ORNL-1816.txt

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ORNL-1816

This document consists of 177 pages.

Copy ?3 of 218 copies. Secwimsrimm

Contract No, W-7405-eng-26

AIRCRAFT NUCLEAR PROPULSION PROJECT
QUARTERLY PROGRESS REPORT

For Period Ending December 10, 1954

W. H. Jordan, Director
S. J. Cromer, Co-Director
R. I. Strough, Associate Director
A, J. Miller, Assistant Director
A. W, Savolainen, Editor

DATE RECEIVED BY INFORMATION AND REPORTS DIVISION
(JANUARY 6, 1955)

DATE ISSUED

JAN 21 1955

OAK RIDGE NAT!ONAL LABORATORY
e “Operated by
CARBIDE AND CARBON CHEMICALS COMPANY
© A Division of Umon Carbide and Carbon Corporation
' Post Office Box P
QOak Ridge, Tennessee

 

(EARATARTIIR

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1. G. M, Adamson
2. R. G, Affel
3. C. R. Baldock
4. C. J. Barton
5. E. S. Bettis
6. D. S. Billington
7. F. F. Blankenship
8. E. P. Blizard
9. G. E. Boyd
10. M. A. Bredig
11. F. R. Bruce
12. A. D. Callihan
13. D. W. Cardwell
14. J. V. Cathcart
15. C. E. Center
16. G. T. Chapman
17. R. A, Charpie
18. G. H. Clewett
19. C. E. Clifford
20. W, B. Cottrell
21. D. D. Cowen
22. S. Cromer
23. R. S. Crouse
24. F., L. Culler
25. L. B. Emliet (K-25)
26. D. E. Ferguson
27. A. P, Fraas
28. J. H. Frye
29. W. T. Furgerson
30. W.R. Grimes
31. E. E. Hoftman
32. A. Hollaender
33. A. S. Householder
34. J. T. Howe
"~ 35. R. W. Johnson
36. W. H. Jordan
37. G. W. Keilholtz
38. C. P. Keim
39. M. T. Kelley
40. F. Kertesz

41. E. M. King
42. J, A, Lane
43, C. E. Larson
44, M. E. LaVerne

 

ORNL-1816

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Reports previously issued in this series are as follows:

Period Ending November 30, 1949
Period Ending February 28, 1950
Period Ending May 31, 1950
Period Ending August 31, 1950

Period Ending December 10, 1950

Period Ending Mcrch-IO, 1951
Period Ending June 10, 1951
Period Ending Sep’remrber 10, 1951
Period Ending December 10, 1951
Period Ending March 10, 1952
Period Ending June 10, 1952
Period Ending September 10, 1952
Period Ending December 10, 1952
Pericd Ending March 10, 1953
Period Ending June 10, 1953
Period Ending September 10, 1953
Period Ending December 10, 1953
Period Ending March 10, 1954
Period Ending June 10, 1954
Period Ending September 10, 1954

 

 

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FOREWORD

This quarterly progress report of the Aircraft Nuclear Propulsion Project at ORNL re-
cords the technical progress of the research on circulating-fuel reactors and all other
ANP research at the Laboratory under its Contract W-7405-eng-26. The report is divided
into three major parts: . Reactor Theory, Component Development, and Construction,
Il. Materials Research, and ill, Shielding Research,

The ANP Project is comprised of about 400 technical and scientific personnel engaged
in many phases of research directed toward the achievement of nuclear propulsion of air-
craft, A considerable portion of this research is performed in support of the work of other
organizations participating in the national ANP effort. However, the bulk of the ANP
research at ORNL is directed toward the development of a circulating-fuel type of reactor.

The effort on circulating-fuel reactors was, until recently, centered upon the Aircraft
Reactor Experiment. This experiment has now been completed, and the operating ex-
perience is described in Section 1 of Part |,

The design, construction, and operation of the Aircraft Reactor Test (ART), formerly
the Circulating-Fuel Aircraft Reactor Experiment (CFRE), with the cooperation of the
Pratt & Whitney Aircraft Division, are now the specific long-range objectives. The ART
is to be a power plant system that will include a 60-Mw circulating-fuel reflector-moderated
reactor and adequate means for heat disposal. Operation of the system will be for the
purpose of determining the feasibility, and the problems associated with the design, con-
struction, and operation, of a high-power, circulating-fuel, reflector-moderated aircraft
reactor system. The design work, as well as the supporting research on materials and
problems peculiar to the ART (previously included in the subject sections), is now re-
ported as a subsection of Part |, Section 2, ‘‘Reflector-Moderated Reactor."

   

   

 

 
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CFOREWORD v o et eeeeen e e e e e e e o

SUMMARY & i i ettt i i it e i . e e e

PART |. REACTOR THEORY, COMPONENT DEVELOPMENT, AND CONSTRUCTION
1. CIRCULATING FUEL AIRCRAFT REACTOR EXPERIMENT ........... ... ...,

Operation of the Aircraft Reactor Experiment ... ......... e e .
Low-power experiments . ..ttt it ot eetoens sttt onenens e e e e
High-power experiments ... v v v vt v v ittt ian ot enossotoanosssnononens
Mathematical analysis of approach to critical ... ..o i, P e e b e e s

Operation of ARE Pumps « v v i v it i ittt it ittt sttt et ie st s oaronnenen
Loading of the ARE v 4 vttt it ittt et eeteosoneononseenenoeenenenes
Anclysis of Ledk in Sodium System . . ih i v e e o st e s e ettt

Preliminary Estimates of Corrosion in ARE . ..... St e e e e e

2. REFLECTOR-MODERATED REACTOR .t vvuvvuennenunnnnn, P ..
Expunsion-Tank;n’d'Xenon?Removal System ........ e e e e e
Control Rod Desrigrn ConSIerations « v v e v v ensneenenenennn, B
Fill and Drain System « v v v e v e e nenenns et e e e e e e

Design Physics v v v v v vttt i i i ittt tennnonnnnsennns e
Ac'rlvcmon of the Inconel core shells Gt a e e e et e ettt e e

3. EXPERIMENTAL REACTOR ENGINEERING + vttt vvtennerorteeeneneenennes

In-Pile Loop Component Development ....... e e e e e e e
Horizontal-shaft sump pump . o v v vttt it ittt ittt ittt ittt sttt nnannsse
Heaot exchonger .. .o v ittt ittt it i ittt st i it v et et an et nasansans

Pump Development Program .« v o v v v v v vttt et et e ettt et e e,
ARE-type sUMP PUMPS « v v v v v et e o v v o v e o e e e e e Ve e e Ve
Large-scole pump developmenf co e

--------------------------

   

 

Demgn and Operqtlon of Forced- Clrculahon Corrosmn and Mass Transfer Tests

Beryihum-sodtum-lncone! mqss frcnsfer

 

- Notura!-gas hect sourc:es for forced-curculahon Ioops

oooooooooooooooooooooooooo

    
  
 
  
 

  
 

'Pump for heat exchcnger tests

Gas-

 
 

Furnace Heat Source 'Deve'l'opment ceeeas

" Natural-gas burner .. ....... ettt ee e e e e

Trap for Fluoride Vapors « v v v v ittt vttt ittt et o nstneeoooeesnosnsosenesss

 

 

'Hec:f, Exchanger Deve[opment oo RS e e e e e e

“Sodium-cooled’ 100-kw furnace tests . e

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Fused salts in Inconel . .......... e e aeaes 44
Sedium in mulhmetcfl loops e e e ee e e T o . 45
e
~~Heat exchcnger tests . ettt e e T U 46
Header ook tost ... LoL oLl

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4. CRITICAL EXPERIMENTS &0t oteretereneeesenenenaneneanananennns 52
~ Reflector-Moderated Reactor . v v v v ev e ti e oo rnsenesosoesonssasonssnsss 52 .
PART Ii. MATERIALS RESEARCH
5. CHEMISTRY OF MOLTEN MATERIALS ... ovvvrinteitennenennennen, weeeee. 57 Yy
Phase Equilibrium Studies « v v v v v v v it ettt i ettt eneeeenas BRI 57
Solid phase studies in the NaF-Zr ,-UF  system . .....coovvivn it e 57
Phase relationships in UF _-bearing systems . .o v v v vt i iv e vvntenonnoonnoonnns - 58
Solubility of UF, in NaF-KF-LiF eutectic . ....ovvviveriiiii iy 59
UF , stabilify v v e e it ittt i ettt eneenenssossnneroonssnessnsnooessns 60
Dif\?erenfialr’rhermcl ANAlYSIS & i v ittt e it e e e e e 61
Chemical Reactions inMolten Salts .« . . v v ittt i ittt i sttt e it e 61
Reduction of FeF, by H, in NaF-ZrF, systems .. ....oooviniiiiii ., 61
Reduction of Ul':4 to UI:3 influoridemelts v v v v i s ittt et te et i e e 61
Electrochemistry of fused salts .. .. i ittt i it ie it vessonnnnnneesss 62
Stability of chromous and ferrous fluorides in molten fluorides ......... ... .cvvuy 63
Reduction of UF, by structuralmetals ... oovvviv e, 64
Production of Purified Molten FIUotides « v e v e v e vononensnnnsonsnsns e e _ 66
Electrolytic purification of zirconium-base fluorides + v v e v v v v et v n i vnronssons 66
Fluoride production facility . . v v i ittt ittt e it ittt e s e cennennsnnnens 68
Alkali fluoride processing facility ..... e it e s e e 69
Purificationof KF and RbF .o o i i i i i i it ittt iiei ittt ennenns cees e /0
Preparation of various fluorides + v v v v v v et vt v eveeenens Cesve o c et 71 -
Chemistry of Alkali Hydroxides v uuveeueeuenneenrenreneensoennenesennnns 71 j
Purification of hydroxides v v v it inenienietvannosononsnsonrsosnsas 71 -
Effect of additives on hydrogen pressure over NaOH-Ni system . .. v vt e v et v vveanns 71 §
Fundamental Chemistry of Fused Salts + v v v v ettt it i in it oo neenonoanannsnos 72
Solubility of xenon in fused salts . ........ Gt ettt et ettt 72
X-ray diffraction studies in salt systems ... v it vi v et nnennas cevees e e 72
Physical chemistry o vu e i v et innstennetnneseeessosennsssenssssnnos 73
Viscosity measurements o4 v v v oo v v s aeeoeessnossosersossssosossanas s 75
6. CORROSION RESEARCH ................. ® & & & 5 5 & # 8 2 F & & & 2 B " b . ..... V'.- 76
Thermal-Convection Loop Corrosion Studies v v v e v v v evsenovvosensonoenenss e 76
Inconel foop containing UF; in alkali-metal-base mixtures ... ... .o 0ol., e 76
Type 316 stainless steel loops containing UF, in alkali-metal-base mixtures . ... .. ce 76
Effect of induced potentials ... v i i ii ittt e essensar e 77
Hastelloy B thermal-convectionloops ...t veive i v e e ettt 77
Special alloy thermal-convectionloops .+ vttt v vt vt e v eonnenen e e e PP 4
Sodium-Beryllium-Inconel Compatibility ... ciivitie vttt sneesasnsassssnnnsss 78
Static Corrosion Studies .o e ittt it ittty e e 80
Brazing alloys on stainless steel .. ... .. o0y et he e 80
Brazing alloys on nickel .. ... .. .. t e et e e e em et e e 80
Screening tests of carbides v v v i v vttt i i e e e raee s e R 82 : g‘
Magnesium-lithium alloy inwater . ..o i ittt ittt ie it iiieeeeneenonns 86 .
Rubidium Investigations « .. vuvevon. C et et e e . 87 - « [

 

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. 'RADIATI'ON_DAMAGE . e e T R T
MTR Stahc Corrosmn Tesfs e '
~ Effect of lrrodlahon on UF -C F e
| Minldture 1n- P|le Loop N Ty v eewnn i e e

 

Fundamental Corrosion Research v v v e v ch e e b b e et i ettt
Mass transfer in liquidlead « o v v oo v v v v v et et ettt st e ce e
Fused hydroxides as acid-base analog systems . ......... e d e e e
Flammability of alkali metal ‘solufions at high ’remperotUres e ceeees

Chemical Studies of Corrosion v veveeeeeeensos et s e e s et e sttt
Effect of temperature on the cortosion of Inconel melts with N|F2 additions ... .o
Effect on chromium additions on fhe corrosion of Inconel and fluoride melts with

NIF AddifioNS v vt v st eevesttnncancsssans bt e s et et st e e

Effecf of chromlum valence state on corrosionoflnconel .+ v v v v e v v s st oo e v cvosoeos

. METALLURGY AND CERAMICS +.vuvenuvennnennn, e e -

Development of Nickel-Molybdenum Base Alloys ... v i i eiinanevenns coene
Fabrication experiments « .o e vt e ot evovsossoonvenes et e ettt e
New alloys ...... c e eesa et et s s e st et an e C e sttt
Oxidation and oxidation protection + .. vv s os. Gt ettt et et e s
Radiator fabrication i v v it v et oot aeneanssarosessosssstosesassssonss

Welding « v ot vt vttt sttt snsosonsssnssssessosstosstossatsaroerocesss

Mechanical properties studies v oevee et eonesnnoessns e e cons

Special Materials Fabrication +iveeseeoeeeseesoosoesoatoaanssesoosnnsss
Duplextubing v iiinennoaesen Gt i e s e e et e oo e
Boron carbide shielding ...... e e e e e v e et et e s et es et
Tubular fuel elements «vvvveennensas Ceees s esa e e esena Cee et
Control rods v v vvvoes v e et ee s Gt et et e e i e st e
Al-UO, elements for shielding experiment . .... ettt et r e e e e e e

Brazing Alloy Development ... ... C et e e cesecs e a e ce s e
Heat Exchanger Fabrication . . ... ... et ee et cr et e e

Bery”iumoxideFUelElemenfs -oo.oocc ooooooo lol.otlnooocooooo'o;no-oooo

. HEAT TRANSFER AND PHYSICAL PROPERT[ES .................... coenus

Fused Salt Heat Transfer +uvvuveeensenn et ees s e e vee b e s e e s
In-Pile Loop Heat Exchanger Analysis ...... s ee s s et s sasesasane s e s e nen e
Free Convection in Fluids with Volume Heat Sources . ..veeeeveeenne. e v
Heaf Transfer Effecfiyenes.s of Reactor Coolants v v v v vt o v v en v ensnse e e

ART Core Hydrodynamics ... ..vuu ee s et

. ART Heot Trunsfer » @ .» *» e . ......... g . * ’. . * L] . . L L . - ® & & 0 @ - . . ..-‘.. . * .. . .. ‘. _':-..7&.‘.
' Physmc[ Properhes Measurements . e e e e e e e e e
Heat capacity « oo e e v v e v eonnss et i e e i e e e e e e e "
VISCOSIi‘y ceee e s e e e e et e

 
 

Bench £S5t + o v o o o v e v v o s s oo e oo oo oottt ss e oot
Heat transfer calculations . v v v o v vt vttt ettt et es s vt ot aoanonosnoassssen

 

