the

Conflgurcflon Tank: Compurlson of Uncorrected Meos-
urements wlfh Those Corrected for Flux Depression,
Self—Absorphon, and Self-ShieIding.

299

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

difference between the two curves is due only to
this 11% correction.

There have been efforts to predict theorehcally
the attenuation of gamma rays and neutrons in
water.?=11  The data presented here will be of
value to any future work on this problem. It
should be noted that nc attempt has been made to
predict attenuation in the oil or the borated water.

STUDY OF ADYANCED SHIELDING MATERIALS
E. A. Warman?2

The thermol-neutron data that were omitted from
the previous report® on the group of tests per-
formed to aid the ANP shield design effort of
Pratt & Whitney Aircraft are reported here. These
data’ present detector response measurements
normalized to cadmium-difference gold-foil fluxes.
The gold-foil fluxes have been corrected for flux
depression by the 2-mil-thick foils and for self-
absorption and self-shielding in the foils, as
described in the preceding paper.

The configurations tested included Hevimet,
stainless steel, boral, and lithium hydride, pre-
ceded by a beryllium reflector-moderator and backed
by borated water to mock up borated alkylbenzene.
The materials and their experimental arrangements
were described previously;5 however, subsequent
chemical analysis of the borated water has shown
that the boron content of the water was 1.34 wt %
rather than the 1.46 wt % reported previously.

A comparison of the thermal-neutron measure-
ments taken in water and in borated water (refer to
Fig. 5.1.1 in preceding section) showed that the
addition of the boron was quite effective in re-
ducing the thermal-neutron flux. The borated
water curve had a slightly steeper slope than that
of the water curve. The thermal-neutron flux in
the borated water was a factor of 29 below that
in plain water 15 cm from the source plate and a
factor of 36 below the flux in water at 120 em. It
should also be noted that the hump in the curve in
the forward region was not as pronounced in
borated water because of the greater percentage of

 

Ve P, Blizard and T. A, Welton, The Shieldin e of
Mobile Reactors, ORNL-1133 (Part 2) (June 30, 195

”D R. Otis, Neutron and Gamma-Ray Attenuation for
a Fission Source in Water — Comparison of Theory with
LTSF Measurements, ORNL CF-57-3-48 (March 12,
1957); ANP OQuar. Prog Rep. Dec. 31, 1956, ORNL-
222% (Part 1=5), p 331,

300

absorption of the neutrons being thermalized in
the first few centimeters of water.

A short discussion of the measurements made in
the borated water beyond the various configura-
tions is presented below, although a complete
analysis has not yet been made. N

Configurations 25, 26, and 27. ~ Thermal-
neutron flux measurements beyond conflguruhon 25
(Fig. 5.1.6) show the effect of replacing 4/ in, of
borated water with ‘/ in. of Li-Mg alloy plus 4 in.
of beryllium. The fhermul-neutron flux was raised,
as was expected, because of the addition of the
beryllium.

Conflgurahon 26 consisted of 4 in. of Hevimet
preceded by / in. of Li-Mg and 4 in. of beryllium,
The slope of a plot of the thermal-neutron flux
curve for this configuration is considerably altered
from that for configuration 25. - This is possibly
due to the Hevimet not being os effective as
borated water as a thermal-neutron shield, and, as
a result, there is a higher thermal-neutron flux at

VR
2-01-057-70-376

7
10
5
2
10°
5 ONLY (134 wi %
2
10° .
5 ' .
4 in.OF Be +4 in.OF HEVIMET +
2
10t TION 27: ¥4 in OF Li-Mg
5 +4 inOF Be+% in.OF +4 inOF
HEVIMET + B-H,0
2

S
W

%
Li-Mg ALLOY 4+ 4 in. OF Be
+B-

ONN\ 18]

THERMAL-NEUTRON FLUX {neutrons/cm?.sec- watt)
N =N O N o

<

N o

 

<3I
nN

0 {0 20 30 40 50 60 VO B8O 90 {00 {10 120 430
Zo, DISTANGE FROM SOURGE {cm)

Fig. 5.1.6. Thermal-Neutron Fluxes Beyond Con-
figurations 25, 26, and 27, (Secrety

 
 

 

 

short distances beyond the Hevimet. However,
the attenuation of fast neutrons in the Hevimet
resulted in a reduction of the thermal-neutron flux
at greater distances. '

In configuration 27 a lé-in.-thicl«: boral curtain
was placed between the beryllium and the Hevimet.
The thermal-neutron flux was somewhat lower than
that beyond configuration 26 for short distances
beyond the Hevimet but was raised so that curves
of the data for the two configurations approxi-
mately coincided at greater distances. This pos-
sibly indicates that the effect of the boral be-
comes negligible at these distances.

Configurations 28, 29, 30, and 31. — The thermal-
neutron flux measurements beyond configuration 28
(Fig. 5.1.7) show the effect of adding 12 in. of
lithium hydride behind configuration 27. The flux
was lowered by about a factor of 2 immediately
following the lithium hydride and approximately a
factor of 2.7 at a distance of 100 cm from the
source plate. This indicates a softer neutron
spectrum. The thermal-neutron measurements
beyond a configuration having an additional 12-in.

 

SEenTT™
10°
5 : Yy in, OF Li-Mg ALLOY +
o 4in. OF Be + Y2 in. OF + 4in, OF
HEVIMET + B~
10*
’g 5
g 2
% 103
o
§ 5
v
5 2
3 10°
o
= 5
5
J e
= 10
g s
& CONFIGURATION 28: 4 in.OF
@ 2] ALLOY +4in.OF Be + Y2 in. OF .
T | BORAL + 4in. OF HEVIMET +
3 12in. OF LiH + B-Hp0
= 5
&
= A
o CONFIGURATION 29: Y, in. OF Li-Mg
S ALLOY + 4in.OF Be +¥,in. OF BORAL
+4in. OF HEVIMET +24in. OF LiH + B~H,0
2 -
2

30 40 50 60 70 8C 90 {00 110 120 130
© z,, DISTANCE FROM SOURCE (cm)

o 10

'Fig.r 5.71.77.' Thermal-Neutron Fluxes Beyond Con-
figurations 27, 28, and 29. {Secred)

PERIOD ENDING SEPTEMBER 30, 1957

thickness of lithium hydride (configuration 29)
appear to be inconclusive owing to the decreased
thermal-neutron flux and to instrumentation dif-
ficulties; no thermal-neutron data were obtainable
for configurations 30 and 31.

Configurations 32, 33, and 34. — The effect of
placing a l-in.-thick stainless steel pressure
shell beyond configuration 27 was studied in
configuration 32, The change in the slope of the
thermal-neutron flux curve may be partially ex-
plained by the fact that steel is less effective as
a thermal-neutron shield than is borated water.
Also, the iron in the steel has a dip in its cross-
section curve for 23-kev neutrons, which allows
a large number of neutrons of this energy to escape
into the borated water where they are subsequently
thermalized. These effects would cause an in-
crease in the thermal-neutron flux in the forward
region of the curve. On the other hand, scattering
of fast neutrons in the steel increases the prob-
ability for their absorption in the region further
from the source and causes the curve to go below
that of configuration 27 in this region.

In configuration 33 the effect of increasing the
thickness of the beryilium reflector-moderator used
in configuration 32 was studied. The addition of
4 in. of beryllium resulted in an increase in the
thermal-neutron flux near the shield surface be-
cause of the large percentage of thermalization of
neutrons in the beryllium. The flux dropped in the
region further away from the source because of
the capture of these same neutrons by the boron
and hydrogen.

Configuration 34 was an attempt to determine the
effect of reducing the thickness of the Hevimet
gaomma-ray shield used in configuration 32. Sub-
tracting 2 in. of Hevimet allowed on additional

2 in. of borated water to be added to the back of

the shield, which resulted in a decrease of the
thermal-neutron flux at short distances out in the
borated water. The flux increased, however, at
greater distances and was slightly above that of
configuration 32. The results for configurations
32, 33, and 34 are compared with those for con-
figuration 27 in Fig. 5.1.8.