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Removal of Xe!35 from Molten Fluoride Fuels « v v v v v i vnvevnenene IEERE .. .' R 124 b
LITR Horizontal-Beam-Hole Fluoride-Fuel Loop .o v v v v v iv v iins ceeee e vee. 125 .
Crégp and 'Sfress-Corrosion T eSS v i e v i et et ittt et i e e e e 126 -
Remote Me:tdllography ..................... e e e Ry e ]27 -
‘ Mass Speétrogmphic Analyses ...ttt eri i it e e e . 127
10, ANALYT]CAL STUDIES OF REACTOR MATERIALS P DR e . 129
o :Anafy'rlccl Chemlstry of Reactor Materials  « v v v e v v s vevnnenneeeneeenns e 129
Determination of uranium metal in fluoride salt mixtures .. ... . i, 129
, Determmatlon of trivalent uranium in fluoride fuels ...... e et 129
, Determination of oxygen influoride fuels .. v v e it i it it i et e 131
: -~ Determinationof sulfur .. .. i i i i i s i i e 131
: Deterin_incl_tion of fluoride in NaF-KF-LiF-base fuels ............. e veee e 132
Petrographic Investigations of Fluoride Fuels .. ....ovvinen . e e 133
- ANP Serwce Laborafory ................................... e .-_. . 133
11, RECOVERY AND REPROCESSING OF REACTOR FUEL ... vvvvenns e e 134
_Flssmn Pr_oduc? Removal v vvnin it ittt iiiiieiaanes T 7
‘ Appliéafions of Fused Salt~Volatility Processes v v v v v v s vt v vt s oncnoeenens .. 137
Aircraftreactor fuels . oo v v ittt i i it e it i i e e i e 138
Heterogeneous reactor fuels .. ... . ... oo e e s et 138
PART Ill. SHIELDING RESEARCH .
12, SHIELDING ANALYSIS &+ v et v veeeeeeeeeneneeieneaeenenns eeee.. U3 |
:; Slant Penetration of Composite Slab Shields by Gamma Rays .« o v vv v v v v v nnseannen 143
Emergy Absorption Resulting from Incident Gamma Radiation as a Function of
Thickness of Materials with Slab Geometry . . . . . i ittt it i i et ieenen 144
:';: A Formu\a_fion for Radiation Injury to Include Time Effects .. .. v v v i v i iiiiiien 145
Analysis of Some Preliminary Differential Experiments .. .. ... ... 0o 146
Interpretation of Air and Ground Scattering at the Tower Shielding Facility .. .......... 147
i 13. LID TANK SHIELDING FACILITY .. \vuvintiniineentanennanennns S |
Effective Neutron Removal Cross Section of Lithium , ... ... ... e 151
VGVE-ANP Helical Air Duct Experimentation . ..o vv v v esnenenonnnnnnns e 153
Reflector-Moderated Reactor and Shield Mockup Tests .. ..... it e st . 155
14. BULK SHIELDING FACILITY vttt et it eiite e eencaoannnneenes e - 156
" A Search for Short-Half-Life Nuclear Isomers in K39, Rb83, Rb87 and Zr90 e 156 ,
5 15. TOWER SHIELDING FACILITY & tttttttteeeeeeeeeeeeaneeneenneenenenns 158 o}
| TSF Experiment with the Mockup of the GE-ANP R-1 Shield Design . ................ - 158 ]
' R | PART 1V. APPENDIX . o
: ORGANIZAT]ON CHART OF THE AIRCRAFT NUCLEAR PROPULSION PROJECT ......... 67 - E
. SECRET

 

 
 

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' ANP PROJECT QUARTERLY PROGRESS REPORT

SUMMARY

PART I. REACTOR THEORY, COMPONENT
DEVELOPMENT, AND CONSTRUCTION
1. Circulating-Fuel Aircraft Reactor Experiment

The Aircraft Reactor Experiment was operated
successfully for over 100 Mw/hr. During this
time the maximum equilibrium outlet fuel tempera-
ture was 1580°F, and the maximum power level
was over 2.5 Mw. The temperature coefficient of
reactivity 2 min after the introduction of a pertur-
bation was —5.5 x 10~3, although its instantaneous
value was considerably higher. As a result, the
reactor was extremely stable while operating at
power, and there were no control problems. The
critical mass was 32.5 Ib of U235 and thus was
in reasonable agreement with the calculated
value of 30 ib. 7

Operation of the process systems and their
instrumentation was exceptionally trouble-free
during the six days of operation rom the time the

‘equipment was sealed in the pits until the sodium

and the fuel were dumped.
It is believed that most of the fission-product

- gases evolved from the liquid fuel. In a 25-hr run,

it was determined that no more than 1 part in 30 of
the xenon poison remained in the fuel.

An unexpectedly steep initial slope that was
observed in the usual plot of [1 = (1/multiplication

constant)} was found to be due to the formation of
a_thermal-flux maximum near fhe flssaon chombers\_“
"“dumps in heat exchanger fes'rs.' Imenswe work
has been done on the development of a hydraullc
“:'cn'cun fhat prowdes a fuel pump, a “fuel expansion
__tank, an& a xenon-removcl system clnd that performs
_the varlous funct:ons requ:red ‘of the sys’rem.
) lnmal fesfs of a “Lucite f]ow model have given
“oromisingresults. Des;gn criteria were esfabl:shed_» -
f'for the contro| rod fho will

.as the mulhpllcahon increased.  Once the shape -
‘of fhe spct;al distribution of the thermal-neu’rron:‘%_’\'
flux was set up, the flssuan chambers reglsferedi“f
'_,,ron!y ‘the generci increase in flux |eve| and the
“Tf:,‘;‘:kcoun’r ra’re mcrecsed slow!y '

  
 

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subjected, it was interesting to note that all

 
   
 
 
 
  
   
     
 
   

t ;_-'_.‘for remot
power runs. Smce this was  the hlghest “ambient
temperature to which ARE pumps have been

'systenis.

rotary elements functioned satisfactorily.

Fuel samples were withdrawn from the ARE
during operation below criticality and at very low
power. As was anticipated, the chromium content
of the fuel increased after each addition of fuel
concentrate. Just before the first additions of fuel
concentrate were made, the chromium concentration
of the carrier material was 100 ppm and the carrier
had been circulated for 155 hr. Since the system
had been operating isothermally, the chromium
content of 100 ppm was considered to represent
0.5 mil of attack on the Inconel. By the time the
final sample was taken, after a total of 307 hr of
operation, the chromium content was 445 ppm,
which represents about 5 mils of attack. A plot
of the data obtained showed that the chromium
content of the fuel had started to level off. Since
it was not possible to sample the fuel during full
power operation, no information on mass transfer
can be obtained until the equipment can be sec-
tioned.

2. Reflector-Moderated Reactor

The reactor assembly and the installation for
the Aircraft Reactor Test (ART), formerly the
Circulating-Fuel Reactor Experiment (CFRE), have
been the subject of analytical layout and detailed
design studies, Component development tests are
proceeding as scheduled, and specifications have
been complefed for radiators to be used as heat

 
 
 

r be prmcapally a means
for confrolhng ‘the meda

ing formu]ated Desngn work 'is Under way on @
f[l@;and drain sys’rem “for the reactor assembly that
will provlde a qunck disconnect coupling suitable’
operahon w:th hlgl’l-iemperafure llqwd

 

  

An analysis has been made of the activation of
the Inconel core shells that would come from two

péi’qture level of the
h:mrecc’ror fuel. Plans for coohng the control rod are

 
oo i s

 

 

 

ANP QUARTERLY PROGRESS REPORT

main sources: the activity of the cobalt in the
Inconel and the activity of the fission fragments
that strike the Inconel and remain there. The
analysis showed that little would be gained by
trying to obtain cobalt-free Inconel, since the
activity could not be reduced below that caused
by the fission fragments. Adjustments are being

‘made in the constants used in multigroup calcu-

lations in order to decrease the disparity between
the calculations and the results of critical experi-
ments. '

The various types of facilities suitable for the
“operation of the ART have been considered, and it

is now planned to install the reactor assembly in a
double-walled container with a water-filled annulus.
The type of container was selected after careful con-
sideration of the hazards involved in the operation

~ ofthe reactor. This typeof installation provides for

containing the products resulting from a nuclear
accident and/or chemical reaction of all combusti-
bles in the installation, minimizing the likelihood
of serious damage from an explosion caused by
sabotage or bombing, and removing and disposing
of volatile fission products evolved during the
course of operation,

3. Experimental Reactor Engineering

Progress has been made in the design and testing
of components for a horizontal, entirely enclosed
loop for insertion in an MTR beam hole. The loop
will be used to circulate proposed reactor fuels in
the MTR flux so that radiation effects on the fuel
can be studied. A layout arrangement of the loop
has been made and is being improved consistent
with specification and component changes. Tests
of two horizontal-shaft sump-type pumps were made
with the fluoride mixture NaF-ZrF, circulating at
a temperature of 1350°F for 500 and 1000 hr,
respectively.  The Graphitar-lapped steel face
plates wore considerably in both tests, but they
did not fail. A single-tube salt-to-air heat ex-
changer that will fit inside the beam hole has been
fabricated.

- Sump pumps of the type used in the ARE are

being evaluated for application in large-scale heat

exchanger tests, and pumps for ART use are being
developed. One ARE-type sump pump is being

tested at K-25 for performance and life determina-

tions. The pump is operating at pump speeds of

" up to 1500 rpm at a maximum capacity of 40 gpm,
~and it is circulating the fluoride mixture NaF-ZrF,-
UF, at a maximum pump inlet temperature of

1350°F. Operating time now exceeds 3000 hr,
The pumps for the ART inherently possess many
imposing design problems because of the large
capacity requirements:  fuel pumps, 650 gpm,
50-ft head; sodium pumps, 430 gpm, 125-ft head;
NaK pumps, 2800 gpm, 280-ft head. Investigations
of impeller cavitation characteristics and inlet
and entry conditions are being made, along with
studies of sealing, cooling, lubricating, and driving
problems, _

Forced-circulation corrosion and mass transfer
tests with fused salts in Inconel systems are under
way. Considerable difficulty has been experienced
with failures of the thin-walled tubing (l/z-in. oD,
0.020~in. wall thickness) at bends and welded
joints. The loop design has been altered by elimi-
nation of the economizer and by using 0.045-in.-
wall tubing. One loop of the modified design is
now being operated and has accumulated about 150
hr at a Reynolds number of 10,000 and a tempera-~
ture gradient of 200°F.

A third beryllium-sodium-Inconel mass transfer
test has been completed following 1000 hr of
operation. The maximum operating temperature (at
the beryllium section) was 1300°F, and the
Reynolds number was about 190,000. Another
similar loop has been fabricated and is being
assembled. A loop in which sodium will be circu-
lated in type 316 stainless steel is being fabri-
cated.

The heat exchanger test program has included a
header leak test and a series of pump tests with
an ARE moderator-coolant type of pump. The
header leak test (fuel to NaK) showed that self-
plugging of an NaF-ZrF ,.UF , fuel through 0.002-
in.-dia leaks can occur. The pump tests showed
that speeds of up to 3700 rpm could be safely
achieved with this type of pump.

Developmental work on a gas-furnace heat source
that can be used in large heat exchanger tests was
continued. A sodium-cooled 100-kw furnace was
tested for about 120 hr before a gross leak termi-
nated further operation. An output of 85 kw was
obtained at a sodium outlet design temperature of
1500°F. Two natural-gas burners, based on @
design development of the Esso Marketers, were
built and tested. Heat releases of 400 kw at
3100°F were achieved with the first burner, and
1 Mw ot 3300°F was obtained with a second
burner that was improved on the basis of the daia
obtained from the first burner.

ol

e

 

 

 

 
i

 

 

«

N

A trap for fluoride vapors was developed that
can be used in any high-temperature molten fluoride
system to prevent plugging of the gas lines leading
to or from the vessel containing the fluoride,
particularly if the salt contains ZrF ,. Successful
operation of this device was achieved during the
ARE pump tests at K-25. '

4. Critical Experiments

The second critical assembly of the reflector-
moderated reactor was constructed that incorporated
a beryllium island. Criticality was achieved with
4.66 kg of U235 at a density of 0.092 g of U235 pe
cubic centimeter of fuel annulus. This Iocding
was shown, by calibrated controf rods, to contain
0.92% excess reactivity, equivalent to about 0.31
kg of U235, and therefore the critical mass of the
unpoisoned assembly was 4.35 kg of U235, The
experimental results thus did not agree well with
the calculated critical fuel loading of 7.16 kg of
U_235 or 0.142 g of U235 per cubic centimeter of
fuel annulus. A series of measurements of neutron
and fission-rate dls’rrlbuflons and of reactivity

_coeffuments will be made prior to the incorporation,

in the structure, of additional materials to simulate
reactor components.

PART Il MATERlALS RESEARCH
5. Chem:stry of Molten Mufermls

Additional data obtained in solid phase studles
of the NaF-ZrF ,-UF, system have essentially
confirmed the previous findings. There was some
indication from thermal analysis of mixtures along

the Na_U,F ., -Na Zr ¢F3y ioin that these com-

ponents do not form a complete set of solid solu-
‘tions. A new quenchlng appcratus is bemg used
for the studies that is expecfed to’ ‘accelerate 'rhe
~‘rate ‘at which data can be obfamed Sfudles of ’rhe;_\
_ fi-,_;phase relahonshlps m UF -bearlng systems haved
EE ?_con’rlml

S

 

in the NaF KF LIF eufectic to provide a usable

R SYSfem for the reflector-moderated circulating-
el recctor. EXberl
,Of UF as 0 fun‘ ' P S a.wn
‘that UF is more stable at elevated’ temperatures

 
 
 

than eshmofes of its thermodynamic properties
had indicated. It was also fournd in studies of

\

 

o d, nd work was s'rarted on NaF-LaF, and -
. ‘KF- LGF “systems, since Lul':3 is known to be a

“stable ¢ stand-m ' for the more drfflcultly handled
UF,. Experlmem‘al evudence was obtained which
lndicc'red that UF is probcbly sufflmenfly soluble

PERIOD ENDING DECEMBER 10, 1954

UF ; mixed with KF and heated in Inconel ccp'su.les

that disproportionation of UF, occurred at a lower
temperature than would have been expected in the
absence of KF. For differential thermal analysis
of the uranium-bearing mixtures, improved graphite
containers with provision for the thermocouples to
be immersed directly in the melt are now being
used,

Investigations of the chemical reactions in
molten salts are continuing in an effort to under-
stand the mechanisms involved in the purification
of fluoride mixtures and in the reduction of UF
to UF, in fluoride melts. Similar studies of the
corrosion products of Inconel in fluoride mixtures
and of the reduction of UF, by s’rructurai metals
are under way.

Production of purified zirconium-base fluorides
mixtures has been continued on a large scale in
the 250-lb-capacity facility. |t is anticipated that
by full-time operation of this facility by a crew
of one engineer and eight technicians, a sufficient
stock pile of material will be available by December
31, 1954 to allow termination of this operation for
some time, This stock pile {to be ~7000 Ib)
should be sufficient to meet the requirements of
Pratt & Whitney Aircraft, other outside requesters,
and those of the Laboratory for research in this
field.

The pilot-scale processing facility continues to
be used for small-scale development of purification
processes for new fuel compositions of interest.
Considerable effort has been expended on attempts
to prepare UF ,-bearing mixtures of NaF-LiF-KF.
Careful control of operating conditions has allowed
production of sufficiently consistent UF j-content

""m'crz"fér‘i'dl to permit its release for corrosion testing.

E[ectrolyhc purlflcchon of NaF ZrF mixtures
hcs made poss:ble a &gmficam‘ cut in processmg
fime of these ma'rerict!s._ Mechnc impurities have

been found to be removed by electrolysis in a

fraction of the time requured by hydrogen strlppmg
It has not been possdaie to cut the processing

“time of UF -bearing mixtures by using electrolysis,

bfb'bd‘b[y'bkecause of the simultaneous reduction of
4 1o UF at the cathode and oxidation of UF

'cn‘ the anode.

‘The tentative values previously obtained for the
solubility of xenon in fused salts were essentially

“confirmed by data obtained with new, improved

apparatus, The new values were 8 x 10~8 and
9.3 x 10~ 8 mole of xenon per milliliter of melt at

T e

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

260°C and 8.9 x 10~8 and 9.6 x 10~8 at 450°C.
X-ray diffraction studies of the binary systems of
alkali  fluorides with uranium trifluorides are
under way,

6. Corrosmn Research

Stud:es of the corrosion of Inconel, type 316
stainless steel, Hastelloy B, and several special
alloys when exposed to the fluoride mixtures of
interest were continued through the use of thermal-
convection apparatus. In the studies of Inconel
loops containing UF, in alkali-metal-base mixtures,
considerable difficulty was encountered in making
the mixtures, in controlling the total uranium con-
tent, and in determining the ratio of UF, to UF,.
Conflicting data were obtained, and efforts are
now being made to achieve better control of the
variagbles involved. An alkali-metal-base mixture
~ containing UF, was also circulated in type 316

stainless steel loops, and, as in the Inconel loops,
‘considerable mass transfer occurred.