301

 
 

 

ANP PROJECT PROGRESS REPORT

302

SECRER
2-01-057-70-378

[¢2]

10
5 ATION 34: Y%4in. OF Li-Mg ALLOY +
4in.OF Be + Y% in.OF BORAL + 2in. OF HEVIMET
% +in. OF STAINLESS STEEL + B-H,0
10
CONFIGURATION 27: Y, in.OF Li-Mg ALLOY
S +4in. OF Be + Yin. OF BORAL + 4in. OF
= 2 HEVIMET + B-
S . 4
210
O
8 9 CONFIGURATION 32: Y in. OF Li-Mg
o ALLOY + 4in.OF Be +% in. OF BORAL
S 3 +4in.OF HEVIMET +1in. OF
¢ 10 STAINLESS STEEL + B-H,0
o =
..:_’ ~
-
2
% 10
2 e
. J
=
e 2
5 10
Lt
< 5
g
g 2
oo
L
E 5
CONFIGURATION 33: ¥, in. OF Li-Mg ALLOY
2 + 8in. OF Be + Y%, in. OF BORAL + 4in.OF
10! = HEVIMET + tin. OF STAINLESS STEEL + B-H,0
5
2
1072

 

0O 10 20 30 40 50 60 70 80 90 {00 HO {20 130
2y, DISTANCE FROM SOURCGE (cm)

Fig. 5.1.8. Thermal-Neutron Fluxes Beyond Configurations 27, 32, 33, ond 34.
 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

5.2. BULK SHIELDING FACILITY
F. C. Maienschein

THE SUPPRESSION OF CAPTURE GAMMA RAYS
BY LITHIUM IN LEAD

J. D. Kington

One of the problems that confront reactor shield
designers is that of shielding against secondary
gamma rays produced by the capture of thermal
neutrons in lead. A search of the litercture re-
vealed that a process for fabricating a lithium-
lead alloy was patented in Germany in the 1930’s

and that the eutectic contained 0.67 wt % lithium.
A lead sample containing 0.67 wt % lithium was
then fabricated by the Metallurgy Division, and a
preliminary experiment at the Bulk Shielding
Facility was performed in order to investigate the
effectiveness of the lithium in reducing the cap-
ture gamma-ray production in the lead. It was
found that lithium distributed throughout lead was
approximately as effective as boral covers on a
pure lead shield.

303

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

5.3. TOWER SHIELDING REACTOR-Il

C. E. Clifford

MECHANICAL DESIGN'

A method of suspension for the TSR-Il that will
allow a beam of radiation from a collimator through
its shield to sweep both a vertical plane and o
horizontal plane has been chosen and the design
is being examined from structural and stability
standpoints. In order to permit rotation of the
beam in the vertical direction it was necessary to
reduce the length of the reactor tank from that
used in previous designs. The relative dimensions
of the TSR-Hl, as now designed, are shown in

Fig. 5.3.1.

NUCLEAR CALCULATIONS?

As was reported previously,3 a nuclear parameter
study was carried out with the 3G3R code on the
Oracle to establish the dimensions of the TSR-II.
The resulting configuration had a 17.5-in.-dia
internal water reflector, a 5.5-in.-thick core region,
a 7.875-in.-thick external water region, and an
aluminum-to-water volume ratio of 0.57. Similar
calculations have now been performed on the
UNIVAC with the Murine code. A comparison of
the values of concentration vs reactivity from the
two calculations revealed that @ change of 0.3%
in the thermal-neutron macroscopic absorption
cross section, % o for the Oracle calculations was
all that was required to obtain complete agreement.
Results of the Oracle calculations with the revised
value of X are compared with the UNIVAC cal-
culations in Fig. 5.3.2. The agreement is such
that. control problems and the effects of various
shells in the design are to be investigated almost
entirely with the Oracle 3G3R code, and only the
final results will be checked by a repeat calcula-
tion with another code.

Calculations have also been performed to de-
termine the total amount of reactivity that could be
controlled with a boral shell in an internal water
reflector. An ideal case was used with the dimen-
sions given above. A boral shell was placed at

 

Mhe en ineering design of the TSR-1l is being per-
formed by the ORNL Engineering Department.

2The nuclear calculations for the TSR-Il are being
performed by M. E. LaVerne.

3¢. E. Clifford and L. B. Holland, ANP Quar. Prog.
Rep. March 31, 1957, ORNL-2274 (Part 1=5), p 294.

304

L. B. Holland

the core-internal reflector interface and the trans-
mission of thermal neutrons through the shell was
varied from 0 to 100%. With zero transmission of
thermal neutrons the reactivity that could be con-
trolled was found to be 11%. For these calcula-
tions the transmission of fast neutrons was as-
sumed to be 100%, and thus the result should be
conservative, The change in the thermal-neutron
flux produced by the insertion of the zero trans-
mission shell is shown in Fig. 5.3.3.

FUEL ELEMENT DEVELOPMENT?

The fuel element development work is near com-
pletion. Of the 71 different fuel plate sizes re-
quired for the two types of elements, 60 plates
with depleted fuel have been rolled to the correct
shape. It is necessary to determine the proper
rolling sequences to produce plates with the proper
thickness and with length and width both held
closer than the usual MTR tolerances. Delivery
of a three-cylinder hand roller sufficiently long to
accommodate the longer fuel plates of the annular
elements has made it possible to fabricate several
dummy annular elements. Inspection of these
elements shows them to be well within the toler-
ance necessary for operation in the TSR-1l. One
of the annular elements is shown in Fig. 5.3.4.
Two views of a quarter section of the reactor core
formed with the dummy aluminum elements are

shown in Fig. 5.3.5.

DEVELOPMENT OF THE CONTROL DEVICE®

Improvements were made in the TSR-Il prototype
control mechanism and a new method was employed
to measure the distance traveled by the control
plate under scram conditions as a function of time
after scram. The design changes were of a minor
nature but tended to reduce friction in the posi-
tioning device and guiding system. The effect of
the changes was to bring the operating pressures
more in line with the calculated values. As

 

4The fuel element development work is being per-

formed by the ORNL. Metallurgy Division.

5The control system for the TSR-1l is being developed
by the Reactor Controls Group.

(‘ |
 

14

 

" CONTROL MECHANISM

FISSION CHAMBER
POSITIONING MOTCR

CONTROL MECHANISM
POSITIONING DRIVE MOTOR

 

HANDLING LUGS

FISSION CHAMBER ——— ]

 

 

.:i
CONTROL MECHANISM

POSITIONING DEVICE

A

ALUMINUM

BORAL

Fig. 5.3.1. Proposed Tower Shielding Reactor

T

 

 

1]

 

|

q

 

 

L)

ok Fo
b
i

 

 

   

 

 

 

 

 

 

 

 

 

i

 

 

I

K3

 

 

    

 

 

 

 

 

‘ =1

 

 

PERIOD ENDING SEPTEMBER 30, 1957

 

 

 

    
 

UNCLASSIFIED
2-04-060-1-R2
WATER INLET
| WATER OUTLET
[
|
]
| ¥
D]
o

 

R R
:.\\:s:a"\\.\
Ry

—=——REACTOR TANK

H—— IONIZATION CHAMBER

 

 

INTERNAL REFLEGTOR

 

CORE REGION

LEAD

I1 {(Vertical Section).

305

 
 

 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

19 - UNCLASSIFIED
{(x 10" 2—04—~060—27

'2 T T

e UNIVAC CALCULATIONS
@ — ® ORACLE CALCULATIONS

 

[ ————

 

~§

 

 

P
e

 

 

 

 

 

6 /

REACTOR GEOMETRY: 0.57

i " ALUMINUM-TO-WATER VOLUME

5 - RATIO, 445-cm~dia INTERNAL  —|

- WATER _REF LECTOR, _lS-cm-THICK

CORE, '25-cm-THICK EXTERNAL
WATER REFLECTOR

. /

06 [o)rg 08 09 10 14 12 3

REACTIVITY, &

 

 

U23® CONCENTRATION (atoms /cc)

 

 

 

 

 

 

 

 

 

 

 

Fig. 5.3.2. Concentration of U235 in the Proposed
TSR-1l as a Function of Reactivity. Reactor geometry:
0.57 aluminum-to-water volume ratio, 44.5cm-dia
internal * water reflector, 15-cm-thick core, 25-cm-thick

external water reflector.

previously reported,® a 66-psi line pressure should
be sufficient to hold the poppet against the posi-
tioning device for normal operation; reducing the
pressure below 50 psi should allow the spring
force to overcome the pressure in the cylinder, and
the mechanism would move away from the posi-
tioning device. (In the reactor the control plates
would then move toward the fuel.) The actual
pressures observed with the design improvements
were 72 psi to operate the mechanism satisfactorily
ond less than 53 psi to permit the spring to over-
come the hydraulic force.