Thermal-convection loops constructed of Hastel-
oy B were operated with NaF-ZrF -UF (50-46-4
moie %), with NaF-KF-LiF (11 5-42 0- 46 5 mole %)
containing 12 wt % uranium as UF,, and with
sodium as the circulated fluids. The loops which
circulated the fluoride mixtures showed only small
amounts of corrosion — 1.5 to 3 mils, The period
of operation of the loop has not had a noticeable
effect on the corrosion; therefore it is felt that the
small amount of corrosion that occurs takes place
in a short time and may be a function of the condi-
tion of the original surface. The Hastelloy B
loops operated with sodium exhibited considerable
mass transfer. Since the loops were known to be
covered with an oxide deposit and were not cleaned
before they were filled, additional loops are to be
operated to determine whether the mass transfer
was caused by contamination. Loops constructed
of several modified Inconel-type alloys were
operated with NaF-ZrF ,-UF,. The data obtained
from these loops indicate that reduction of the
chromium content of the alloy to 5% or less greatly
reduces corrosion.
~Tests of the compatibility of sodium, beryllium,
and Inconel in a system have indicated that at
1300°F the flow rate of the sodiumand the spacing

-~ between the beryllium and the Inconel are the con-

trolling factors, Beryllium surfaces exposed to
flowing sodium are unattacked, but surfaces ex-
posed to relatively stagnant sodium at 1300°F
show dissimilar metal mass transfer between the

beryllium and the Inconel. Tests are under way to
determine the minimum spacing that can be tolerated

between beryllium and Inconel when exposed to

slow-moving sodium,

Several type 304 stainless steel T-joints brazed
with experimental alloys prepared by the Wall
Colmonoy Corporation were fested in sodium and
in NaF-ZrF ,-UF ,. Most of the alloys had ‘good
resistance to the fluoride mixture, but only the
10.2% P-13% Cr-76.8% Ni alloy had good resis-
tance in both sodium and the fluoride mixture.
Several T-joints of A-nickel were also brazed
with various alloys and tested in the fluoride
mixture NaF-ZrF -UF , and in sodium hydroxide.
All the alloys tested hod good resistance in the
fluoride mixture, except the 69% Ni-20% Cr-11%
Si alloy, whereas all the alloys had poor resistance
to sodium hydroxide, except the 82% Au-— ]8% Ni
alloy.

The carbides of titanium, zirconium, chromium,
and boron were corrosion tested in fused fluorides,
in sodium, and in lithium. For the most part, these
carbides showed fair resistance in the various
media, except B,C, which was rather severely
attacked in sodium and in lithium. The agueous
corrosion resistance of an 80% Mg-20% Li alloy
which has been proposed as a crew-compartment
shielding material is being studied. Dynamic
tests of the resistance of Inconel to attack by
rubidium are under way.  Preliminary results
showed maximum attack in the hot leg to a depth
of 1 mil,

Previous tests on the mass transfer character-
istics of container materials in liquid lead indi-
cated that alloys in which intermetallic-compound
formation was possible showed a marked increase
in resistance to mass transfer as compared with
the pure components of the materials. Additional
tests of this hypothesis have now been conducted.
The work on fused hydroxides has centered on the
development of a systematic chemistry of these
substances. By an application of an acid-based
theory, at least 12 types of acid-base analog
reactions are predicted for the fused hydroxide
systems. Each of these types is briefly discussed

in the text. Work on the flammability of alkali-

metal solutions at high temperatures has continued,

and additional data have been obtained for bismuth-
rich alloys with sodium which provide a more com-

plete picture of the reactivity of this system. The
effect of water vapor on the reactivity of these
alloys has also been studied,

L

 

 

 

 
 

 

‘ ‘f;‘k

¥

R

 

In further chemical studies of corrosion, informa-
tion was obtained on the effect of temperatures
and of chromium additions on the corrosion of
Inconel by fluoride melts with NiF, additions.

7. Meta"urgy and Ceramics

The nickel-molybdenum-base alloys are being
studied extensively as possible reactor structural
materials with qualities superior to those of Inconel.
Attempts are being made to improve the ductility
and the fabricability of commercial Hastelloy B
through purification and to find another suitable
and improved nickel-molybdenum-base alloy that
has the strength and corrosion resistance of Hastel-
loy B. The results of attempis to improve the
fabricability of the materials are being evaluated
by extrusion experiments because extruded seamless
tubing will be required if any of these alloys are
to replace Inconel as a reactor structural material,
Temperature-cycling tests in air were performed
with Hastelloy B and other nickel-molybdenum
alloys to determine the behavior to be expected in
an NaK-to-air heat exchanger such as will be

required for aircraft application. These tests have
shown that oxidation protection will be required.

Three radiator test specimens have been con-

structed of Hastelloy B for studying the fabrication
methods and for determining the thermal shock and
oxidation resistance of a complicated assembly,
The specimens contained 20 tube-to-header joints,
which were inert-arc welded by using semiautomatic
welding equipment, and 3 in. of Inconel-ciad copper
fins spaced 15 fins per inch and brazed to the tube
with Coast Metals alloy No. 52. The individual

fube—to-header ]omfs were buck-brazed The fcbrl-___
"f)these test ussembhes was rouhne, and
'..f'no unusuol dlfflcuihes were encountered Prlor:
. to evoluatlng the properhes of Hostelloy B weldl‘ixs"l
"f.-f»_'-_-:."metal “the sfrengfh ‘and ducflllty properhes offl'”

ca’rlon

 

commercually avmlabie Hcs’re”oy B plate were

  

.anneohng and (2) solution cnnealmg and aging cn‘h.’

1950°F, These » very prellmmary experumen'rs hqve

~shown that the room-fempera’rure duc'rlln‘y can be
greatly lmproved by an aging freatment prior to
7 service at 1500°F, S ;

  

P "ma (::Hl‘ escnd o g

  

Six new 'Iv

on the creep properties of Hastelloy B, Inconel,
and other alloys for use at high temperatures.

  

“additional” 'rube-burst units _Have been mstol[ed for
studying the effect of Hluoride mixtures and sodium

PERIOD ENDING DECEMBER 10, 1954

Several of the new nickel-molybdenum alloys are
being tested to determine their load-carrying abili-
ties, :

A study of the flow of metals during an impact
extrusion is being made as a part of the efforts
to produce duplex seamless tubing that will have
good oxidation resistance on the outer surface and
good corrosion resistance on the inner surface.
Several suitable B ,C-containing compositions
have been developed for the shield for the ART
heat exchanger; various nonmetallic bonding ma-
terials can be used.

Thirty-five control rods are being prepared for
the GE-ANP project that contain a mixture of 50%
aluminum powder and 50% B,C. Fuel plates are
being prepared for a study of the effect of delayed
neutrons on the over-all shield weight of circu-
lating-fuel reflector-moderator reactors,

A finned surface for a fused salt-to-air heat
exchanger being considered for use in conjunction
with an in-pile forced-circulation loop was fabri-
cated and tested for corrosion resistance to the
fused salt., The corrugated fins were formed from
0.010-in,-thick nickel sheet and were brazed to a
3/a-in.-OD Inconel tube with an 82% Au-~18% Ni
brazing alloy. A 0.020-in.-dia nickel wire was
used as a spacer between each fin segment, and
it served to provide a capillary path for the braze
alloy, which was preplaced at the fin interlock
joint only. The braze alloy used was found to have
excellent resistance both to air and to the fused
salt.

Beryllium oxide ceramics were produced, proba-
bly for the first time, by casting from a basic slip.

’Hifhen‘o an aud had been considered to be neces-_
“sary to assure a well- defloccuhfed ‘slip.  This
“":work is belng done in an efforf to prepore berylllum
';oxlde fuel elemenfs.

8. Heat Transfer cmd Physlcul Properties |

Preliminary  forced-convection heat transfer

" experiments with the ARE fuel, NdF;ZrF4¥UF4
"~ (55.5-40-6.5 mole %), in an Inconel tube, yielded
heat transfer coefficients which differed from
expected values by only 24% after 115 hr of oper-

ation at abou'r T300°F no corrosion deposits were

“observed on the inner tube wall at the end of the

eXperlment., The performance charqcterlshcs of an

“in-pile loop heat exchanger were determined. The

mathematical velocity and temperature distributions
for several free-convection volume-heat-source

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

systems were compiled, and apparatus was de-
veloped for measuring the over-all radial tempera-
ture difference (or heat transfer coefficient) for a
modified version of the flat-plate system in which
it will not be necessary to measure fluid temper-
atures. _

" Several approximate mathematical temperature
"soiuflons for forced-flow volume-heot-source en-
fronce systems were developed which can be used
to predict the temperature structure of struts or
screens located in circulating-fuel reactor flow
passages. The question as to whether electric
currents, which generate heat in the circulating
fluids of experimental volume-heat-source systems,
affect the fluid flow characteristics was investi-
gated. It was found that the hydrodynamic structure
was not mfluenced by the presence of the electric
currents.

The enthalpies and heat capacities of NaF- -ZrF -
(53-43-4 mole %) were determined; the heat
_ éqpacity in the solid state over the temperature

~range 70 to 525°C was found to be 0.18 cal/g.°C,
and the heat capacity in the liquid state over the
temperature range 570 to 885°C was found to be
0.26 cal/g-°C. The enthalpies and heat capacities
of LiF-KF-UF, (48-48-4 mole %) were also ob-
tained; the heat capacity in the solid state over
the temperature range 125 to 465°C was found to
be 0.234 + (0.95 x 10=4) cal/g-°C, and in the
liquid state over the temperature range 565 to
880°C it was found to be 0.657 - (3.93 x 10~4);
cal/g-°C. The viscometry equipment used earlier
was modified so that the accuracy of liquid vis-
cosity measurements was significantly increased.
The more accurate measurements obtained for
i\iaF‘-ZrFfl-UF4 (53.5-40-6.5 mole %) were 30%
lower than the preliminary values reported previ-
ously. The thermal conductivity of molten NaF-
ZrFA-UF4 (53.5-40-6.5 mole %) was found to be
1.2 Btu/hr-ft.°F. The thermal conductivities of
the alkali fluoride mixture NaF-KF-LiF with and
without UF, were compared in the liquid and solid
states.

- 9. Radiation Damage
The program of MIR

capsules containing fused fluoride fuels has been
continued, and irradiated capsules containing both

irradiations of Inconel

- UF,- and UF ,-bearing mixtures have been ex-

-3 ) .
" amined. There was practically no corrosion of the

irradiated UF j-bearing capsules nor of one of the

irradiated UF ,-bearing capsules. There were no
significant differences in the iron, chromium, or
nickel contents of the irradiated fuels, as com-
pared with the starting fuel batches, and there was
no evidence of segregation of either uranium or
impurities, An improved version of the capsule
irradiation facility has been put into service, and,
to speed up the program, arrangements have been
made for irradiating two capsules simultaneously
in separate, but adjacent, facilities in the MTR.
Examinations of a welded nickel capsule con-
taining a UF ,«C,F | solution that was irradiated
in the ORNL Graphite Reactor indicated the un-
suitability of the solution for use as fuel in a
mockup of a circulating-fuel reactor.

A miniature in-pile loop designed for insertion
in a vertical hole in the LITR was bench tested
and was found to be satisfactory. The bench test
included four freezing and melting cycles in 260 hr
of operation at a maximum temperature of 466°F,
Since it was possible to freeze and melt the fused
salt without causing failure of the loop, it may be
advisable to fill the in-pile loop before it is in-
serted in the reactor and then to melt the fuel
mixture after the loop is in position, A Delco
motor has been rebuilt to withstand radiation and
the high temperatures to be used for the in-pile
joop. Heat transfer calculations that predict the
thermal behavior of the loop were completed. A
second model of the loop for insertion in the LITR
is being constructed.

Plans were completed for an in-pile study of the
removal of xenon from molten fluoride fuels under
ART conditions,

The horizontal type of fuel-circulating loop
designed for irradiation in the LITR has been
operated out-ofpile with a non-uranium-bearing
salt and is now being inserted in the HB-2 facility
of the LITR. The creep-test apparatus for testing
Inconel at high temperatures in the MTR is being
bench tested. The stress-corrosion apparatus for
LITR operation has been successfully bench tested,
and an in-pile apparatus is being constructed,
Remote metallographic studies of solid fuel ele-
ments were continued, and additional information
on the relationship between UO, particle size and
radiation damage was obtained,

10. Analytical Studies of Reactor Materials

Methods were developed for the determination

of uranium metal and UF3 in NaF-KF-LiF-base

 

 

 

 
 

 

1)

)

reactor fuels. Uranium metal is determined by
converting it to UH, with hydrogen at 250°C, then
increasing the temperature in an atmosphere of
carbon dioxide to 400°C to decompose the hydride,
and finally measuring the volume of gas liberated
as a consequence of thermal decomposition. Under
these conditions, CO, was reduced to CO by urani-
um metal and also by UF,. A trap of 1,0, was
incorporated to oxidize CO and thus remove this
source of error,

A solution of methylene blue was found to oxidize
trivalent uranium quantitatively to the tetravalent
state without liberation of hydrogen. A procedure
in which methylene blue is used as the oxidant was
developed for the determination of trivalent uranium
in a variety of materials, The method appears to
be applicable to routine analysis.

Two other reagents, cupric chloride and titanium
tetrachloride, will, under selected conditions,
oxidize trivalent uranium to the quadrivalent state
only. The latter reagent can probably be adapted
to an automatic coulometric titration procedure for
this purpose, '

Calibration of the apparatus for the determination
of oxygen as oxide in fluoride reactor fuels was
completed for quantities of oxygen up to 235
mg/liter, In this range the relationship log
k/c = A\Jc + B isvalid; k is the specific conduc-
tivity of water in HF, ¢ is the concentration, and
A and B are constants,

A colorimetric method was adapted for the de-
termination of sulfur as sulfate or sulfide in fluoride
salts. The sulfur is used to form mefhylene blue,
an intensely colored dye, for which the absorbancy
is readily measured. The method was also applled_

to fhe defermmcmon of sulfur m sodlum. A Semi-i_"j_"_‘

in off-gas streams wds set up ‘that is based on the
""rurbldn‘y of bismuth sulflde in a buffered solution,
-~ The specfrophofometrlc htra’non “of fluorlne |nf:"
~ fluoride fuels w1'rh a zwcomum ‘complex was in-
“vestigated ‘and was found fo ‘be unschsfacfory for
routine application,

F’etrogrcphtc examinations were made of severalq'
~ hundred samples ‘of fluonde melts,  Most of the
" samples were from “alkali fluoride systems “eon-

The Anolyflcai ‘Service Laboratory”

taining UF ..
reported 1587 samples which involved 11,541 de-
terminations,

PERIOD ENDING DECEMBER 10, 1954

11, Recovery and Reprocessirfig of Reactor Fuel

A plant for recovering (in seven batches) the
uranivm from the ARE fuel and rmse by the
fluoride-volatility process is being demgned and
construction is scheduled for completion by
December 31, 1955, It is estimated that the amount
of material to be processed will be 12,4 #3 of
NaF-ZrF ,-UF, containing 65 kg of uranium, This
plant will demonstrate, on a pilot-plant scale, the
feasibility of the fluoridesvolatility process as
applied to the processing of a circulating-fuel
aircraft reactor, The feasibility of the process has
been established on a laboratory scale. The basic
equipment, as now envisioned, will consist of a
fluorinator, an absorption column packed with NaF,
a cold-trap system, and a fluorine disposal unit,

This method of recovery and decontamination can
also be used for processing heterogeneous reactor
fuel elements of the type that can be dissolved in
fused fluoride salt by means of hydrogen fluoride.
Compactness of plant, operation at atmospheric
pressure, and economical waste disposal are some
of the advantages of this type of process,

PART I, SHIELDING RESEARCH
12. Shielding Analysis

Calculations made by the Monte Carlo method
with the use of the ORACLE were complefed for
the attenuation of gamma rays in the sides of a
a two-component crew shield and the heating by
gamma rays in the beryllium slabs adjacent to the
gamma-ray slab source. The resulis of the gamma-

ray hecmng study are of interest for calculations of
thermal stresses and consequenf coohng require-

ments for the reflecfor _region of the CIrculatmg-

fuel reflector-moderoted reactor. Attempts were

made to_incorporate the Biology Planning Chart

_No. 1 of the ANP Med:cal Adwsory Group and other

:_l_recommenda"hbrnws into shield designs. It was

demons’rroted that the chcrt could be supplcmfed by

a fhree-parameter mathemahcol ‘formulation which

 lends considerable f]exrblll’ry in the oppllccmon of
_the tolerance limits. Anaiyses of the Tower Shield-
_ing Fccnln‘y (TSF) data mclude a phenomenologlcal -

ancly5|s of the behavnor of neutrons which Ieove

“the reactor shleld cmd dre cnr scaflered mto the'

crew shield. Relatlvely ‘simple ‘models are shown
to explain some of the values of scattered flux

 
s |

 

 

 

 

incident on crew shield sides and rear. In ad-
dition, more detailed calculations of the TSF data
were made to estimate the fraction of scattered
radiation at the maximum altitude which is attribute
able to ground scattering. There is evidence that
for many measurements it will be about 1%, and for
highly anisotropic shields such as the GE-ANP
R-1 it will be no more than about 5%. The con-
clusion is reached that the TSF towers are high
enough for experiments applicable to high-flying
alrplcnes.