 

8¢, E. Clifford and L. B, Holland, ANP Quar. Prog.
Rep. June 30, 1957, ORNL-2340 (Part 1-5), p 323.

306

 

UNCLASSIFIED
- 2=-04~—060—25

BN | INTERNAL CORE _+exTERNAL__{- |
B ™ REFLECTOR ] REGION | REFLECTOR |-

S

.........

RADIUS {(cm)
(@) No Boral Shell.

' INTERNAL ’ CORE , EXTERNAL_‘_i
: REFLECTOR REGION . REFLECTOR :

   

.ave.

.....

.........

..................

RADIUS (cm)

(6) Thermolly Black Shell at Interface of
Core and Internal Reflector.

Fig. 5.3.3. Effect of a Boral Shell at the Core ~Internal
Reflector Interface on the Neutron Flux in the Proposed
TSR.Il.

 
 

3

 

PERIOD ENDING SEPTEMBER 30, 1957

8 UNCLASSIFIED
57

¥

Fig. 5.3.4: TSR-11 Dummy Annular Fuel Element.

The control-plate travel as a function of time
under scram conditions was obtained by observing,
with an oscilloscope which has a persistent screen,
the time required for the piston and control plate
to traverse a known part of its travel. The time

~was defined by having the control. p!ate break an

electrical ‘contact when motion started and then
make electrical contact ‘when the desired. distance
had been _covered.
peared as pips on the sweep sngnul moving across
the’ oscnlloscope screen. The. values of distonce

traveled as a function of time thus obtained are
- shown in Fig. 5.3.6, fogether with the same values
obtained - by an’ iterative solution “of a force-
‘balance equafion that was written to design the
‘prototype. A full-scale model is being desngned

on the same prmmples.

' The change in contact ap-

CALCULATION OF THE GAMMA-RAY
HEATING IN THE LEAD SHIELD

W. W. Dunn’ D. K. Trubey

Calculations have been carried out in order to

- predict the gamma-ray heating in the lead region

around the TSR-ll core. For, these calculations,
the core was assumed to contain 9 kg of U233 and

to have an aluminum-to-water core volume ratio of

0.57. The caleulations involved two steps: (1) o

“caleulation of the heating in the core at the core-

shield interface, and (2) a Monte Carlo Oracle
calculation of the heating in_arbitrarily chosen
regions {as shown in Fig. 5.3.7) throughout the
lead shleld The calculated heating in the core at

 

70n assignment from USAF,

307

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Rl UNCLASSIFIED|
¥-23553

 

 

() INSIDE VIEW [ERE.

 

N JHCLASSIFIED
¥-23548

o L3
itlt”lzll'llll )

(5) EXTERNAL VIEW

Fig. 5.3.5. Dummy Fuel Elements for the Proposed TSR-I1, Assembled in Quarter Sphere.

the core-shield interface, coupled with an as-
sumption for the angular distribution at the inter-
face, was used as the source term for the Monte
Carlo calculation. The two steps are described in
detail below.

Heating at the Core-Shield Interface

For the calculation of the core heating at the
core-shield interface, the core was divided into
seven spherical shells (see Fig. 5.3.7) chosen so
that a reasonable average flux could be assumed
for each shell. The source distribution was as-
sumed to be proportional to the thermal-neutron
flux as calculated by the Oracle 3G3R code for an
aluminum-to-water volume ratio of 0.707 and core
dimensions of 20 and 40 cm. This flux plot was
scaled to the assumed core dimensions (22.225-cm
inner radius and 37.46-cm outer radius) and was

308

vused for the calculation, although the aluminum-
to-water volume ratio in the calculation was
specified as 0.57. The gomma-ray spectrum from
each source shell was then divided into six energy
groups, which were arbitrarily chosen as 0.5, 1, 2,

4, 6, and 8 Mev.

According to the usual transformation from a
spherical-shell source to an equivalent infinite
plane source (neglecting the back plane),® the
core heating at the core-shield interface as a
function of the initial energy E, leaving the source

 

8This transformation has been described by E. P.
Blizard; see, for example, S. Glasstone, Principles of
Nuclear Reactor Engineering, p 592, Van Nostrand,
New York, 1955. In summary, H(shell) = (r/ro) X
[H(infinite plane ot rg = 1) ~ H(infinite plane at rg + Il

o

 
 

 

 

 

9

UNCLASSIFIED
2—01—-060-30

0.9 //—"//’

0.8 /
0.7 //

0.6

CALCULATIONS ~ / ]/

0.5

/

[

 

 

 

 

 

 

 

DISPLACEMENT (in.}

~~ LABORATORY DATA

 

e /
i/

 

0.2

 

/
s
 ~

 

 

 

 

 

 

 

 

 

 

0 5 t0 5 20 25 30 35 40
TIME {(msec)

Fig. 5.3.6. Piston Displacement in Control Mechanism
Test for Proposed TSR-Il as a Function of Tirne.

plane was determined by the equation

' Sr
- H(shell,Eo) := -".; [21TKS(E0) Eqy Fa(EO)] X

x [ B(REq) GIR,Eg) R dR 3

where

rq = distance - from center - of core (or
center of internal water reflecfor) to
. core-shield interface, cm, :
r = distance from center of core to
. spherical source shell cm, '
.z = o — r,cm, S

_S(Eo) - .
- gamma rays of energy E, generated
 per square cennmeter per: hssron'
: | ‘per second, - |
- (Ey) = energy ubsorphon | coefficient» for

gamma rays of energy E, in the
core,

source term for each sheH number of,

PERIOD ENDING SEPTEMBER 30, 1957

"R = distance from the source point on
the spherical shell to the core-shield
interface, cm,

B(R,E,) = energy absorption buildup factor in
the core as a function of distance R
and source energy E,,
G(R,Ey) = oftenuation kernel for a point iso-
fropic source
= e-#'R/47TR2,

#, = total absorption coefficient,

K = term to convert to watts of heating
per cubic centimeter of material per
megawatt of reactor power.

The correction for the heating in the 7, + r infinite

plane, that is, the back plane, was made in the

actual calculation, but it amounted to only a small
per cent and then only in the innermost core source
planes.

The source term, S(E;), which was ussumed to
be proportional to the average thermal-neutron flux
in that shell, can be given by the expression

@ S(Eg) = Ayt [N + o) u) +

+ Nfp) ZA2) + N(AI) Z(A)) +
+ NH) X H)] ,

where

A = factor for normalization to one fis-
. sion per second in the core,
$,, = average thermal-neutron flux in the
source shell,
t = shell thickness, cm,
N(p + c) = number of prompt fission and capture
gamma rays inthe energy group about
E, produced in uranium per fission,

2/(") = macroscopic fission cross section of
U235, _
number of fission-product decay
gamma rays in the energy group
‘about E, per fission,
N(Al) = number of aluminum capture and de-
' cay gamma rays in the energy group
about Eo per capture,
2 (A1) = macroscopic absorption cross sec-
' tion of aluminum, :

N(fp)

309

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

  

BORAL

INTERNAL WATER REFLECTOR

315‘“\-

 

UNCLASSIFIED
2—01—-060-28

123 4 5 67135 iom._v o2

. 46 11
NOT TO SCALE IN " Y v

. ‘ 1 ‘ _
/a-in. Jq—in. Yo-in.
CORE REGION SPACES  SPACES SPACES

Fig. 5.3.7. Shield Configurction Used to Calculate Gamma-Ray Heating in the TSR-Il.

N(H) = number of hydrogen capture gamma
rays in the energy group about E,
per capture,

Z (H) = macroscopic absorption cross sec-
tion of hydrogen.

The values used for the N's as a function of E,
are listed in Table 5.3.1.

310

 The linear and energy absorption coefficients of
the core as a function of E, were determined by
the following expression, where p, is the fractional
density of the ith material, '

K
Ficore) = E(_) Pi -
i \P /i

 

o

 
 

"

 

Toble 5.3.1. Equivalent Number of Gamma Rays In the Core for Each Energy Group E,

PERIOD ENDING SEPTEMBER 30, 1957

 

 

(i?') N(p + c)/fission? N(fp)/fission® N(H)/captureb N(Al)/capfureb
0.5 3.31 2.972 0 0
1 2.55 1.864 0 0.10
2 1.687 0.89 1L.115 1.53
4 0.368 0.11465 0 0.77
6 0.050 0.0084 o 0.21
8 0.0067 0.0006 0 0.35

 

“H. W. Bertini et al., Basic Gamma-Ray Data for ART Heat Depo&ition Calculations, ORNL-2113, pp 8,16 (Sept. 17,

1956).

b4, C. Claiborne and T. B. Fowler, The Calculation of Gamma Heating in Reactors of Rectanguloid Geometry,

ORNL CF-56-7-97, p 20 (July 20, 1956).

The energy absorption buildup factors for each
E, in the core were obtained by fitting the ex-
pression:

B(R) = 1 + du) + blun)?

to the tabulated NDA values for aluminum and
water.? Where possible, the empirical expression
was weighted as the core density ratio, two-thirds
aluminum to one-third water. The values of 4 and
b for each value of E,, which are good up to ap-
proximately 7 mean free paths, ore given in
Table 5.3.2.

The core heating at the core-shield interface
was obtained by summing the result of Eq. 1 over
all seven core planes. The results are shown
in the second column of Table 5.3.3.

Because anofher method of cclculaflon was used -

for. the second step, it was felt that the heating in

some overlappmg region should be computed by
~ both methods for purposes of comparison. There-
fore, the method described above was also used to.
_determine the heating in the shield at the first
'alummum ‘shell and in. the initial layer of lead.
~This was uccompllshed for the aluminum shell by

assuming that the flux entering the aluminum was

,'th"e"éqm'e_ usfi thct-whi ch left ihe cbre. The oluminum

 

9H. Go!dstem und J E Wllkms, Jr., Calculauons o/.

the Penetrations o Gamma Rays. Final Report, NYO-
3075 (June 30, 1954).

- Toble 5,3,2. Values of the Coefficlents a and b
Used in Determination of B(R) for the Core

 

 

Eo a b
0.5 1.10 0.45
1 0.955 0.185
2 0.78 0.05
4 0.57 0.0

6 0.42 0.0

8 0.33 0.0

 

Table 5.3.3. Heat Contributions from Eqy Energy Groups

 

Heat Generation (fiafis/cma-Mw)_

 

In First

 

Ey | Ar In First
(Mev) Core-Shield Aluminum Lead
' ' interface Shell interval
0.5 0.034 0.047 0.649
1 0.042 0.063 0.389
2 0.050 0.074 - 0.360
. 0.017 0.026 0.251
6 0.003 . 0.005 0.055
8 0.002 0.003 0.025
Total 0.148 . 0.218 - 1L729

 

31

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

heating was then the product of the heating.per
energy group in the core and the ratio of the
aluminum-to-core energy absorption coefficients.
The results are tabulated in the third column of
Table 5.3.3.

For the lead heating, Eq. 1 was resolved for the
heating in the aluminum at the aluminum-lead
interface, by using the new distances involved and
by including the effect of the thin shell of alumi-
num on the linear and energy absorption coef-
ficients. The buildups in the core and aluminum
were assumed to be the same. The resulting heat-
ing as a function of the energy group (which
agreed to within several per cent of that estimated
in the preceding paragraph) was multiplied by the
ratio of the lead-to-aluminum energy absorption
coefficients to give the results tabulated in the
fourth column of Table 5.3.3. Naturally, it is not
expected that this estimate for lead will be as
good as that for the aluminum shell, and no reli-
ance is placed on it other than for comparison with
the calculation discussed below.

 

) Hicollided) = [¢4-Me§ (4 Mev) + &9 y0
collided collided

Since the Monte Carlo code yields gamma-ray
heating as a function of initial energy and angle of
incidence, it was necessary to find the energy and
angular distribution of the flux incident upon the
shield. The energy flux incident upon the shield
was estimated by dividing the calculated heating
for each initial energy group at the core-aluminum
interface into collided and uncollided fluxes. The
uncollided flux for each E, was determined by
solving Eq. 1 without buildup factors or absorption
coefficients.

The total collided heating for each E in the
core was found by:

Hicollided) = Hiuncollided) [B(R) = 11 .

This was assumed to be made up of collided fluxes
of the energy group E, under consideration plus
collided fluxes of each lower energy group E cor-
responding in size, except for the upper group, to
the initial groups E,. For example, for E, at
4 Mev:

Ho(2 Mev) + &y oy #5(1 Mev) +

collided

+ ¢0-5-Mev ua(O.S Mev) + @9 1 .Mev £ (0-1 Mev) | K .

 

Heating Within the Shield

The Oracle Monte Carlo code'® for the heat de-
position resuiting from the penetration of gamma
rays through stratified slab shields was used to
calculate the heating in various TSR-1l shield
layers which were arbitrarily chosen as follows:

1. the first layer of aluminum,

2. the six equal regions in the first layer of lead,
3. the second aluminum layer,

the water layer,

the third aluminum layer,

the five equal regions in the second lead layer,
the two equal regions in the final borated
aluminum layers which were considered all
aluminum,

These layers are shown in pictorial form in

Fig. 5.3.7.

N A

 

10¢c. D. Zerby and S. Auslender, ANP Quar. Prog.
Rep. Dec. 31, 1955, ORNL-2012 (Part 1-3), p 203.

312

collided collided

The relative values of the collided fluxes in the
above equation were obtained from the ratios of the
areas under the differential energy spectrum curve
for each E,, as given by Goldstein and Wilkins?
for a point isotropic source after passage through
one mean free path of water. This was possible
since the relative spectrum for each E; is nearly
independent of the distance from the source for a
number of mean free paths. Figure 5.3.8 illustrates
such a curve for an E, of 4 Mev. With the ratios of
the collided fluxes for each E known, it was pos-
sible to solve Eq. 3 and obtain the collided num-
ber flux for each energy group. The sum of the
collided and uncollided fluxes for each E gave the
total incident flux for each energy group.

It was assumed for the uncollided fluxes that the
radiation was isotropic in the source planes and,
since the back plane correction is unimportant, the
angular distribution should also be nearly the
same as for infinite plane sources. This angulor

 
 

 

9

41rrze"°rlo

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

175 i" 2-01{-060-34
®
150 -
- T o
o WATER
o ALUMINUM
.25 A WATER !
A ALUMINUM [ fror=2
flN /
’/
A
1.00 ' /
| /f/
I ../
Hor =1
0.50 I \ =T
' \‘ /:fl?’
nfil;-,. ] #:’:‘:—-"—’:fi
a a ==
0.25 g % {=-Mev 2-Mev 4-Mev
|l —— jentlf ——— — Sttt ————— - —— T
é “"g > GROUP GROUP GROUP
-~ 0
o o
0 l L 1
0 05 1.0 1.5 2.0 2.5 3.0 35 4.0 45 5.0
£ (Mev)

Fig. 5.3.8. NDA Moments Method Calculation of Collided Gamma<Ray Spectra in Aluminum ‘and Water Media
from a 4-Mev Point Source.

distribution was determined by weighting the con-
tribution per solid angle with the exponential at-
tenuation with distance,

The uncollided flux arriving at point P from a
source ring of radius p on an infinite plane at
distance z (Fig. 5,3.9) is

e~ HR
41TR?

 

S( =) 2np dp .

This flux arrives in 40 about @ and

 

p'dp __,z.?ztaneseczedfl_ _
RZ z2 sec? @ = fon 040 "

Therefore, -

(4)

S
F(6,2) d6 = 5 fan g e~H2/cos 8 g9

UNCLASSIFIED
2-01-060-36

 

 

Fig.
Uncollided Flux.

5.3.9. Geometry Used to Determine Incident

313

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

where

0 = angle between the incident radiation and
the normal from the shield to the point,
z = normal distance from plane to point.

By assuming that the source is uniformly dis-

tributed over the core planes, Eq. 4 can be in-
tegrated over the core to give

B s’ cos 0
# .

where S’ is the volume source strength. Let
A = $’/2p and normalized by. the requirement that

n/2 g
6) ¢ = f A sin 0 d6 (l -e Homax/ €0 ),

which yields
% ¢ = ALl = Eyfpz_ I,

where

 

E,(x) = x f 2 dy -

x

The heating for each E, at any point in the
shield from the uncollided flux is given by the
expression

(8)
H= | f(6) AsinOcos 6d0|1 - e
j; in 6 co [

’

max/cos 9]

where

cos @ = conversion from fluxes to current as
required by Monte Carlo calculations,
f(60) = Monte Carlo result for fraction of en-
ergy deposited in a porticular region
for a particular E, and angle of inci-
dence.