- 13, L:d Tank Shielding Facility
At the Lid Tank Shielding Facility (LTSF) the

effective neutron removal cross section for lithium
was determined as 1,01 + 0.04 barns, The meas-
urements were made in a medium of oil behind a
solid slab of lithium. The experimentation with

_GE -ANP hehcal air ducts was continued, and

measuremenfs have now been made beyond an array

‘of fhlrfy-flve 3-in.-dia ‘ducts both in a medium of

water and in a gamma shield medium of steel
Raschig rings and borated water. The presence of
the ducts in plain water increased the thermal-
neutron flux by a factor of 300; in the gamma shield
medium the ducts increased the flux by a factor of
3000, Preparations for the reflector-moderated
reactor and shield mockup tests are being com-
pleted. It is now planned to use solid UO, as the
simulated fuel in the mockups. The UO, will be
mounted on a movable belt.

14. Bulk Shielding Facility
The Bulk Shielding Facility (BSF) was shut

down during much of the quarter for modifications

to the reactor pool. These modifications will per-
mit the use of demineralized water in the pool,
and thus they will eliminate corrosion of the
reactor fuel elements. Several BSF staff members
participated in a program to discover any short-
period isomeric gamma-ray transitions which might
be present in K39, Rb85, Rb87, and Zr?0. Measure-
ments were made at the ORNL Graphite Reactor,
and no previously unknown short-period isomers
were found in K39, Rb®3, or Rb®7. Zirconium-90
yielded a 2.30 £ 0.03-Mev gamma ray with a half-
life of about 0.83 * 0.03 sec. This may be of
importance in design considerations for a circu-
lating-fuel reactor. |

15. Tower Shielding Facility

Measorements made around the GE-ANP R-1
reactor shield in the TSF pool were compared with
similar measurements made earlier at the BSF.
The variation of the thermal-neutron intensity as
the altitudes of the mockup and TSF detector tank
were varied simultanecusly was also measured.
The data indicated that ground-scattered neutrons
are still observable at the maximum altitude for
this particular reactor-shield combination.  For
tests at the 195-ft altitude, lead was placed in the
rear of the detector tank to simulate the shielding
in the crew compartment, and the thermal-neutron
flux and gamma-ray dose rate distributions within
the tank were determined. This experiment was
interrupted for o period of 2"/2 weeks so that the
TSF could participate in an Air Force Project in
which a group of monkeys were exposed to massive
neutron radiation doses.

I‘l :

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Part |

REACTOR THEORY, COMPONENT DEVELOPMENT,

AND CONSTRUCTION

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1, CIRCULATING-FUEL AIRCRAFT REACTOR EXPERIMENT

E. S. Bettis

J. L. Meem

Aircraft Reactor Engineering Division

OPERATION OF THE AIRCRAFT REACTOR
EXPERIMENT

- The Aircraft Reactor Experiment was success-
fully operated during the quarter. Uranium in the
form of molten Na,UF, was added to the barren
carrier, NaZrF., with which the fuel system was
initially filled, to make the reactor critical. The
fuel composition at initial criticality was 52.8-
41.5-5.7 mole % (NaF-ZrF,-UF,), which has a
melting point of 990°F, whereas the final fuel
mixture {which included excess uranium) had a

~ composition of 53.2-40.5-6.3 mole % (NaF-ZrF,-

UF,) and a melting point of 1000°F,

It was initially intended to remotely add the
concentrate to the fuel system from a large tank
which contgined all the concentrate, after first

‘passing it through an intermediate transfer tank.

This system was discarded when temperature-
control and continuous-weight-measuring instru-
mentation on the transfer tank proved to be un-
satisfactory. Instead, o less elaborate, but more
direct, method of concentrate addition was em-
ployed. This enrichment operation invoived the
successive connection of numerous small concen-
trate containers to an intermediate transfer pot,
which was, in turn, connected to the fuel system
by a line which injected the concentrate in the
pump above the liquid level. Each of the concen-
trate containers was weighed before and after a
transfer in order to determine the amount of
uranium injected into the system. The concentrate
was supplied in batches in cans containing from
about 0.25 ib of N°2UF5 (for rod calibration) up
to about 33 |b (as was used during the first
subcritical loading). In the enrichment operation
the pump bowl served as a mixing chamber and
uniformly distributed the concentrate inté the

 circulating stream. (For details of the loading

operation see following subsection on ‘‘Loading

of the ARE.”)

 .The first concentrate addition was made on
October 30, but the reactor did not become critical

until three days later (3:45 PM, November 3).
Most of the intervening time was spent in clearing
the end of the transfer line at the pump, which,
because of limitations inherent to only this par-

ticular design, was difficult to heat and even more
difficult to service.

The approach to criticality was carefully charted
after each fuel addition. The resultant curve of
reactivity [1 — (1/multiplication constant)] as a
function of the addition of fuel (in terms of pounds
of U235 per cubic foot) is presented in Fig. 1.1.
The data from three different ionization chambers
are presented; meters Nos. 1 and 2 were fission
chambers located in the reflector, and the BF3
counter was located external to the reactor at the
mid-plane of the cylindrical side. The unique
shape of these curves (which, when first extrapo-
lated, suggested a much lower critical mass than
was actually required) is believed to be due to
the particular radial flux distribution of the reactor
at the location of the chambers and the change
in this distribution as criticality was approached.

The calculated volume of the carrier in the fuel
system (before concentrate addition) was 4.82 3.
(The only significant check on this value was

-EURET
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0.3

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o 4 8 12 16 20 24 28

U CONCENTRATION (Ib/ft3)

Fig. 1.1. Approach to Criticality.

11

 

 

 
 

 

i B g et s oo b

 

ANP QUARTERLY PROGRESS REPORT

obtained from subsequent analyses of fuel samples,
together with the known amounts of concentrate
added.) While the total amount of vranium (U23%)
added to the system in order to make the reactor
critical was approximately 135 |b, because of the
amounts withdrawn from the system for samples
and in trimming the pump level, the uranium
concentration at criticality was 23.7 Ib/#3; or,
since the calculated volume of the 1300°F core
was 1.37 3, the ‘‘cold,”” clean critical mass of
the reactor was 32.5 1b of U235,

~ Low-Power Experiments

Several ‘‘experiments’’ were performed on the
critical reactor at low power, including reactor
power and rod calibrations. In addition, the
effects of the process system parameters on re-
activity were noted, and a preliminary measure-
ment of the temperature coefficient was under-
taken. The tests were started on the morning of

November 4 and were completed by noon on

November 8.

The regulating rod was calibrated both by the
addition of fuel and by a determination of the
resultant pile period upon withdrawing the rod

(as derived from the inhour equation). The value
of the rod was first obtained by noting the amount
of rod insertion required to maintain a constant
power level as a finite amount of fuel was added
to the system. This information, together with
a calculated value of the mass reactivity coef-
ficient (Ak/k)/(Am/m) of 0.232, permitted a de-
termination of the value of the rod. The technique
of rod calibration by pile period was also employed
both at design fuel flow (48 gpm) and with no fuel
flow. The data from each of these tests are
presented in Fig. 1.2. Although there is con-
siderable scatter, the data from the different rod
calibration techniques appear to be mutually con-
firmatory, and a rod value of 0.032 Ak/in, obtained
from the data was used throughout the remainder
of the experiment. It should be noted, however,
that the period calibration with no flow is believed
to give the best data, since the inhour equation
is applicable without correction and the resultant
value of the rod is not dependent-on a reactivity
coefficient, ,

The reactor power was first estimated from the
fission chamber counting rate, but attempts were
made to confirm the estimate by operating the

. SECRET
ORNL-LR-DWG 3900A

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0.04
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A °
| . ® A a A A
0.03 %
A X
X
X A
£ A
< 0.02
x o
®
0.0 A FUEL ADDITION
@ PILE PERIOD ZERO FLOW
X PILE PERIOD 48gpm FLOW
. ’
0 2 4 6 8 10 12 14
ROD POSITION (in.) | -
Fig. 1.2. Calibration of Regulating Rod.

12

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":'vented\ by“ operating the pit at s

clean reactor af a low power for a 1-hr period and
then withdrawing a fuel sample and taking a count
of the sample. This calibration was attempted
first at an eshma’red power of 1T w and then at
10 w. The fuel activity from the 1- whr run was
too low for an cccurate ‘count to be made, but
that from the ]0 whr run indicated a power of
13.5 w. The nuclear instrumentation was cali-
brated on the basis of this power determination.
t developed later that almost all the volatile, as
well as the gaseous, fission products were ap-
parently continuously removed from the fuel at
the pump, and consequenfly the actual power was
probably much greater than ‘that mdlcafed by the
fuel sample.

‘Attempts were made to measure the temperature

'coefflcrent when the reactor was subcrmcql and
‘again during the low-power operation. In both

instances it was estabhshed that the coefflctenr
wds hegative, and, in the latter case, it was

~ determined that the magnitude was approximately

5 x 103 Ak/F, A more accurate determination
of the magnitude of the temperature coefficient
was deferred until the high-power runs were made.

As a part of the low-power operation, the shim
rods were calibrated in terms of the regulating
rod. Each of the three shim rods had approxi-
mately 0.15% Ak/in. for most of their 36 in. of
travel.

ngh Power Experiments

The reac’ror was fmally taken to hlgh power
(estlmated at 1 Mw from the nuclear instrumen-
tation) at 6:20 PM, November 9, some six days
after it first beccme cri’ncal

durmg which there were perlods of opercmon ‘with

power levels of ]0 ]00 end 500 kw, and fmcl!y”'
-1 Mw. Power incre ined '
hcnpai‘ed merely by mcreqsmg ’rhe speed o
* heat exchanger.f"j'

  

   

whlch ‘cooled fhe “fuel

 

  

pressures and remotely exhausting the pit gases
to the atmosphere.

 

This power level
was attained after a 30—hr perlod of operohonj“

' _._“_‘_wh[ch recor

 

   
  
  
 

o bt ai ined.
“":i.'of cpprOleOTeIY 2.5 Mw) After about 18 min at

 

msamas'pfie'rsc*

PERIOD ENDING DECEMBER 10, 1954

Once high power was attained, the reactor was
operated at various power levels during the next
several days, as required, to complete the desired
tests, These tests included measurement of the
temperature coefficient of reactivity, a power cali-

 bration from the process instrumentation, and a

determination of the effect of large increases in
reactivity, and they were concluded by a 25-hr
run at full power to determine whether there was
a detectable buildup of xenon.

The temperature coefficient of reactivity was

determined simply by placing the regulating rod

on the flux servo and then increasing the speed
of the blower cooling the fuel. The change of
rod position (converted into reactivity) divided by

. the change in the reactor mean temperature de-

termined the reactor temperature coefficient. The
absolute value of this coefficient was initially
quite large, and it decreased after 2 min to a
relatively constant value of —5.5 x 107°% Ak/°F
Further analysis of the data is under way to
ascertain the precise value of the instantaneous
fuel temperature coefficient, which is, of course,
the most important characteristic affecting the
control of a power reactor. It is certain that this
coefficient was considerably larger than was ex-
pected and that the reactor was exceptionally
stable. ,

In this, as in any potential power reactor, the
reactor behavior as a result of large increases
in either reactivity or power demand is of par-
ticular interest. With a circulating-fuel reactor
operating to produce power, insertion of the safety
rods reduces the reactor mean temperature. The
power level, on the other hcmd is controlled by

fthe rate at which heat is withdrawn from the
J_ACIrcula'rmg fuel ‘The effects of severol of these
'operaflons are :llustrcn‘ed graphlcol]y in Fig. 1.3,

 

) l,the reqcfor mle'r and outlet fuel
emperaiures o er a 100-min ‘test. The 12 thermo-

couple recdmgs on rhe lnle’r ‘and outlet of each
' '\‘h'efstx paralle[ fuel c1rcunrs through 'rhe reactor
_ e}iown in fhls ftgure.. ‘At zero time and with
"'d ‘mean femperurure of ‘about 1320°F,\ the fuel

"_:her“ -

   

shown la’rer,_corresponds to a reoctor power

 

high power the helium blower motor was turned
off, and the temperature gradient was eliminated

13

 

_.blower speed was graduclly increased over"'
min perrod untll a reactor AT of 250‘°F_ was__”
(Thls ’rempercfure difference as w1[|4

 

 

 

 
 

 

 

 

ANP QUARTERLY PROGRESS REPORT

 

1900

Y REDUGED

SPEED

1800

SHIM ROD.
SUBCRITICAL
SHIM RODS

1700

TEMP. INTERLOCK
BLOWER (FULL SPEED)

!690
1500
1400
1300

1200

REAGCTOR INLET AND OUTLET TUBE TEMPERATURES (°F)

1100

1000

ON FUEL HELIUM
{FULL SPEED)

. BEOREI-
ORNL-LR-DWG 3916

 

SPEED TO MAXIMUM||
CRITICAL:

SHIM RODS
FUEL HELIUM
SPEED TO ZERO

AT 72 Mw

SLOWLY INCREASED FUEL

104 100 96 92 88 84 80 76 72 68 64 60 56 52 48 44 40 36 32 28 24 20 6 122 8 4 O

TIME {min)

Fig. 1.3. Power Excursions.

in about 8 min. However, by this time the reactor
mean temperature was 1380°F, and it was there-
fore decreased (to 1350°F) by inserting the shim
rods. The reactor was again brought from low
power to high power in 2 min, and after 8 min at
high power the shim rods were withdrawn to in-
crease the mean temperature. It is most sig-
nificant that during the 2-min interval required to
bring the reactor from a nominal power of 100 kw
to 2.5 Mw the shortest recorded pile period was
only 14 sec. Furthermore, when at high power,
group withdrawal of the shim rods (0.02% Ak/sec)
. resulted in a pile period of only 10 sec., These
““two limiting periods were consistently repro-
ducible. |

Additional insight into the behavior of the
negative temperature coefficient is also afforded
by Fig. 1.3. By time 84 min, the blower was off
and the reactor power was reduced to about
100 kw. The shim rods were then inserted to
make the reactor subcritical, At 86 min, the fuel
helium blower was turned on. The higher density

14

cooled fuel made the reactor critical, and in 2 min
a AT of about 200°F was obtained. At this time,
however, the blower speed was automatically
reduced by a low-temperature signal. The blowers
were subsequently shut off and the shims inserted.