A typical plot of f(6) for an initial energy E of
2 Mev is shown in Fig. 5.3.10. Substitution of
Eq. 7 for A in Eq. 8 yields (for heating in the ith
region of the shield)

314

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
2-01-059-237
50
40
30
<
s
2
=
.
20
Pby
/4/%
e
Pby Pbg
Pbyp —
A Pbryy
0
0 30 60 o0
8 (deg)

Fig. 5.3.10. Percentage of 2-Mev Incident Gamma Rays
Deposited in Varicus Regions of the TSR-ll Shield as
a Function of 6.

©) HfE) =

i mux/cosfl
-

j;/,.(e) ¢(E) sin 0 cos 0 df T B0

ITIGX)

Although Eq. 9 was developed for the uncollided
flux, for this calculation it was assumed that the
collided flux reaching the shield had the same
angular distribution as the uncollided flux. Hence
Eq. 9 was solved for the total heating in each
region in the shield by using the total flux in each
energy group. One example of the integrand of
Eq. 9 for one energy is shown in Fig. 5.3.11.

a9

 
 

b

 

 

UNCLASSIFIED

 

 

 

   

 

 

 

 

 

 

 

 

2-01-059-236
0.0
H = F(8)sin8 cos 8
' b ' Pb,
0.08 //\
=
s /
~ 006
§
°
s /
O
=
e /
<L
Y 004
¥
/&
0'02 /\
% .
0 — 1 :
o 30 60

 

8 (deg)

Fig. 5.3.11." Heating in Various Reglbns of the TSR-II
Shield Resulting from 2-Mev Incident Gamma. Rcy: as
a Function of 0. i : '

PERIOD ENDING SEPTEMBER 30, 1957

The total heating evaluated in each region in the
shield by Eq. 9 is shown in the form of a histo-
gram in Fig. 5.3.12. The heating in the aluminum-
water-aluminum region and in the final lead and
aluminum region was not significant compared
with that in other regions and hence is not illus-
trated. Also shown in Fig. 5.3.12 are the results
of the estimates for the first aluminum ond lead
regions calculated with Eq. 1 and described in
the last two paragraphs of the preceding section.

Conclusions

The calculation of heating in the regions outside
the core should be an upper limit of the actual
case and therefore serve as a safe basis for de-
sign. This is due to the source of leakage radia-
tion being slightly overestimated, since the buildup
factors used for attenuation within the core apply
to an infinite medium and ignore the depression
from the lead layer around the core. This depres-
sion is largest for collided photons arriving in
directions nearly tangential to the core-shield
interface., These photons would be expected to

contribute strongly to the heating just at the

inner edge of the lead where the heating is the
most important. .

A further extension of this calculation is pres-
ently under way to remove the dependence of the
calculation on the energy and angular distribution
at the shield interface by utilizing the data from
the calculation by Bowman and Trubey (see

 Chap. 5.4, this report) to extend the Monte Carlo
- calculation from the shield to the source planes

in the core.

315

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

2-04-060-29
———— MONTE CARLO

1.5 , wwew ESTIMATED
=
=
")
§
s
£ 40
o
z
o
=
b
Lt
I

0.5

Pbyg Pby Pb,  Pby Al Al
. At |Pb, | Pb, | Pbs) Pb, | Pbs | Pog| Al H20I Al ey f——p—— | |
0 05 1.0 15 20 25 30

DISTANCE FROM CORE —SHIELD INTERFACE (in.}

Fig. 5.3.12. Gamma-Ray Heat Generation Rates in the Proposed TSR-11 Lead Shield Region as a Function of
Distance from the Core Shield Interface.

316

 

It
 

 

 

 

PERIOD ENDING SEPTEMBER 30, 1957

5.4. SHIELD DESIGN

C. E. Clifford

MONTE CARLO CALCULATION OF GAMMA-RAY
PENETRATION OF ALUMINUM SLABS

D. K. Trubey

A Monte Carlo calculation of the energy spec-
trum and angular distribution of gamma rays which
penetrate 3-, 6-, and 9-in.-thick aluminum slabs
has been carried out by using the Oracle code
discussed previously,! For this calculation the
gomma rays were assumed to be incident normal
to the slabs and to have initial energies of 0.662
Mev (the energy of Cs'37 gamma rays).

The energy balance for the three slabs, that is,
the fraction of energy reflected, absorbed, or trans-

 

F. L. Keller

and the energy in each sector is ¢(i, i + 1), where
i=12 ..., 13, then

(1) 2mA f"f wbdy = Hli, i + 1)
i1
and
1
(2) 204 [ pbdp = N,
0

Table 5.4.2. Angular Sectors and Energy Groups for
Transmitted Collided Energy

 

 

 

 

 

 

 

 

 

mitted, is given in Table 5.4.1. The fraction

] : . Angular A Energy Energy
transmitted is fu:;ther brolfen down into fhe. un- Sector, Polar Angle* || Group, Range
collided and collided portions. The transmitted s . (de

. . 9) E (Mev)
cotlided-energy fraction is then subdivided into
10 energy groups which are transmitted through 1 05 1 0-0.03
12 angular sectors, as shown in Table 5.4.2. Since 10
the incident energy is normal to the slab, there is 2 5= 2 0.03-0.05
no dependence on an azimuthal angle. The dis- 3 1015 3 0.05-0.07
tributf:‘on of the transmitted collid:d-ener;;.;y fraction 4 15-20 4 0.07-0.09
int ari ener oups and a ar t
info the various energy groups and angular sectors 5 2025 5 0.09-0.19
obtained from the calculation is presented in
Tables 5.4.3 through 5.4.5. The total collided 6 25-30 6 0.19-0.29
energy transmitted per unit solid angle for the three 7 30-35 7 0.29-0.39
slabs is shown in the histograms in Figs. 5.4.1,
8 35-40 8 0.39-0.49

5.4.2’ and 5.4.3.

An attempt has been made to fit the histogram 9 40-45 9 0.49-0.59
plots with analytical expressions, |f the distribu- 10 50 -60 10 0.59-0.662
tion is represented by : n 6070

#(f) = Acosb o, o 12 70-90
Is, Auslender and A. T. Futterer, ANP Quar, Prog. *Angle with the normal.
Rep. March 31, 1957, ORNL.-2274, p 286.
Table 5-.4.1.. Fractions of Incident Energy Reflected, Absorbed, and Transmitted
stab Energy Fraction
Thickness Transmitted
' (i)’ Reflected Absorbed
* Uncollided Collided
6 0.0498 0.802 0.0457 0.102
9 0.0483 ~ 0.909 0.00977 0.0329

 

317

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 5.4.3,

Fraction of Incident Energy Transmitted (Collided) in Energy Group E and
3-in.+Thick Aluminum Slab

Sector S for

 

Fraction of Incident Energy

 

 

 

 

 

 

 

 