Although the reactor power level had been cali-
brated against the activity count of a fuel sample,
the actual reactor power remained in doubt through-
out the experiment, not only because of uncertainty
regarding the retention of fission products by the
fuel but also because of discrepancies in heat
balances in the process systems, that is, heat
removed from fuel and sodium vs heat picked up
in their water heat dumps. While the causes of
these discrepancies are now being analyzed in
detail, the most reliable estimate of the reactor
power is believed to be that obtained from the
temperature differences and flows in the fuel and
sodium systems. During one typical period of
power operation, the fuel AT of 370°F at 45 gpm
accounted for more than 1.9 Mw in the fuel, while
the sodium AT of 115°F at 150 gpm accounted

148
 

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At

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for more than 0.6 Mw in the sodium, which result
in a total reactor power in excess of 2.5 Mw. The
temperature differences quoted were obtained from
the fuel pipe temperatures to and from the reactor
at the time the reactor inlet and outlet core tube
temperatures (as shown in Fig. 1.3) recorded a
AT of only 250°F. While there are several
possible phenomena which could contribute to

‘these low core tube temperatures (external cooling,

surface layers, radiation, etc.}, there were 10 to
15 thermocouples on the inlet and outlet pipes
which gave consistent readings. The outlet pipe
temperature was at an equilibrium of 1580°F and
was in excess of 1600°F during transients.

The last scheduled experiment to be conducted
on the reactor was a measurement of the xenon
buildup during a 25-hr run at high power. The
amount of xenon buildup was observed by the
amount of regulating rod which had to be with-
drawn in order to maintain a constant power level.
However, during the 25-hr power run the regulating
rod had been withdrawn only 0.3 in.,, or one-
thirtieth the amount calculated on the assumption
that the xenon would remain in the fuel.

The scheduled tests were completed by 8:00 AM,
November 12. The reactor operation was then
demonstrated for all those who attended the ANP
Information Meeting on Friday, November 12.
During the 12-hr period of 8:00 AM to 8:00 PM,

the reactor power was cycled 21 times. The

resulting temperature cycling was probably as

severe as that to which an aircraft reactor would
be subjected. At 8:00 PM, November 12, with
the scheduled experimental program completed and
over 100- Mwhr of operation logged, the reactor
was made subcritical, but circulation of the fuel
and sodium was contmued The followmg morning,

the’ fuel qnd ’rhen 'rhe sodmm were. dumped lnfo '
, thelr respecflve dump tunks. ' '

Maihemutlcal Anu|y5|s of Approach to Crmcul L

W. E Klnney

Alrcraff Reoctor Engmeermg DIVISlOn o

When 'rhe ARE ‘was brought to critical by suc-f ”
cessive fuel addlflons, it was observed ‘that the‘_ '
~ usual plo'f of [] - (]/multlpllcohon constant)] vs

“granium concen'rrahon lncreased at flrs'r very
' rapldly, but when ’rhe curve go’r close to, l the_.

is observed in many reactors, but the ARE ex-
hibited the effect to an unusual degree. In order

PERIOD ENDING DECEMBER 10, 1954

to explain this, the ORACLE three-group, three-
region code was modified so that flux shapes at
successive fuel additions could be calculated.
Group constants were obtained by flux weighting
with fluxes from an Eyewash calculation on the

ARE.

Figure 1.4 shows the space distribution of the
thermal flux for no fuel and for runs 2 through 6.
The effect of the reflector as fission neutrons
become more numerous can be seen. Figure 1.5
compares experimental and calculated startup
curves for the fission chambers which were
located in the reflector as indicated in Fig. 1.4.
In the caleulation,

where CR is the counting rate of the fission

chamber, ;. is the fission cross section for group
i

i, and ¢ is the group 7 flux. For the ARE startup

CR

e

where m is the multiplication constant, CR. is
the counting rate on run j, and CR is the counting
rate with no fuel. |t seems, then, that the unex-
pected, rapid initial rise in the counting rate of
the fission chamber and the {1 ~ (1/m)] curve is
due not only to the general rise in flux level but
also to the formation of the thermal-flux maximum
near the fission chambers. Once the shape of
spatial distribution of the thermal neutron flux
is set up, the fission chambers register only the
general increase in flux level and the count rate
increases slowly.

OPERATION OF ARE PUMPS
W ‘G. Cobb A. G. Grindell
... W.R. Huntley

Alrcraft Reactcr anmeermg Division

The operation of fhe ARE pumps durmg the

. prenuclear and nuci’ear phases of the experiment
‘can be regorded as successful A ’rotc] of 635 hr
of operatlon at h:gh temperature was accumulated ‘

with the sodlum sys'rem pump and 462 hr with the

fuel system pump The latter total includes 169 hr
~with fuel carrier, 220 hr with eritical fuel bu’r no
“heat removal, and 73 hr with heat removal,

The preoperation check of the ARE pumps
consisted in a functional check of the seal gas-

15

 

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ANP QUARTERLY PROGRESS REPORT

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RADIUS (em)

Fig. 1.4. ARE Thermal Flux vs Radius.

balancing systems, installation of a vacuum sink
on the reactor-fuel inlet-pressure-transmitter pilot
valve to protect the pressure transmitter during
the vacuum fill of the fuel system, installation of
a gas nozzle antiplugging device on the fuel pump
tank to reduce the zirconium fluoride vapor-
condensate plugging problem, and a functional
check and setting of low-flow alarms on water
and lubricating oil systems.

After about 24 hr of hot prenuclear operation,
some noises from the bearing housing of the main

,'--'ff_?_____fu'eri pump were noted. This disturbance was first
““heard on the crystal noise pickups. After many

increases and decrease. in intensity, the noise

finally subsided, and the bearing housing per-

formed quietly during the sealed pit operation.
No detectable increase in drive-motor power oc-
curred during the time the noises were heord.

16

Durin.g the filling of the fuel system 'cmd‘s‘ub':-'fi.

sequent nuclear operation, there was some leakage
from the gas space of the main fuel pump tank.
This pump tank served as a fluid expansion and

degassing chamber, as well as the pump sump

tank. The gas leakage may have been associated
with any or all compression joints (spark plug
probes, Swagelok fittings), with any welded gas
flange seal, or with the two compression joints
in the rotary element,

The water and the lubricating oil auxiliary
systems on all four pumps (including two standby
pumps) functioned without incident during the
entire experiment. The seal gas-balancing systems
on the pumps also functioned without incident.

The liquid level indicating probes on the two
sodium pumps and on the main fuel pump (the
standby fuel pump was isolated from the fuel

 

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PERIOD ENDING DECEMBER 10, 1954

ORNL—~LR—DWG 4451

 

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1.0 — Rzt el
/__—_—-—f:" — T
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U2 CONCENTRATION (ib/ft°)

Fig. 1.5. {1 - (l‘/m)] vs U23% Concentration for ARE.

system) functioned satisfactorily., The maximum
level probe in the main sodium pump tank shorted
occasionally, but it subsequently cleared. Appli-
cation of the *‘short burner’’ (high current source
to separate accumulated condensate from the
probe) had no effect., The maximum operating
level probe in the main fuel pump tank shorted
several times during early stages of hot operation.

Each time, the short was cleared instantaneously

by use of fhe short burner.

After approxlmafe!y'

five such occurrences, the circuit was arranged

for continuous burning, cnd no furfher mdlcahon'

of shortmg appeared
'The pump power-’rransmmmg V-belts functioned

‘sahsfocfonly durmg the entire experiment. The
effect of radiation damage upon the belts may be

determined upon inspection.

The power traces indicated good operation of

the pumps undér load. The voltage regulcmon

of the pump motor supply (an’a-c~d-c motor gener-:
ator set) became somewhat unsteady with passage’

of time. This unsteadiness of voltage regulchon
could be seen in pump speed and motor current
indications.

The pump tachometer generators performed well
after some initial troubles were corrected. The
tachometer generator on the main fuel and main
sodium pumps failed during the prenuclear hot
operation. These failures, it is believed, can
be traced to improper fit-up of the generator shaft
into inner races of ball bearings and, possibly,
to excessive tension on the V-belt.

Indications were that the ambient temperature
was approximately 175°F in the pump region
during the nuclear and power runs. This is the
highest ambient temperature fo which ARE pumps
have been subjected, but the rotary elements

- functioned satisfactorily at this temperature. The

lubricating oil and the water temperatures were
maintained af about the values experienced during
hot shakedown development and acceptance

testing. -
- The vapor trap installed in the main fuel pump of
the ARE system was used successfully during the

enrichment procedure. Several hundred cubic feet
of gas was vented through it. The heated line
and trap were still open ot the termination of the
experiment. A report on the vapor trap design

23

’\_ !»“ Ly 4«?. ]7

N

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* ANP QUARTERLY PROGRESS REPORT

and operation is being prepared. Modifications of
this vapor trap design could be used on other
fused salt systems in whuch reliable gas con-
nectlon is imporfcn'f '

LOADlNG OF THE ARE

e  _ " G. J. Nessle |
. C.M. Blood J. E. Eorgan
UL P Boody N. V. Smith
C.R. Croft J. Truitt

Mcn‘erlais Chemlstry DIVISIon. '

R Wiley |

A:rérafi Reactor Englneermg Division

 
   
 
 
   

Fmal prepcranons ‘and installations for fher

'Ioadmg of the ARE were completed during the
month of October.. Transfer of the barren carrier

(NaF ZrF 50-50 mole %) into the ARE fill tank

. No. 2 was completed in approximately 43 hr

w:fhouf mmdeni’ Approximately 2750 Ib of carrier
| looded into the fill tank, and sufficient
: maferta[ was on hand to fill the fuel system to
its operating liquid level.

The enriched fuel storage tank had previously
been loaded with 60 Ib of carrier material for
practice injections into the reactor. Tests indi-
cated that the semiautomatic enrichment system
was inadequately heated, and therefore equipment
for addition of portions of the fuel concentrate
(Na,UF,) directly to the circulating fuel pump
was designed and assembled.

In order to provide adequate means for ap-
proaching criticality safely, some of the original
30-1b batches of concentrate were reduced to
10-Ib batches, and, similarly, to provide accurate
measurements for calibration of the control rods
after criticality, 12 small batches of approximately
0.5-1b size were prepared from a 30-1b batch.

The injection system consisted of a main con-
centrate’ can (containing 30, 10, or 0.5 lb as
required) connected to an intermediate transfer
can by a resistance-heated 1/4-in. Inconel tube.
The intermediate transfer can was designed so
that it could be filled to a predetermined level
(approximately 5.5 b) when the 10- or 30-Ib
batches were in position to be transferred. The
0.5-1b batches were transferred through a smaller
mfermed:ate can, which contained the entire small
batch each time of transfer. The intermediate can
was, in f't._ayrn','connec’red to the fuel pump. The
main transfer line to the pump was heated by
“calrod units, WHi_le the nozzle within the pump

s

was heated by resistance heating. The close
tolerances of the injection nozzle resulted in an
inherent tendency to short out and cause the
nozzle tip to be cold. " This, coupled with the
drainage of high-melting-point material from the
fong transfer line, always presented the possi-
bility of a frozen-fluoride plug.

The first 30 b of concentrate was mlected into
the fuel pump without apparent difficulty. After
this injection, however, difficulties in the form

~of line plugs and venting troubles frequently
occurred.

After 48 hr, most of these problems
were under control, and loading of the enriched
concentrate proceeded smoothly. The reactor was
brought to criticality, and small injections of
0.5-1b batches were made for the control rod

“calibrations. At the completion of the final small

addition, the transfer line broke and simultane-
ously plugged ot the injection nozzle. Efforts to
repair the damage were fruitless, and the final
injection of 30 Ib of the concentrate to allow ARE
operation at power was made through the sample
line into the pump.

ANALYSIS OF LEAK IN SODIUM SYSTEM

E. E. Hoffman W. H, Cock
C. F. Leitten
Metallurgy Division

The sodium system of the ARE was filled
initially on September 26, 1954, but the sodium
had to be dumped on September 27 because of a
leak at a tube bend in the sodium purification
system. The section of austenitic stainless steel
pipe in which the leak occurred was located just
ahead of the filter traps. The system had been
filled with sodium for 37 hr and had reached a
maximum temperature of approximately 1150°F
before the leak was detected. A portion of the
section removed and the location of the leak are
shown in Fig. 1.6. The leak occurred where a
metal thermocouple protection tube had been at-
tached to the stainless steel pipe by Heliare
welding. The attack around the weld was due to
the formation of sodium oxide as the sodium
leaked through the weld. The inside of the pipe
at the weld may be seen in Fig. 1.7. An ex-
cessive amount of penetration was obtained during
the welding operation. The large weld nugget
cracked during cooling after welding. A cross
section of the weld crack is shown in Fig. 1.8.
Examination of the pipe in this section revealed

-AY
i ge 2 de i docpd Rie . R ohl: Kol o iy i L Cojo

= i 0 2

 

i

W

o ¥

o

-p %

E T

T Fig.‘

 
  
    
 
 
 
  
 
  
 
 
 
 
 
  
 
  
     

 

o

     

. Fig. 1.7.

fweld 12X, Redueed 27%.

 

 

Where Thermocouple Well Had Been Attached.

B UNCLASSIFIED
Y-13310

Inside View of Crack in Thermocoup!e‘

PERIOD ENDING DECEMBER 10, 1954

UNCLASSIFIED
Y-13298
(INSERT Y-13309)

1.6. Sect'ion' of Tube Bend Showing Location of Sodium Leak in Type 347 Stainless Steel Pipe

Inset shows location of leak at higher magnification.

 

“a variation in wall thickness from 76 to 135 mils,
and the spectrographic examination revealed that
two pieces of pipe in this section were type 347
stainless steel, while the other two were type 316
stainless _steel.  Chemical analyses of three
sodium samples taken from this section indicated
an 'average sulfur conte of 31 ppm. The actual
cause of the exce;suve penetrahon can probably
be aflnbufed to a lack of knowledge on the
we!der s part of ‘the vqnahons in pipe wall 'rhlck-
_ nesses m thts sectlon.ww

 

 

PRELIMINARY ESTIMATES OF CORROSION
IN ARE

W, R Gfi.mes DA R Cfiheo

Materials Chemlstry DlVlSlon

 

It was possible to wnhdrow' samples of the ARE
fuel for analysis during operation of the reactor

19

5

.,

b
st
Ty
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4
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1

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ANP QUARTERLY PROGRESS REPORT

 

B UNCLASSIFIED)
3 y-3418 |

 

 

 

 

 

 

Fig. 1.8. Cross Section Through Thermocouple
Weld Crack.

below criticality or at very low power levels.
Samples were withdrawn by pressurizing the pump
bowl sufficiently to cause flow through a heated
Inconel tube into a sampling device, which was
constructed to allow line flushing before col-
lection of the sample.

Analyses of the samples were carried out by

the ANP Analytical Chemistry Group end are
shown in Table 1.1. As may be seen, the
chromium content of the fuel was essentially
constant at 100 ppm after 155 hr of circulation
of the fuel carrier through the reactor. If corrosion
of the Inconel is assumed to have been uniform
over the surface of the system — and since it was
operating isothermally during this period, such
an assumption is reasonable — the chremium con-
centration of 100 ppm would represent about 0.5

20

TABLE 1.1. ARE FUEL SAMPLE ANALYSES

 

Hours After

 

Uranium Chromium
Filling with
Barren Carrier (wt %) (epm)

19 81

60 90
110 102
155 100
157 1.84 150
178 3.45 190
182 5.43 200
205 9.54 205
242 12.21 300
246 12,27 320
268 12.54 378
286 12.59 420
307 13.59 445

 

 

mil of attack on the Inconel.

As was anticipated, the chromium content of
the melt was found to increase a few hours after
each addition of fuel concentrate (Na,UF ;). By
the time the fing! sample was taken, after 307 hr
of molten fluoride circulation, the chromium con-
tent was found to be 445 ppm. Again, if a uniform
rate of attack of the Inconel is assumed and if
the removal of circulated material from the system
during this time (to maintain desired pump liquid
level) is taken into account, this chromium content
would represent about 5 mils of attack on the
Inconel system. I|f the data in Table 1.1 are
plotted as chromium confent vs time, it is seen
that the chromium content had begun to level off
by the time the final sample was taken. A pro-
jection of the slope of the chromium _cur-ve- indi-
cates a maximum chromium concentration of about
600 ppm after several hundred hours of operation.

There is no doubt that mass transfer of chromium
metal began soon after power operation started,
since there were then large temperature differ-
entials across portions of the system. Sampling
wa_s'not possible at this time and could not, of
course, have given information on mass transfer.
Heat transfer characteristics of the ARE did not
appear to have changed during its 150 Mwhr of
operation, and so whatever mass transfer occurred
was unimportant. Details of mass transfer will
be studied when sectioning of the equipment
becomes possible.

-
 

R b il

 

i

- r¥

C» 1954

PERIOD ENDING DECEMBER 10, 1954

2. REFLECTOR-MODERATED REACTOR

wr

A. P. Fraas

Aircraft Reactor Engineering Division

-~

Analytical studies, layout work, and detail de-
sign have proceeded on both the reactor assembly
- and the installation for the Aircraft Reactor Test
(ART), formerly the Circulating-Fuel Reactor
Experiment (CFRE). Work has also continued on
the component development tests outlined in the
previous report. | Reports have been completed on
the high-power-density beryllium thermal stress
test and the first fuel-to-NaK heat exchanger
test.2:3  Specifications have been completed for
radiators to be used as heat dumps in heat ex-
changer tests, It is expected that essentially the
same specifications can be used for procurement
of the radiators for the ART heat dumps and that
endurance test experience gained in the course
of these small heat exchanger component develop-
‘ment tests will establish the reliability of the
product of at least one vendor. |

EXPANSION-TANK AND XE NON-
REMOVAL SYSTEM

#

, G. Samuels |
Aircraft Reactor Engineering Division

W. Lowen, Consultant

¥

Recent work on the expansion-tank, fuel-pump,
and xenon-removal system has been directed
primarily toward the development of a hydraulic
circuit that performs all the various functions re-
quired of the system. The basic components and
their principal functions were described in an

earlier report4 and may be summarized as fo“ows
(1) an orifice for bypassing a mefered qmoun'r of”v
fuel from the mcnn system for processmg, @ a

- 'swul chamber for producmg a large gas- ]quId

  
  

ffec'r the prompt releuse of xenon

 

 
 
    

£ lA P. Fraas, ANP Quar. Prog. Rep. Sept. 10, 1954,

ORNL-1771, p 25.

 

m'Beryllzum Tesi No. 1, ORNL CF 54 10-106 (Oct. 25,

 

“Stress Study of the Intermediate Heat Exchanger Test,
ORNL CF-54-11-69 {to be issued).
‘4R, W. Bussard and A. P. Fraas, ANP Quar. Prog.
Rep. Dec. 10, 1953, ORNL-1649, p 39.

9

 

interface area by v:olem‘ cgl'rahon “and gas en'rrcnn-

 
 
 

for the main fuel system with a liquid surface that
is stable for all attitudes (in flight); (3) a centri-
fuge cup integrally mounted on the back of each
fuel pump impeller for degassing the processed
fuel, pumping the processed fuel back into the
main system, and providing a seal for the fuel
pump.

Several Lucite models that differ in component
design and circuit arrangement have been built and
tested, and they have led to the design illustrated
in Fig. 2.1. In this model the swirl chamber is
mounted between the two fuel pumps and raised
above the centrifuge cups to assure their priming
under starting conditions.  The swirl in the
chamber is produced in two ways: by nozzleslo-
cated in the swirl chamber floor and by swirl
pumps which are needed to maintain a high swirl
velocity when the fuel level rises above the 25%
full condition owing to thermal expansion in the
main system. The nozzles and swirl pumps are
arranged to deliver high-velocity fuel jets tangenti-
ally at the periphery of the swirl chamber to give
good agitation and to assure a strong centrifugal
field and a stable free surface.

The nozzles serve both to meter the bleed flow
through the expansion tank and to control the fuel
pump suction pressure relative to the swirl chamber
gas pressure. This important function is perhaps
more evident in the schematic circuit analog pre-
sented in Fig. 2.2. By tracing the bleed flow, it
can be seen that the pressure drop across the

_ nozzles approximates the heat exchanger resis-
. tance.

Consequently, the fuel pump suction pres-
sure is maintained near the helium pressure
exisfing in the swirl chamber.

..The configuration of nozzles and parts is so
proportioned as to oversupply the centrifuge fuel

Aflow demand in_order to minimize the possibility
of the cen’rrlfuge becommg emp’ry, “especially
during speed transients,

o centrifuge is dlverfed by the flow-split stator to
R, W, BUSSGrd und R E MccPherson Tbermal Stresses

The excess fuel to the

the swirl pump, which, as previously described,
ejects this fuel back into the sw:rl chamber. Thls_

"arrangemenf of mcorporcn‘mg a sec0ndcry circuit

within the bleed circuit contributes toward stabi-
lizing the three free surfaces inherent in this
layout.

m

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y paven
i ,' ) P
i‘ ,’A STy T
£ ;
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¥

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T o6 &
ST

21

 

 
4

   

@

SWIRL CHAMBER

SERVES AS EXPANSION.TANK
PRODUCES GAS ENTRAINMENT.
PROVIDES NEGATIVE 6 STABILITY

       
 
    
 
 
     
  
 
 
    

SINGLE SHAFT —OVERHUNG ‘/{
1

 

 

" FUEL UNDER
HELIUM PRESSURE

UNIT ASSEMBL:/
REMOVABLE

L

7
////////

- FUEL PUMP

- R

-y LARGE RADIAL AND D'SCHAFSEOF;,RS.ESSU E
mend AXIAL CLEARANCES

A A

 

 

 

Fig. 2.1.

Expansion-Tank and Xenon-Removal System,

X

“SECRET
ORNL-LR-DWG 4497

SLINGER [MPELLER
SEALS AGAINST PRESSURE
ENABLES INVERTED CPERATION

)

AI———— SWIRL PUMP

PROVIDES ADDITIONAL ENERGY
FOR SWIRL AGITATION

SERVES AS GAS TRANSFER PUMP
ASSURES VARIABLE SPEED STABILITY

 

| T——— FLOW-SPLIT STATOR
PRIMES CENTRIFUGE

ASSISTS IN KEEPING CENTRIFUGE
SUPPLIED WITH FUEL

CENTRIFUGE SEAL IMPELLER
MAINTAINS PRESSURE IN SYSTEM

 

PREVENTS CENTRIFUGE BY-PASS

q_,jjk CENTRIFUGE

REMOVES BUBBLES

ACTS AS A TRANSFER PUMP
OF THE BLEED FLOW

1dOdIY SSIYOOYd ATIFLIVNO dNV
 

PERIOD ENDING DECEMBER 10, 1954

HELIUM PRESSURE

sEewET
ORNL—LR~DWG 4498

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

  
  

£
| ]
. O e |
: Ap~60psi
VSLJNGER
& IMPELLER -BUCKING 777 T I IMPELLER
LEAKAGE
SWIRL
PUMP
= | -
Ap~30psi > 2 |
' 2|9
XENON- REMOVAL SYSTEM < o e -
BLEED CIRCUIT -/ >
< fi—
\SWIRL STATOR
* s )] x% s CHAMBER 1
SEAL FLOW T &
BUCKING PUMP o313 & - NOZZLES
5* o ™ CENTRIFUGE
229 SEAL
.J IMPELI_ER
. W@ CENTRIFUGE
FUEL PUMP
*
 REACTOR CORE <
MAIN FUEL CIRCUIT™

 

23

A R i i

 
Cold, ;e e

 

 

i

ANP QUARTERLY PROGRESS REPORT

The addlhon of 'rhe swirl pump made it neces-
sary, however, to replace the sllnger ring, previ-
ously employed, with a slinger impeller in order
to buck any leakage from the swirl pump. Although
the complete assembly has not yet been tested in

~an inverted position, the slinger impeller seems to

be very effective. A similar arrangement is now

‘being tried as a means of controlling the leakage

from the main pump discharge through the clearance

_ between 'rhe impeller and the casing past the

cem‘rlfuge. More conventional sealing arrange-
ments were found to be ineffective because of the
abhb'rmoily 'ldrge‘ radial clearances (0.060 in.)
specified in the design objectives. Initial tests
currently in progress with the centrifuge seal
impeller have been quite promlsmg Although no
detailed studies of the off-gas system and fuel
pill addition sysfem have been made, the changes
in fhe xenon—removal system will make it neces-
sary that the off-gas system connection and the
fuel pill addition be made through the swirl
chamber rather than through the pump housing as
illustrated previously.>

CONTROL ROD DESIGN CONSIDERATIONS

The design of the ART has been predicated on
the belief that no fast-moving control rods will be
required because of the inherent stability of the
reactor that will arise from the strong negative
temperature coefficient of the circulating fuel.
It is further expected that only about 1% Ak/k
will be required to compensate for xenon poisoning,
since it is anticipated that with the bypass ex-
pansion tank and gas-scrubbing system planned,

the xenon will be removed almost as rapidly as it

is formed. |t is expected that compensation for
burnup and the accumulation of fission-product
poisons can be effected by adding pills of solid
fuel having a high percentage of U235, Thus the
principal function of the control rod will be to
control the mean temperature level of the reactor
fuel.  Since the temperature coefficient will
probably be approximately 5 x 10=3 Ak/k per °C
and since a temperature level variation of the
order of 150°C will probably be required, provision
of a Ak/k of 1% in the control rod for temperature

level control should be more "rh‘an adequate. ln

reviewing the various requirements for control,
thus follows that a control rod having a total k e ff
of 2% should prove to'be adequate. This reactivity

 

 

Sibid., Fig. 3.4, p 40.

24

canbe readily incorporated in a single slow-moving
control rod. [t would seem entirely adequate to
move this control rod at a rate of 1% Ak/k per
minute, a slow and conservative value.

The size and stroke of the rod are important
considerations, Two items of experimental data
are availoble fo indicate the size of the rod re-
quired. In the course of the second reflector-
moderated reactor critical experiment (that making
use of powdered fuel placed inside 1%-in.-square
aluminum tubes), it was found that with an 18-in.-
dia core and a 9%-in.-dia island a control rod
effectiveness of about 0.6% Ak/k was obtained
from a 346-|n.-0D stainless steel tube filled with
powdered boron to a density of about 50% and
inserted in one-half of the critical experiment
assembly. A second indication of the effective-
ness of control rod material is available in that a
piece of gadolinium oxide approximately the
diameter and thickness of a 25-cent piece was in-
serted at the center of the island in the three-
region octahedron critical assembly (without core
shells). The value of this gadolinium oxide wafer
was found to be approximately 0.2% Ak/k. Some
additional data are also available from multigroup
calculations. All these data seem to indicate that
a rod approximately 3’/8 to 1/2 in. in diameter having
a stroke of approximately 20 in., about 71/2 in, of
which would be below the equator, should prove
adequate to give a Ak/k of 2%.

The heat generation to be expected in the con-
trol rod will establish the amount of cooling that
will be required and, to a large degree, much of
the detail design of the rod. The heat generation,
in turn, is dependent, in part, on the number of
neutrons absorbed in the rod at full power. Molti-
group calculations indicate that 1.3% of the
neutrons produced will be absorbed in a control
rod having a Ak/k of 1%, if it is assumed that the
average effectiveness of the control rod is half
that of the portion at the center of the island.
Thus a control rod having an effectiveness of 2%
Ak/k would, when in the ““full in’’ position, absorb
2.6% of the neutrons. [f a value of 3 Mev per
neutron absorbed is assumed, the energy associated
with neutron absorption in the control rod would
amount t0 0.039% of the total power generated in the
reactor. This would total 24 kw for a 60-Mw power
level. The control rod cooling requirement would
be a minimum if gadolinium rather than boron were

used as the absorbing material in the rod. This

3
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Hcnn ool e 1, < el | v BmBe

 

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_amount to about 20 w/emS,

' quire'd.

 
 
  
  
 
   
  

  

g frangi

would mean that most of the energy associated
with the hard capture gammas emitted from the
gadolinium would appear as heat in the surrounding
moderafi-ng material rather than in the gadolinium.
Thus a gadolinium rod would have the advantage
'rhcn‘ relatively little of the heat associated with
neutron captures in the rod would appear in the

red, and therefore little provision for cooling

would be required. On the other hand, if boron
were used, the very short range ofpho emitted
from neutron captures in boron would cause
virtually all the energy associated with the
neutron capture to appear in the boron.

‘The principal heating in a gadolinium rod would
be that induced by gammas from the fuel region.
For a 60-Mw reactor it appears that the power
~density in the gadolmlum from this source would
Estimates indicate
that about 20 cm® of gadolinium would be re-
The total power generation in the rod
from gammas from the fuel region would amount to
opprox1mate|y 400 w, or about one-sixtieth the
power ‘generation in a boron rod. It should be
noted that gadolinium oxide is available at a cost

of about $4000/1b and that it can be fabricated

readily by using conventional ceramic techniques.

Cooling of the rod presents quite a number of
problems. If the rod were immersed in flowing
sodium, cooling would present no problem. How-
ever, the control rod octuotmg mechanlsm would
have to operate, to some degree at least, in
sodium, and self-welding might present some
difficult problems. If the rod were placed in a
thimble in the center of the core and an atmosphere

of helium were mamtamed in the 1h:mb|e,7|f mlghfr
be posmble to effect coohng through rodlahop‘_or:;i S

conduchon from the rod materlal to the sodiv
cooled wo!ls of 'rhe thlmble._”

crrongemehf;mlght prove to be scflSfccfory WIth o

 

'umfomcle or ‘f' ‘Haf u

might be circulated over the rod to cool it, in
which case the outboard end of the control rod

 

  
  

Thus_‘ sor'r of an
1 " Valves and thetr consequent enfralnment in la'rer
_”\fllhng and drommg operohons “and’ por'rly to mini-
n"'m:zeu the _amount ‘of raduoochvfiy ‘that might be

 

»Jilnclude’ prov151on for accurafe m
quantity of fluid in the “drain tank at all times

PERIOD ENDING DECEMBER 10, 1954

mechanism could include a heat exchanger so that
the heat would be removed from the circulating
helium and transmitted to the shield water, Un-
fortunately, such a system would be dependent on
the satisfactory operation of quite a number of
moving parts, always a likely source of trouble.

FILL AND DRAIN SYSTEM

A good, reliable, relatively simple fill-and-drain
system incorporating a quick disconnect coupling
suitable for remote operation with high-temperature
liquid systems is clearly needed. In the design of
the ARE, fixed tanks withremotely operated valves
were used in order to avoid a remotely aperated
coupling. This approach led to a quite complicated
collection of tanks, plumbing, and valves, and to
somewhat clumsy arrangements for system drainage.