S Ea1 E-2 E-3 E=d  E=5 E<6  E=7 E=8  E=9  E=10  Tofal
1 0 -0 0 -0 0.0001268 0.00004886 0 0.00008103 0 0.002545 0.002802
2 0 0 0.00004422 0.0001214 0.0005534 0.0002395 0.0002185 0,0002553 0,0003704 0,007583 0.009386
3 0 0 0.00005692 0.0001443  0.0007589 0.0006092 0.0004436 0.0003113 0.0008273 0.0114) 0.01456
4 0 0 0.00008955 0.0002400 0.0009322 0.001012  0,0009466 0.0007345 0.001046 0.01344  0.01844
5 0 0 0.0001136  0.0002382 0.001167 0.0009087 0.0002115 0.001380  0.001411 0.01181 0.01724
6 0 0 0.0001244 0.00018110 0.0009606 0.001033 0.001344 0.001136  0,01356 0.001181 0.01952
7.0 0 0.0001645 0.0002012 0,001573 0.001028  0.000951¢ 0.001722 0.01308 0.0004495 0.01917
8 0 0 0,0001836 0.0004086 0.001685 0,001071 0.001019 0.001518 0.0131¢9 ‘ 0 0.01908
9 0 0.00001872 0.0002882 0.0005362 0.003537 0.003130  0,002531 0,01222 0.007077 0 0.02934
i0 0  0.000008793 0.0003237 0.0006943 0.003529 0.002386 0.002794 0.01414 0.0008499 | 0 0.02473
11 0 0.00001838 0.0002007 0.0044540 0.002870 0.001584  0,005738 0,003876 0 0 0.01473
12 0 0.00001828 0.0002145 0.0005365 0.003065 0.001385  0.003609 0.001155 0 0 0.009983
Total —6_ 0.00006417 0.001804  0.003747 0.02075%90 0.014365  0.019807  0.03854 0.05141 0.04842 0.19898
Table 5.4.4, Fraction of Incident Energy Transmitted (Collided) in Energy Group E and Sector § for
6-in.-Thick Aluminum Slab
s Fraction of Incident Energy
E=1 E=2 E=3 E=4 E=5 E=6é E=7 E=8 E=9 E =10 Total
1 0 0 0.00004386 0 0.0001805 0 0.0001698 0 0.00006487 0.001164 0.001623
2 0 0 ¢ 0.00009864 0.0002970 0 0.0002733 0.0004263 0,0001370 0,002511  0,003752
3 0 0 0.00004889 0.00001509 0.0004989 0.0003587 0.0001356 0.0003558 0,0006073 0.005327 0,007347
4 0 ¢ 0.00002740 0.00003203 0.0007486 0.0006221 0.0002833 0.0005175 0.0008115 0.005891 0,008934
5 0  0.00004552 0.00006332 0.0001901 0.0004834 0.0008475 0,0005601 0.001407 0,0003428 0.005755 0.009695
6 0 0.00001986 0,0001945 0.0002363 0,000336% 0.0006139 0.0009168 0.0005214 0.006614 0.0009825 0.0l1044
7 0 0 0,0001201 0.0002634 0.001028 0.0002668 0.001813 0.0007509 0,008405 0 0.01265
8 ¢ 0.00001183 0,0001232 0,0003411 0.001845 0.001374 (0.0007652 0.001347 0,005172 0 0.01098
9 0 0.00004562 0.0003138 0.0006895 0.001695 0.001313 0,001417 0.005916 0.,002624 0  0.01401
10 0 0.000055982 0.0002887 0,0002954 0,002164 0.001961 0.001914 0.003420 0.0004832 0 0.01054
- H 0 0.000010710 0.00007994 0.0002982 0,001628 0.0008028 0,001932 0.002165 0 0 0.006917
12 0 0 0.0003549 0.0002493 0.001207 0.001075 0.001813 0.0004993 0 0 0.005198
Total ? 0,0001396 0.001659 0.002711 0.01212 0.009243  0.01199 0.01734 0.02526 0.21691 0.10208

 

318

 
 

}]

 

PERIOD ENDING SEPTEMBER 30, 1957

Table 5.4.5. Fraction of Incident Energy Transmitted (Collided) in Energy Group E and Sector S for
9-in.-Thick Aluminum Slab

 

Fraction of Incident Energy

 

 

 

 

 

 

 

 

E=1] E =2 E=3 Em=d E=S5S E=b Ew=7 Ew§ E=9 E =10 Total
1 0 0 0 0.000019192 0.00001206 0 0.00001067 0.00009016 0 0.0005113 0.0006434
2 0 0 0.000002177 0 0.0001452  0.00002060 0.00003799 0 0.0000266% 0.001168 0.001400
3 0 0 0.00001126 0,00004094 0.,0001060 0.00001336 0.00002675 . 0.00005468 0.0004522 0.001172 0.001877
4 0 0 0.00003177 0.00003685 0.0001807 0,0002483 0.00002514 0,0006556 0.0003300 0.001517  0.003045
5 0 0 0.00007779 0.00023096 0.0002761 0.0002738 0.0002246 0.0002301 0.0006285 0.001189 0.003131
6 0 0 0.000021 19  0.0001118 0.0004371 0.0001221 0.0004751 0.0008794 0.001685 0.0001525 0.003885
7 0 0.000001519 0.000009835 0.00002760 0.0003723 0.0003694 0.0001137 0.0001820 0.001323 0.0002243 0,002623
8 0 0.00001083 0.00002665 0.00004738 0.0006920 0,0007257 0.0001406 0.0002184 0.001169 0 0.003031
9 0 0 0.0001301 0.0001344 0.0006778 0.0008160 0.0003214 0.0009388 0.002061 0 0.005149
10 0 0.000003085 0.0001184 0.0001807 0.0011319 0.0006917 0.0007707 0.001425 0.0001359 0 0.004457
1" 0 0.00001308 0.00009961 0,0001479 0.0004565 0.0002915 0.0007973  0.0003005 0 0 0.002106
12 0 0.000000630 0.00003136 0.00008799 0.0005592 0.0006130 0.0002402 0.00003278 0 0 0.001595
Total 0 0.00002914  0.0005602 0.00]0658 0.005057 0.004206 0.003274 0.005007 0.007811 0.005934 0,3294
UNCLASSIFIED UNCLASSIFIED
2=-04-059-224 " 2-01-059-223
0450 04502 0° 20° 30° 40°
0425 50° 0425 50°
& >
& &
Z 0400 g 0400
-
g 60° g 60°
= Q
2 oo7s 2 0075
5 5
= 2
2 ' 2 70°
§ 0.050 o G 0.050
g &
0.025 80° 0025 80°
ace ot 90°

 

 

 

.'A

. ,Elg.,".5.4.l.=:- F'lh'c'(:t_iorln of ,lr‘:cfrdént_' Energy

Tfuhsmi"ed

(Collided) per Unit Solid Angle: 0.662-Mev Gomma

Rays No_rrnul_ly’ incident on _d 3-in.-T_hIc
Sltab. ' -

k Aluminum

 

 

 

~ Fig. 5.4.2. Fraction of Incident Enefgf Trensmitted
(Collided) per Unit Solid Angle:

Slab.

0.662-Mev Gamma

Rays Normally Incident on o &in.-Thick Aluminum

319

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
2-01-059-226

i0* 20° 30° 40°

 

o
Q.030

0.025

g
O

FRACTION OF INCIDENT ENERGY
o o
o o
o &

0.005 80

 

 

 

20

 

Fig. 5.4.3. Fracticn of Incident Energy Tronsmitted
(Ceollided) per Unit Solid Angle: 0.662-Mev Gamma
Roys Normally Incident on a 9-in.-Thick Aluminum
S|ub.

where u = cos 0 and N = the total collided energy
transmitted. Equation 2 reduces to
2rA

3) b + 1

 

and Eq. 1 reduces to

(4) NE*' — b = ¢, i + 1)
or
$i, i + 1)
min = w7 - N

for py = 1, py4 = 0; therefore

=V G, i+ 1
5 b =1-F q“__rvll-

=1

, forj# 1,13 .

Equation 5 may then be used to get 11 estimates
of b, It was found that b did not vary greatly, and
so an average was used to determine A from Eq. 3,
The resulting equations, ¢ = A cos® 6, are also
shown in Figs. 5.4.1 through 5.4.3 and indicate no
change in the angular distribution with slab thick-
ness.

320

MONTE CARLO CALCULATION OF THE GAMMA-
RAY PENETRATION OF LEAD-WATER SLABS

L. A, Bowman? D. K. Trubey

The DO2C55 series of calculations of the gamma-
ray penetration of lead-water slab shields, described
previously,! has been continued, and a total of
512 problems has now been computed. All combi-
nations of the following parameters have been used:
slab thicknesses of 1, 2, 4, and 6 mean free paths,
consisting of eight varying percentages of lead
preceding or following water; initial gamma-ray
energies of 1, 3, 6, and 10 Mev; and angles of
gamma-ray incidence of 0, 60, 70.5, and 75.5 deg
from the normal. The results include information
on heating, dose rates, and energy fluxes through-
out the slabs; energy and angular distributions
reflected back by the slabs; and energy and angular
distributions transmitted through the slabs.

The data from all these problems are now being
analyzed. Typical plots are shown in Figs. 5.4.4
through 5.4,7, which represent the heating caused
by 3-Mev gamma rays incident at various angles on
lead slabs 4 mean free paths thick., Note that the
heating values in Figs. 5.4.6 and 5.4.7 are re-
duced by a factor of 2 for ease in plotting.

MONTE CARLO CALCULATION OF GAMMA-RAY
DOSE RATE BUILDUP FACTORS FOR
LEAD AND WATER SHIELDS

L. A, Bowman? D. K. Trubey

The DO2C55 series of Monte Carlo calculations
discussed in the preceding paper also included
calculations of dose-rate buildup factors for lead
and water shields, sketches of which are presented
in Fig. 5.4.8, Only the results for gamma rays
incident normal to the slabs are reported here.