The requirements of such a system are fairly
straightforward. In the first place, a good coupling
that can be operated remotely and that will be
dependably pressure-tight is required.  This
coupling must be relatively insensitive to align-
ment both as to concentricity and to parallelism
of the axes of the mating flanges or surfaces. A
reliable pressure-tight valve on either side of the
coupling must be provided so that, after the
coupling has been made, the space between the
valves can be evacuated and then purged with
helium. The valves on either side of the coupling
can then be opened and the filling or draining
operation carried out. Upon completion of the
operation, the upper valve can be closed and the
space between the valves cleared of liquid by
sevérol short blasts of helium. The lower valve
can then be closed and the coupling broken.
Scavengmg the spoce befween the volves is very
mportant, partly to mlmmlze the possnbihty of
ox:dotlon,of droplefs of mcn‘erml betw<=en the

mitted fo ’rh'e All this implies that

 

"'rhe volume ‘and surface dreu be’rween the two

volves should‘bo_mlmmlzed ‘and fhat the geomefry
k fa "I'tote drcunage and to

  

,surve}nen'r of fhe‘

 

during either the filling or the training operation.
This is particularly important in connection with

STy e 25
T £

 

 

 
 

 

 

 

"ANP QUARTERLY PROGRESS REPORT

reactor fuel systems because it is essential that
the exact amount of fuel in the reactor be known

- at all times. One good way of measuring the

quantity bf liquid in the tank would be to support

the tank on two Hagan Thrustorq units. This

should give an instantaneous measure of the weight
of the tank to an accuracy of 1% of full-scale
reading.

Design work is under way on a system to fill the
above-specified requirements. Complete tests of
mockups of proposed designs will be made before
the final design is chosen.

DESIGN PHYSICS
W. K. Ergen

Aircraft Reactor Engineering Division

' The comparison between multigroup calculations
and critical ‘experiments on three-region reflector-
moderated reactors has shown a certain amount of
discrepancy, which must be overcome by adjusting
the constants used in the caleulations. Investi-
gations were made with the ultimate aim of ob-
taining simpler expressions for the critical mass
and power distribution for reflector-moderated
reactors. This was done in the hope that these
simpler expressions could be fitted to the experi-
ments by adjusting of constants and that this
fitting might be as satisfactory as the one employed
for the multigroup calculations. The three-group
ORACLE calculations of the approach to criti-
cality of the ARE (see Sec. 1) gave a satisfactory
explanation of the strange results observed during
the experiment. An example of the calculations
performed for the ART is the following calculation
of the activation of the Inconel core shells.

Activation of the Inconel Core Shells

"H. W. Bertini

Aircraft Reactor Engineering Division

An attempt was made to estimate the activity of
the Inconel core shells on the inside of a 60-Mw
reflector-moderated reactor after the reactor had
been operating for 1000 hr. The activity will
come from two main sources: the activity of the
cobalt in the Inconel and the activity of the fission

26

fragments that strike the Inconel and remain there.

The method described below for obtaining the
estimate gives an order of magnitude of the total
activity. Reasonable approximations are used to
simplify the calculations, and, in this way, many
of the details that a more complete analysis would
require are circumvented, and yet the required
accuracy of the calculations is maintained.

To obtain the activity due to the cobalt, the
total flux at the surface of the Inconel is obtained,
This flux is multiplied by the product of the
microscopic cross section of cobalt times the total
number of cobalt atoms in the core shells. This
gives the total number of cobalt atoms being
activated per second.  Multiplication of this
number by 1000 hr gives the total number of cobalt
atoms activated during the reactor lifetime. Since
the half life of cobalt is so long (5 years) com-
pared with the reactor lifetime, the decay of the
excited cobalt atoms can be assumed to be con-
stant, and therefore the disintegration rate can be
taken as AN, where A is the decay constant of
cobalt, and N is the total number of cobalt atoms
activated during the lifetime of the reactor.

The following values were used for the calcula-
tions:

Reactor power = 6 X 107 w

Total mass of U235 in the core = 2.5 x 104 g
Density of Inconel =8 g/cm3

Weight per cent of cobalt in the Inconel = 0.1%
Half life of cobalt =5 years

Average flux at the surface of the reactor,

(’és =2X (]Sreacfor
1 curie =3 x 1010 decays/sec

0’5‘,J = 550 barns

GSO = 35 barns

The total mass of Inconel in the core shells of
the reactor is approximated by the mass of two
spherical Inconel shells, one with a 15-cm radius
and the other with a 25-cm radius, each shell
being 0.3 em thick. Then,

+

In
 

Wy

PERIOD ENDING DECEMBER 10, 1954

(total fissions/sec)

 

h“& !
ft
N
X

crfu % (total U233 gtoms)

9 6 x 107w x 3 x 1019 (fissions/w-sec)
% .

 

 

6 x 10 x 2.5 x 104
<550 x 10—=24 ~ - * (fissions/neutron.cm?)
| 235
-1 x 10”(neut’rons/cm2»sec)

The foi;al- lr'wmber‘ of cj:o.boh‘ atoms inl'rhe‘: liner is |
th'?at = [8 (g of Inconel/em®) x 4m (252 + 152) (em?) x 0.3 cm | X 10'3 (g of Co/g of Inconel)
'« 6 x 1023 (afom; of Co/g-at. of Co)l + 59 (g of Cp/g-cflr‘° of Co) -
- 2.6 x 1023 -

The activations per second are

Co

= Co
P ¢A x Ntotal A

I

1074 x 2.6 x 1023 x 35 x 10~24

i

1]

9.1 x 104 (activations/sec) .

The total number of cobalt atoms activated is

P x 1000 (hr)

=
0
o,
i

= 9.1 x 109 (activations/sec) x 1000 (hr) x 3.6 x 103 (sec/hr) -
=33 k102, |

The activity of the shells due to _cobdh‘ activation is given by

    
 
  

=A% Nf") .nr(rai;s'iihlzf.e;;j‘l“a'rions/sec) R

 

 

  

;-'-fvr;z:‘,-_-3.,~3'-;--1021-, L

2 (h/day) x 3.6 x 102 (sec/hr) 3 x 101

 

5 (years) x 365 (days/yea

 

27

 

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

In calculating the activity of the fission frag-
ments, use is made of Fig. 2 in ORNL-53,% which
is a curve of the dose from a uranium slug exposed
~ for three years plotted against time after exposure.
The dose is given in mr/hr at 1 meter per watt of
exposure, ‘

Calculations now show that 1 curie of 2-Mev
gammas gives a dose of approximately 1 r/hr at
1 meter. |f it is assumed that the average energy
of the decay gammas from the fission fragments
is 2 Mev (an overestimate), the curve can be used
to obtain an estimate of the curies of activity for
anv time after exposure. However, it is first
necessary to calculate the watts of exposure in
the regions of the reactor close enough to the
Inconel core shells so that the fragments due to
fissions will strike the Inconel. This is done by
finding the range, R, of the fission fragments in
_ the fuel and again approximating the fuel annulus
by a spherical shell of 25 ecm OD and 15 em ID
“to find the total power generated in the regions
adjacent to the Inconel. This power is divided
by 2 to account for the fact that approximately
one-haif the fission fragments generated in this
region will impinge on the shells. Since the curve
was plotted for a slug exposed for three years,
any valves obtained by using this curve will be
overestimates of the fission fragment activity of
a 60-Mw reflector-moderated reactor.

The range of fission fragments is estimated to

be 0.001 ¢m, and the
1

watts of exposure = 6 x 107 w x —

2

 

47 (252 + 152) x 10-3
X

 

The last factor of 2 in the above equation is due
to the flux at the surface of the Inconel being
twice that of the average flux. Some values of
the total activation of the Inconel core shells
obtained by using the above methods are given in
Table 2.1. One conclusion to be drawn from
Table 2.1 is that little would be gained by trying
to obtain special Inconel with especially low
cobalt content; the activity could not be reduced
below that caused by the fission fragments,

TABLE 2.1. ACTIVATION OF INCONEL
CORE SHELLS

 

Activity (curies)

 

 

Days After
Exposure Cobalt Fission Total
' Fragments
10 480 390 870
20 480 260 740
100 480 80 560

 

 

Mathematical Models for Reflector-Moderated
Reactors

L.. T. Anderson, Consultant

The thermal absorption rate has been calculated
for an idealized reactor consisting of a nonmoder-
ating spherical-shell fuel region surrounding a
moderator and surrounded by an infinite moderating
reflector with a fission source of 1 neutron/sec
in the fuel region, The thermal capture cross
section of the fuel region was taken to be infinite,

 

4 x 2 The nonthermal capture cross section of the fuel
3 7 (253 - 159) region was assumed to be zero, and the thermal
and nonthermal absorption cross sections of the
4 moderafor were likewise assumed to be infinitely
= 1.3 x 10%w . .
small, The neutrons slowed down according to age
theory, and the thermal neutrons diffused accordin
$G. Ascoli and O. Sisman, Absorption o{l Radiation df);’ . h d
from an "X Slug by Lead, Flg. 2, p 7, ORNL-53 (Mdy to diffusion t eory.
1948). The absorption rate is, for the source located at
the inner radius of the fuel region,
1 -
2 e i
(1) Ay =— dx ,
7 2 2
0 x R,
1 +{— = 1] xcotx| + cot?x
Ry
28
Doy T 034

W
 

#ea

g

o

where R, is the island radius, R, s the outer
radius of the fuel region, and 7is the thermal age.
Evidently the absorption rate depends only on two
parameters: R,/R, and T/Rf.

For the source located at R,, the result is

PERIOD ENDING DECEMBER 10, 1954

and this curve is also shown in Fig. 2.3, The
““thermal absorption rate’’ plotted is a direct
measure of the neutron leakage — the higher this
rate the lower the leakage.

 

 

 

R, 2
L R - -(T/R%)xz 1+ —1— 1] x cot x
e
(2) A2 = — ——— dx .
7 R2 2 2
0 x R
1 | — ~ 1) xcotx| + cot?x
R?
For a uniform source throughout the fuel region, the result is
1 Rl (R2 + ZR-‘)
(3) A = A, +3 (A; - A,y .

Equations 1, 2, and 3 were evaluated by numerical

integration for R,/R, = 1.67 and 'r/R2 = 0.33,
0.44, and 0.55. The plots are shown in Flg. 2.3.
For the case R, =0, Eq. 1 reduces to

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1/2
-7/ R% 71/
e erfc ,
sEeRET
ORNL-LR-DWG 4499
‘ t/R?
0.222 0333 0.444 0,555 0.666
09
g
ey
w
5
5
e
~ 08 :
[ ® i . )
. \\4 | —SOURCE AT A,
z A L1 UNIFORM SOURCE -
Q - - g
= o “\‘~\ e \
a "\ *
% 0.7 9., 5\ — T o~
= SOURCE AT 7, | P *-
3 ' 2 ~
S [ NO ISLAND T
Pl
" os : L — ,
0.20 024

  

05 om _ ox
o /R,

   
 
 
 

Fig. 2.3. Thermadl Absorpfi.onr Rate in Mathe-
matical Model of ReflectorsModerated Reurciors.

2 2
R] +R1R2+R2

 

PROPOSED ART INSTALLATIONS
A, P, Fraas F. R. McQuilkin

Aircraft Reactor Engineering Division

A study has been made of the various types of
facilities suitable for the operation of the Aircraft
Reactor Test. In addition to the obvious require-

ments for shielding, heat dumps, and auxiliary

equipment, it is essential from the hazards stand-
point that provision be made to contain the
products resulting from a nuclear accident and/or
chemical reaction of all combustibles in the
installation, minimize the likelihood of serious
damage from an explosion caused by sabotage or
bombing, and remove and dispose of volatile
fission products evolved during the course of
operation. With these criteria in mind, a series
of four basic reactor installations was considered;
each differed in some fundamental characteristics
from the others, The four installations are as
follows: (1) an open test unit mounted over a
water-cooled pan at the National Reactor Testing
Station (NRTS), (2) a circular Quonset type of
hemispherical building, (3) a water-walled tank,
and (4) a reactor submerged in a pool of water.

rl‘Each of ’rhe msfalichons mcluded f;ve malor
w"{:m'[oonen'rs “"a shielded reactor assembly, radi-
ators, blowers, a control system, and a fill-and-
drain system. These components would operate

29

 

T T T

 
 

i G

 

ANP QUARTERLY PROGRESS REPORT

as an integroted system, irrespective of the type
of test installation chosen. In all but the fourth
installation, the assembled reactor would be sur-
rounded by an aircraft type of shield of lead and
borated water, and, in all installations, it would
be coupled to heat dumps consisting of banks of
aircraft-type NaK-to-air radiators through which
cooling air would be circulated. Appropriate con-

_frol systems, along with auxiliary shielding, equip-
-ment, and services, would comprise the balance of
- the test installation. A generalized flow sheet for

th:s setup was shown in Fig. 2.1 of the previous

" quarterly progress report.”

All but the first of the above-listed installations
‘would be located in Oak Ridge. The first installa-
tion is illustrated in Fig. 2.4. It was devised to
permit operation of the reactor surrounded by an

© aircraft-type shield with a heat dump on either side
" fo simulate the turbojet engines. Of the installa-

tions considered, it offers the most compact

“arrangement of the equipment with the least amount

of shielding and minimum provision for contain-
ment, This scheme was developed with the thought
that it could be built in Oak Ridge so that all the
welding of high-temperature-liquid piping could be
made, inspected, and pressure tested and some
preliminary testing carried out, probably including
a hot critical experiment, before the unit was
shipped to NRTS, The dimensions of the unit are
such that it would fit on a flat car and comply
with standard railroad side and overhead clearance
regulations. To do this, it would be necessary to
dismount certain elements, such as the pump drive
motors, the blowers, and the blower drive motors.
This could be done easily, since only bolted con-
nections would be involved, The reactor would be

~ set up with a heavy, water-cooled pan beneath it,
- This pan would catch, hoid, and cool the fuel in
‘the event of an accident, A control room would be

built as a unit and shipped to NRTS on a second

“flat car. The contrél room and the reactor would
“probobly be placed a quarter of a mile to a mile

apart, and the two would be coupled by telemetering

. _equipment, The pumps in the layout are shown as

- -being driven by d-c electric motors, but air turbine
 ‘motors would serve equally well if a source of

, " compressed air were available. If the tests were
. _run at NRTS and Air Force, portable, gas-turbine-
S fipé"dir 'com'pressors were used, a compressed air

 

.- 7A, P. Fraas, ANP Quar. Prog. Rep. Sept. 10, 1954,
ORNL-1771, p 21.

30

source might be more easily arranged than a d-c
generator set. ‘ ‘

In examining the NRTS installation design, a
number of points became evident, First, the prob-
lems associated with operating a reactor at NRTS
seem to be rather serious, particularly from the
standpoint of the amount of time that would be
lost in maintaining an operation 2000 miles from
Oak Ridge. The distance would make it particu-
larly difficult to cope with unforseen problems.
Any relatively small difficulty that might arise
would be likely to introduce a major delay if that
difficulty were not forseen. It appears therefore
that an NRTS installation would entail a loss of at
least six months in getting the reactor into opera-
tion, The second major point that developed in
connection with the examination of the design was
that the major hazards appeared to be much less
serious than had originally been presumed. The
very compact installation achieved through careful
design directed toward simulation of a full-scale
aircraft type of power plant led to a very low in-
vestment of sodium and NaK, about one-twentieth
of that required for the KAPL-SIR reactor designed
for the same power level, Further, the use of
circulating fuel with its high negative temperature
coefficient gives a reactor in which a nuclear
explosion seems almost out of the question,

In view of the relatively small amounts of energy
released under conditions of a total reactor tragedy,
designs were prepared for the installation of the
ART in a closed building. The layout shown in
Fig. 2.5 envisions a sort of circular Quonset
building about 200 ft in diameter. The test unit
would be built and operated on the test floor
behind supplementary shielding, and a heavy,
water-cooled pan would be placed beneath the
reactor, This pan would catch, hold, and cool the
fuel in the event of an accident. Such an arrange-
ment would give plenty of floor area in a relatively
inexpensive building that could be sealed to con-
tain any fission products that might be released,
in line with the philosophy underlying the use of
the Hortonsphere at KAPL., A similar type of
building, but shaped as a hemisphere rather than
as the circular Quonset building, might be used
in order to reduce the amount of steel required in
the framing of the building.

The arrangement shown in Fig. 2.6 follows a
quite different philosophy. It provides for con-
taining the reactor assembly within a pressure

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ORNL-LR-DWG 1391

FUEL PUMP MOTOR
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THERMAL INSULATION

NaK PUMP, HEADER TANK AND
60 H.P. DRIVE MQTOR

 

THERMALLY INSULATED DOORS
s =

 

 

 

 

AIR INLET
SCREEN

 

 

 

 

 

 

 

 

 

 

 

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NaK TO RADIATORS

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FEET NaK RADIATORS

 

SUPPORT FRAME
26 x 7

 

 

 

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BURNER FOR PRE-HEATING

THERMALLY N
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Fig. 2.4. Open Reactor Test Unit.

ye6l ‘0l ¥39W3D30 ONIANI AOId3d

 

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LONGITUDINAL SECTION

 

HORIZONTAL SECTION

 

 

 

 

 

 

ELEVATION

T Fig. 2.5. Circular Quonset Building.
 

w)

PERIOD ENDING DECEMBER 10, 1954

 

Fig. 2.6. Perspective of Proposed ART Installation in Addition to Present ARE Building Showing
Double-Walled Tank.

vessel submerged in water, while the heat dump temperature of the inner tank should not exceed
equipment is located outside, but nearby, A the water temperature by more than 40°F, even if
double-walled tank from 20 to 30 ft in diameter and the bottom of the inner tank were covered with

 

2T e N T L

   

 

 

   
   

from 26 to 36 ft fbl[ is 'enw's.loned "The s space | molten material,  In the event “of. _sorrie'rhing so
between the wa[ls wou‘ld be offlthe order of 18 in,  serious as a meltdown, it seems likely that a part
wcmd would be f:lled w:th wcter. The mner wall ~ of the water shield surrounding the reactor pres-

sure shell would be ruptured and that water would
tend to collect in the bottom of the tank with the
fuel - This water would remove heat from the fuel,
‘vaporize, and then condense on the tank walls
_ so that the atmosphere within the tank would
the heat gwen off by 'rhe decay gamma achvnty ' probably not get hotter than around 300°F,

     

"':would\ Ibe carried off by vaporization ‘of the water ~ The bolt flanges shown in Fig. 2.6 near the tops

       

 

be-fween (fhe tank walls, Since the rate of surface-  of the tanks could be placed about 3 ft above floor

            

      

to- yvatel_' hect_fransfer’under boiling condmon's. “Jevel. The reactor installation and preliminary
would be exceedingly high (on the order of 400,000  shakedown testing could be carried out with the
Btu/hrft2) and since the thermal conductivity of tops of the tanks removed. Since the reactor shield
the fluoride fuel is relatively low, the water-side would be quite effective, the space inside the

33

 

   

 

 

 

 
 

 

 

 

ANP QUARTERLY PROGRESS REPORT

tank wou.ld be shielded fairly well, and it would

be possible for aman to enter the inner tank through '

a manhole at the top for inspection or repair work,
even after the tank had been closed and the ve-

“actor had been run at high power. The water

capacity of the spcice between the tanks, together
~with the water in the reservoir above the inner
tank (apprommotely 10 ft deep), would be of the
order of 110,000 gal, Boiling of this water would
suffice to carry off all the heat generated by the
fission products for about two days without any
fresh water being supphed to the outer tank. All
the vcrlous wires, plpes, tubes, etc, connected to
the reactor and its auxiliaries would pass through
carefully laid-out junction panels such as the one
shown in Fig, 2.7. Such a panel might be installed
in the wall of the tank shown in Fig, 2.6 with a
pressure-hght gaske’red flanged ;uncnon. The
various thermocouples, power wiring, etc. could be
installed on the reactor assembly in the shop and
fitted with Cannon plugs so that they could be
plugged into the panel in a short period of time
after the reactor assembly had been lowered into
position in the test facility. This should minimize
the amount of assembly work required in the field.

The layout investigated as the fourth proposed
installation would involve placing the reactor in-
side o sort of swimming pool with water-tight
thermal insulation surrounding the pressure shell,
lines, and pumps. The lines could then be brought
out the top of the water tank to the instruments
and the heat dumps, In the event of a severe
accident that resulted in a meltdown, there would
be sufficient heat capacity in the water to absorb
the heat from the fission products for some days
before a seriously large amount of water would
have been boiled out of the pool. Such an arrange-
ment could be enclosed, of course, in an air-tight
building.  This might not be necessary, but it
seems likely that, in the event of an abrupt melt-
down type of failure, the high heat capacity of the
region inside the thermal insulation might put so
~much heat into the water in a very short period of

“time that bubbles would boil violently to the
".surface and disperse entrained fission products,

g 'Aftéf,v‘- perhaps, 10 or 15 min, it seems unlikely

that such entrainment would prove to be a problem,

The unshielded reactor assembly will weigh
“approximately 10,000 Ib, the lead gamma shield
approximately 30,000 Ib, and the water in the
~shield approximately 34,000 1b, The first two of

34

these. ttems could be handled: convemenfly with a

20-ton crane, while the borated water could be
~pumped in after the rubber fonks hud been mstaHed

for the water shield.’ :

An armngement such as the double-walled tank
shown in Fig. 2.6 should prove to be cdequate fo
take care of any accident not involving sqbofcge
or bombing. The same should be true for either
the hemispherical or circular Quonset 'rype of
building. Only the NRTS installation would, be-
cause of the remote location, present a not- too-
serious hazards problem if effectively sabotaged.

To evaluate the merits of each of the installa-
tions considered, an attempt was made to envision
as many accidents as possible that might prove to
be serious during the course of operation of the
ART. The worst natural accident that could be
envisioned would result in a meltdown, The only
source of an explosion that has been envisioned
would be either sabotage or bombing. If the
volatile fission products are removed during the
course of operation, the major hazard to the sur-
rounding area would be from the fission products
that might be dispersed in the course of violent
boiling or from an explosion.

Although in the preliminary hazards analysis
consideration was given to operational hazards,
operational sabotage, fire, earthquake, flood,
windstorm, and bombing, the controlling considera-
tions appeared,to be those associated with a total
reactor tragedy. The total reactor tragedy is con-
sidered here as being an accident in which all the
heat that could possibly be released from the
chemical combination of various materials in the
reactor and associated system would be released,
together with the heat from the fission products
accumulated after extended operation at full power
and the energy released in an extreme nuclear
runaway. 1he principal hazard associated with an
ultimate reactor catastrophe is the dispersion of
the fission products that would have accumulated
from extended operation at high power. The key
data on these parameters are presented in Table
2.2,

In examining the data on the heat that could be
released from the combustion of various materials
in the reactor installation, it is immediately evi-
dent that if kerosene or another hydrocarbon were

used in place of water in the shield a very large

amount of heat could be released, an amount almost
one hundred times greater than that from any other

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PERIOD ENDING DECEMBER 10, 1954

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'ANP QUARTERLY PROGRESS REPORT

TABLE 2.2, SUMMARY OF KEY DATA ON ART HAZARDS

 

Sources of Energy

Heat from combination of 1000 |b of Na and NaK with water
Heat from combination of 1000 1b of Na and NaK with air

"Heat from combination of 1200 Ib of zirconium-base fuel with NaK
Heat from combustion of 34,000 |b of shield kerosene with air

Heat from éxtreme nuclear accident

Fission-product decay gamma heat emitted during first 2 hr after shutdown {assuming no

fission-product removal)

Sources of Radicactivity

Total fission-product activity for saturation

2.1 x 10% Bt
2.4 x 10% Bt
0.13 x 10% Bty
635 x 10% By

0.3 x 108 Beu
8 x 10% Bty

6 x 108 curies

Gaseous fission-product activity for saturation 107 curies
Sodium activity in moderator circuit for saturation 10° curies
Sodium activity in NaK circvit for saturation 35 curies

“,Temperafures Associated with Accidents

#I_qme temberfifure for stoichiometric NcK*-H20 reaction 3070°F

Temperature of reaction products for Na*—~zirconium-base fuel reaction 3300°F
_Tefnpératur_e of a‘t'mo’sphere in ]2,000-“3 tank (assuming uniform dispersal of 310°F
.--.com_i.fopstiidn prbduéffi from NG-H2O reaction)

’Tempérafurje rise of NQK-H20 reaction products and shield water if uniformly mixed 62°F

Major Radiation Sources and Doses for Typical Conditions**

Equivalent source for all fission products

Equivalent source for inert gases only

Rate of generation of activity in the form of inert gases (assuming 20-min holdup in

atmosphere of expansion tank)
Same as above but for 420-hr holdup time
Dose at shield surface during operation
Dose at shield surface 15 min after shutdown

Dose at shield surface 10 days after shutdown

Dose outside 24-in.-thick concrete wall 15 ft from center of reactor during operation

Same 15 min after shutdown

4 x 108 curies

108 curies
1000 curies/sec

1 curie/sec
100 r/hr

4 r/hr
1r/hr
0.025 r/hr
0.001 r/hr

 

*|nitial temperature assumed to be 1500°F,

**These key data for typical conditions were developed for the major radiation sources and doses by assuming
1000 hr of continuous operation of the ART at 60-Mw power level with only the gaseous fission products coming
off and by assuming the reactor to be encased in an aircraft-type shield which gives a dose of 1 r/hr at 50 ft.

source of energy. The heat that could be released
from the combustion of the sodium and the NaK in
the moderator and heat dump circuit is relatively
‘small; thus it seems that water should be used as
Vt'he' shielding material rather than kerosene, be-

- cause, even if the water were to combine in stoi-

chiometric proportions with all the sodium and the
NaK in the system, the resulting energy would
still not present a difficult problem, The heat
that would be released from a nuclear accident
depends in large measure on the character of the
accident. However, the value presented in Table
2.2 is that for the worst accident that can be

36

envisioned, that is, one in which uranium abruptly

begins to precipitate out of the fuel in the core so
that the fuel is carried into the core at the normal
rate but no uranium leaves with the exit stream of
fluoride. While hardly a credible accident, this
does seem to represent the maximum rate of in-
crease in reactivity and, hence, the extreme nuclear
accident conceivable for this reactor.

The second portion of Table 2.2 presents the
equivalent radiation sources for the various fluid
circuits. The values given for the sodium and NaK
activities have been obtained from multigroup
calculations.

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It should be noted that the presence of the shield
water would provide a strong damping effect on any
temperature rise associated with an accident, The
third part of Table 2.2 shows the temperature rise
associated with the release of energy from the
various sources for various conditions. As shown
in the table, if all the heat from the combustion
reactions were to be confined in the combustion
products, high temperatures would result from
stoichiometric reactions, Even in such instances,
much of the heat would be dissipated to the atmos-
phere within the containing vessel so that the
temperature would be substantially lowered, If
the reaction were quenched by the shield water,
still lower temperatures would result.

It is clearly essential that provision be made to
contain the products of a total reactor tragedy
except, possibly, if the reactor is installed at
NRTS. A comparison of key data for several types
of container is given in Table 2,3, For this com-
parison the worst set of conditions applicable to
each case was presumed. As may be seen, the
double-walled tank compares favorably with the
hemispherical and ellipsoidal buildings. While
difficult to evaluate numerically, the double-walled
tank also appears superior in that it would be less
subject to sabotage. Even if both walls were
ruptured and the reactor melted down, the residue
would tend to sink to the bottom of the tank pit,

PERIOD ENDING DECEMBER 10, 1954

where it would be flooded by the water that had

filled the region between the walls. This water
would both absorb the heat of any reaction and act
as a shield to reduce the radiation level at the top
of the pit. After careful review of these and a host
of lesser considerations, the 24-ft-dia double-walled
tank was chosen as the most promising test facility.

Although the double-walled tank type of test
installation is adaptable to several types of reactor
test facility, the most promising possibility is to
install it in an addition to the ARE Building. Such
an arrangement would permit the use of services
and facilities that already exist in the ARE Build-
ing, for which no plans have been formulated now
that it has served its original purpose. For ex-
ample, items such as the control room, offices,
change rooms, toilets, storage area, water supply,
power supply, portions of experimental test pits,
access roads, security fencing, and security lighting
are available.

Of the several schemes considered for modifying

the ARE Building to accommodate the ART, the

‘most attractive plan is to construct an addition on

the south end to effect a 64-ft extension of the
present 105-ft-long building by extending the ex-
isting roof and side walls. Figure 2.8 shows a
preliminary design. The floor level of the addition
would be at the ARE basement floor grade, with
the double-walled tank to house the ART sunk in

TABLE 2.3. COMPARISON OF KEY DATA FOR SEVERAL TYPES
OF REACTOR INSTALLATION CONTAINERS

 

 

 

24-ft-dia 17-ft-dia _ _
Doub[e-Walled - Double—quled 209..”-?1'0 11-5-f1-d:la
Tank with 11t Tank with 9-f = 1ipseidal - Hemispherical
Strmght Secflon Srrmghf Sectton Building Building
Heat released, Btu | o 1'0"5' 2.4 % 106 ' 107 107
conm.ner vo[ume, &3 Sy 200 4600 1.2x108 0.4 x 108
”\Confumer surface clr_ea ft2 N o 2640 . ]400 4.3 x 104 2.1 x 10%
. Peak Qus femperaturé,* OF A 44'4' o 970 T 130 ' 220
Peak gas pressure, ps;lg 1 ) | 37 T 1.4 3.9
7_ _Requnred shell 'rhtckness (for allow- T e e k & T '
~able s"?:’S.f,]S,O 00 p51), in. o 0.217 _ 0.251 0.15 - 0.09
: We;ght ofsfeelmsheii, o R e L 5 ]34** SRR

 

 

*Eission-product afterheat was not includedin caleulating peak gas temperature and pressure where adequate, provi-

sion was made for cooling the container wall.

**lncludes steel frame.

 

37

 

I

 
i s i e

8¢

 

s,

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
ORNL-LR-DWG 4471

e 14 Y5 in.

 

 

 

 

 
 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

   
 

L HATCH

 

 

 

POURED GONCRETE
SOLID CONGRETE BLOCKS
[777] CORED GONGRETE BLOGKS

Fig. 2.8.

20 4 8 12 16 20

FEET

oo

 

A 4—| .
i
1
i 106 ft 1in. 64 ft O in, ADDITION——‘
—r——~ /\l /\‘ /\ I/\ - - W WY B e O RN AN 835 KT I
LOCKER ROOM & oea w
INSTRUMENT COUNTING OFFIGE CONFERENGE > 3
ROOM ROOM ROOM - 7 :’oj Zw
— E 20 INSTRUMENT
A N & E = SHOP
=0
I A CORRIDOR 3 % 2
DOWN& g —]
2 n/\}—"b——r\/\y g -
£ i
0 up £
fi [+2]
GENERAL CONTROL, ROOM w
SHOP 3 & FIELD MAINTENANGE SHOP
= <
& £
+1 -
= ] T _‘L * + i K| ]
[i2] 3 M H = R ZZeaTTTT. Ty T A e T T A
L -EXISTING 12-X 12-ft 5- x 8-ff CONTROL INSULATION
ROLL-UP DOOR, 3-x e T T T TRENCH
7-it SWING DOGR AND
COLUMN TO BE RE-
MOVED. J
k !
<7 2 |><]
N =
d b ol b b ) e e o) b b e g%uzl -
T R | L o[ 4 [ o jjg c G
L NEW 26-ft-HIGH X o ) t J
30-ft-WIDE ROLL- =& W
UP DOCR. = = -t £t 4in.
|
P~ o
1 " L 11 1t i3l 1 )I( ")
1 1 e 1 1 1 11 1|1 1 1 m
:O ]
F 7 T 7 il
PART FIRST FLOOR PLAN
EXISTING BUILDING
£ MONORAI
LEGEND AROV

  

PART BASEMENT PLAN

  

101t

- )

14OdIY SSFIO0Yd ATIILIVNO NV
 

Hin

”"L"‘:