In the first group of calculations buildup factors
were determined for all-lead slabs which were 1,
2, 4, and 6 mean free paths in thickness at the
initial gamma-ray energy. The initial energies were
1, 3, 6, and 10 Mev. The results are compared
in Fig. 5.4.9 with the buildup factors computed at
NDA by the moments method for an infinitely thick
slab, The results are in agreement except near
the rear of the slabs, where the Monte Carlo curves
are lower, This is due to the reduction in back-
scattering in the slabs having finite thicknesses,

 

2On assignment from USAF.

 
 

 

 

 

 

 

i

2.00

1.80

1.60

1.40

1.20

1.00

0.80

0.60

0.40

J, FRACTION OF TOTAL INCIDENT GAMMA~RAY ENERGY ABSORBED PER MEAN FREE PATH

0.20

0

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
2—-01-059-249

CODE NO.: 3414
REFLECTED: 0.056
TRANSMITTED: 3.869
ABSORBED:96.075
Eo = 3 Mev
8= 0 deg

LEAD

 

0 o 2 3 4
o, NORMAL THICKNESS OF COMPOSITE SLAB IN MEAN FREE PATHS
AT INITIAL ENERGY

Fig. 5.4.4. Energy Absorption in a Lead Shield as a Function of the Shield Thickness: 3-Mev Gamma Rays

Normally Incident,

321

 
 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
2-01-059-220
2.00

CODE NO : 3442
REFLECTED: 0.759 )
TRANSMITTED: 0.148
1.80 ABSORBED: 99.093
Eo= 3 Mev
6= 60 deg

1.60 _ LEAD

1.40

.20

1.00

0.80

0.60

0.40

J, FRACTION OF TOTAL INCIDENT GAMMA-RAY ENERGY ABSORBED PER MEAN FREE PATH

0.20

 

0
0 1 2 3 4

Lor, NORMAL THICKNESS OF COMPOSITE SLAB IN MEAN FREE PATHS
AT INITIAL ENERGY

Fig. 5.4.5. Energy Absorption in a Lead Shield as a Function of the Shield Thickness: 3-Mev Gamma Rays

Incident at 60 deg. ‘ j

322

 
 

 

2.00

1.8

o
O

o

~RAY ENERGY ABSORBED PER MEAN FREE PATH/2
o

0.8

0.6

J/5, FRACTION OF TOTAL INCIDENT GAMMA

0.2

0

N
o

D
o

0

0

0.40

O,

o

PERIOD ENDING SEPTEMBER 30, 1957

UNCLASSIFIED
2-01-059-224

CODE NO.: 3443
REFLECTED:1.697
TRANSMITTED: 0.053
ABSORBED:98.250
Eo = 3 Mev

8 =70.5 deg

LEAD

 

0 1 - 2 3 4
Hor, NORMAL THICKNESS OF COMPOSITE SLAB IN MEAN FREE PATHS
AT INITIAL ENERGY

Fig. 5.4.6. Energy Absorption in a Lead Shield as a Function of the Shield Thickness: 3-Mev Gamma Rays

incident at 70.5 deg.

323

 
 

 

 

 

e sl AL 12 |

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
2-01-059-222
2.00

CODE NO : 3414

REFLECTED: 3.350

TRANSMITTED: 0.040
1.80 ABSORBED: 96.610

Ep= - 3 Mev

8 =75.5deg

0
o

LEAD

1.40

.20

{.00

0.80

0.60

0.40

J/Z, FRACTION OF TOTAL INCIDENT GAMMA-RAY ENERGY ABSORBED PER MEAN FREE PATH,,

0.20

 

0
0 1 2 3 4

por, NORMAL THICKNESS OF COMPOSITE SLAB IN MEAN FREE PATHS
AT INITIAL ENERGY

Fig. 5.4.7. Energy Absorption in a Lead Shield as a Function of the Shield Thickness: 3-Mev Gamma Rays
Incident at 75.5 deg. : S

324

’)

 
 

 

 

 

UNCLASSIFIED
2-04-059-238

  
  

SLAB CONFIGURATION NO. 4

H,0

H,0

 

TOTAL THICKNESS T
gt
PARAMETERS:

INCIDENT ENERGY = 4,3, 6, AND 40 Mev
TOTAL THICKNESS 7 = 4,2,4,AND 6 MEAN FREE PATHS
sec8=14,2,3, AND 4 -

Fig. 5.4.8. Lead and Water Siab Shield Configurations
Used for Monte Carlo Calculations.

UNCLASSIFIED
2-01-059-249

 

{0 ’ - - v .
[ =——emme MOMENTS METHOD (NDA) FOR
INFINITELY THICK SLAB

MONTE CARLO METHOD (ORNL) FOR
SLABS WITH FINITE THICKNESSES
S 8=0deg

£= INITIAL ENERGY

 

 

 

 

 

 

 

 

 

1 { 3 Mev_l
— VERTICAL LINE INDICATES - —
REAR OF FINITE SLAB—————} ELOI‘/W“-"’

 

oo &

 

 

 

 

8,, GAMMA-RAY DOSE-RATE BUILD-UP FACTOR

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0 ¢ 2 3 4 5 6 7 8 9 0 # 12 13
' 7, SLAB THICKNESS (cm)

| Flg.lhrs.4.r9. Comparison of Dése—Rute Buildup Factors
for Lead Slabs Computed by Monte Carlo Method with
Those Computed by Moments Method. Normally incident

gamma rays.

L—‘/ /.-"

2 — 5 "-' 6Mev]
' / /‘/ —”.ijb/‘::q ‘
/I/fi/;_‘r 10 Mev

S

10

PERIOD ENDING SEPTEMBER 30, 1957

The Monte Carlo buildup factors for all-water
slabs of finite thickness are compared with NDA
buildup factors for on infinitely thick slab in
Fig. 5.4.10. In this case the Monte Carlo calcu-
lations always predicted higher values for the
buildup factors.

Values of the Monte Carlo dose-rate buildup
factors at the rear of composite lead and water
slab shields are plotted in Figs. 5.4.11 through
3.4.14 for total composite slab thicknesses of 1,
2, 4, and 6 mean free paths. The effect of alter-
nating the position of the lead and water is apparent
from these figures. :

All the buildup factors determined for composite
slabs in this series have been compared with

UNCLASSIFIED

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

50 2—-01-059-250
| | l [
| =——aae MOMENTS METHOD (NDA) FOR INFINITELY ___|
THICK SLAB
| MONTE CARLO METHOD (ORNL) FOR SLABS — |
WITH FINITE THICKNESSES
8=0
20 Eo= INITIAL ENERGY
. .
e
o II/VERTICAL LINE INDICATES
: REAR OF FINITE SLAB
=2
g 10 —AE0= ! Mev
2
o
p
et
[+
w
8
- 3M
:[- ev
T
a o 1
= ]%7{' € Mev
-
N o =
% =" | 10 Mev
-
o 45 . 90 135 180 225 270

1,SLAB THICKNESS {cm)

Fig. 5.4.10. Coméuri;on of Dose-Rate Bulldup Factors
for Woter Slabs Computed by Monte Carlo Method with
Those Computed by Moments Method. Normally incident
gamma rays.

325

 
 

 

 

 

 

 

ANP PROJECT PROGRESS REPORT

UNCLASSIFIED
2-01-059-254

H20 Hz0

Pb

Eq, INITIAL 7= tmfp

ENERGY
O ~m===ee-- { Mev
3 Mev
A ————— § Mev
& ———-—= 40 Mev

T ={imfp

By, GAMMA-RAY DOSE-RATE BUILD-UP FACTOR

 

4
O 025 050 075 100 100 075 050 025 O
ptpy/T, NORMAL THICKNESS IN MEAN FREE PATHS AT INITIAL ENERGY

Fig. 5.4.11. Monte Corlo Dose-Rate Buildup Factors
at the Rear of Composite Lead-Water Slab Shields
1 Mean Free Path Thick. Normally incident gamma

rays.

UNCLASSIFIED
2-01—-059-253

H,0

L’T = 4mfp

3

h

----- == { Mev
. 3 Mev

b =———— G Mev
A ——==—10 Mev

8,, GAMMA-RAY DOSE-RATE BUILD-UP FACTOR
n

 

{
0 025 050 075 +t00 1.00 075 050 025 O
2oy / 7, NORMAL THICKNESS (N MEAN FREE PATHS AT INITIAL ENERGY

Fig. 5.4.13. Monte Carlo Dose-Rate Buildup Factor
at the Reor of Composite Lead-Water Slab Shields
4 Mean Free Paths Thick. Normally incident gamma
Rays.

326

UNCLASSIFIED
2-01-059-2%2

S

Hp0

o

’
ENERGY

0 =eemmme==n | Mev
® 3 Mev
A =——= 6 Mev
A ————-—— {0 Mev

N

   

By, GAMMA-RAY DOSE-RATE BUILD-UP FACTOR

\ i
0 025 050 075 100 100 075 050 025 O
pfpb/r, NORMAL THICKNESS IN MEAN FREE PATHS AT INITIAL ENERGY

Fig. 5.4.12. Monte Carlo Dose-Rate Buildup Factor
at the Rear of Composite Lead-Water Slab Shields
2 Mean Free Paths Thick. Normally incident gamma
rays.

" UNCLASSIFIED
2—01-059—254

8 =0deg
H,0 H,0

10 _
£, INITIAL I___
7=6mb=  ENERGY T=6
ommm=-== | Mev
o—— 3 Mev
5 6 Mev
\ A=——-—10 Mev

5., GAMMA-RAY DOSE-RATE BUILD-UP FACTOR

 

1
0 025 050 075 1.00 100 075 050 025 O

Bl /T, NORMAL THICKNESS IN MEAN FREE PATHS AT INITIAL ENERGY

Fig. 5.4.14. Monte Carlo Dose-Rate Buildup Factor
at the Rear of Composite Lead-Water Slgb Shields
6 Mean Free Paths Thick. Normally incident gamma
rays.

)

C
 

 

PERIOD ENDING SEPTEMBER 30, 1957

values obtained by use of o formula, proposed by X, X, = thickness in mean free paths of the
M. H. Kalos of NDA3 in which the buildup factors first and second materials, respec-
independently computed for lead and water are tively. '

combined to determine the buildup factor for a

. i . For a water-lead shield the formula is:
composite slab consisting of the two materials,

 

B,(x,) -1 - (flcs/p') -
1 1.7X, 4 = __._.........'I (1 - e Xz) [Bz(Xl +X2) - Bg(x])] ’

 

B(X.,X,) = B.,(X.,) + | ————— ¢
1402 2v" 2 :

B,y(X,) -1 | (/1)

where

For a lead-water slab ('IhO'l’ is, lead followed by Bes = Compfon scattering cross secfion,
water) the formula is written as follows: i

. p, = total cross section,
B(X],Xz) = BZ(XZ) + Tables 5.4.6 and 5.4.7 show the buildup factors

determined both by Oracle calculations and by the
Kales formula, It should be pointed out that the
numbers representing the Oracle calculations are
. taken from the curves of Figs. 5.4.11 through
where B o , 5.4.14 and do not represent actual data points,

B,(X,) - 1 :
+W[Bz(xl + X,) - Bz(xz)] '

 

B,,B, = gamma-ray dose-rate buildup factors 3M. H. Kalos, Some Theoretical Methods and Results

. , s in Gamma Ray and Neutron Shielding, Shielding Sym-
for the first and second materials, re- posium held at the Naval Radiologicuf Defense Labora-

spectively, - tory, Oct. 1719, 1956, 12NDP 1965, p 93-94.

Table 5.4.6, Monte Carlo Gamma-Ray Dose Rote Buildup Factors at Rear of Lead=-Water Slab Shields: Comparison
of Oracle Calculations with Values Obtained with Kalos Formula

 

Incident Shield (mfp) B

 

 

Gamma-Ray —_— _ T Ratio of B, (Calculated)
Energy Lead Water _Oracle Kalos to B_ (Kalos)
- Cadlculations Formula r
{(Mev)
1 1 1 233 2.18 1.069
2 3.30 3.13 1.054
4 5.81 - 5,60 1.038
| 5 7.51 7.14 1.052
2 1 2.72 2,51 1.084
2 3,90 : 3.47 1.124
3 533 4.73 o aZ
, 4 695 6,06 , 147
3 1 300 3.00 1.033
2 4.50 4,05 1.1
3 6,22 5.36 1.160
4 2

327

 
 

 

 

 

 

ANP PROJECT PROGRESS REPORT

Table 5.4.6 (continued)

 

 

 

 

Incident Shield (mfp) B, e SR
Gamma-Ray _ . Ratio of B, (Calculated) -
Energy Lead Water Oracle Kalos to B, (Kalos)
(Mev) Calculations Formula .

3 1 1 1.95 1.90 1.026
' 2 . 2,58 2,54 1.016
3 3.30 3.10 1.065
4 3.80 3.60 1,056
5 4.2 4.07 1.034
2 ] 2.47 2.45 1.004
2 3.15 3.00 1,050
3 3.69 3.52 1.048
4 4.18 4,03 1.b37
3 ] 2,80 2.85 0.982

2 3.44 2.89 1.190
3 4.00 3.47 1.153
6 1 1 1.60 1.55 1.032
2 2.03 1.98 1.025
3 2.50 2.43 1.029
4 2.66 2.67 0.996
5 2.71 2.74 0.989
2 1 1.98 1.80 1.100
2 2.34 2.14 1.093
3 2.55 2.50 1.020
4 2.62 2.70 0.970
3 1 2.11 1.96 1.077
2 2.31 2.18 1.060
3 2.41 2.52 0.956
10 1 1 1.41 1.38 1.022
2 1.68 1.64 1.024
3 1.90 1.88 1.0Nn
4 2.04 2.03 1.005
5 2.14 2.12 1.009
2 1 1.55 1.50 1.033
2 1.78 1.72 1.035
3 1.97 1.93 1.021
4 2.10 2,07 1.014
3 1 1.65 1.65 1.000
2 1.81 1.84 0.984
3 1.94 2.02 0.960

 

 

328

g

[ 3

 
 

Fbovom s

PERIOD ENDING SEPTEMBER 30, 1957

Toble 5.4.7. Monte Carlo Gamma-Ray Dose Rate Buildup Factors at Rear of Water-Lead Slab Shields: Comparison

of Oracle Calculations with Valuves Obtained with Kalos Formula

 

 

 

 

 

Incident Shield (mfp) B,
Gamma-Ray Ratio of B, (Calculated)
Energy Water Lead 0"‘":'? Kalos to B, (Kalos)
(Mev) Calculations Formula

1 1 1 1.70 1.66 1.024
2 1.86 1.92 0.969
3 2.05 2.12 0.967
4 2.22 2.35 0.945
5 2.42 2.55 0.949
2 1 2.20 1.96 1.122
2 2.18 2.17 1.005
3 2.31 2.45 0.943
4 2.52 2,63 0.958
3 1 2.50 2.27 1.101
2 2,46 2.50 0.984
3 2.72 2.73 0.996

4 2
3 1 1 1.74 1.62 1.074
2 1.92 1.88 1.021
3 2.15 2.16 0.995
4 2.39 2.44 0.980
5 2.66 2.68 0.993
2 1 2.28 1.96 1.163
2 2.39 2.24 1.067
3 2.51 2.54 0.988
4 2.70 2.78 0.971
3 1 2.80 2.33 1.202
2 2.88 2.60 1.108
3 2.97 2.88 1.031
6 1 1 1.55 1.50 1.033
2 1.75 1.79 0.978
3 2.01 2.04 0.985
4 2.20 2.27 0.969
5 2.40 2.56 - 0.938
2 1 2.09 1.89 1.106
2 2.24 2.23 1.004
3 2.38 2.50 0.952
4 2.48 2.82 0.879
3 1 2.48 2,31 1.074
2 2.58 2.68 0.963
3 2.61 3.04 0.859

wa

- . e, T
= o
4 e

329

 
 

 

 

 

 

‘Table 5.4.7 (continved)

 

 

 

hcident

B

 

 

Shield (mfp) ,

Gamma-Ray Ratic of B (Calculated)
Energy Water Lead Oracle Kales ‘to B, (Kalos)
(Mev) Calculations Formula

10 1 1 1.3% 1.41 0.986
2 1.52 - 1.55 0.981

3 1.65 1.70 0.971

4 1.77 1.92 0.922

S 1.90 2.07 0.918

2 ] 1.70 1.72 0.988
2 1.79 1.92 0.932

3 1.88 2.16 0.870

4 1.96 2,39 0.820

3 1 1.95 2.00 0.975
2 1.99 2,34 0.850

3 2.08 2,61 0.797

 

- 330

 

)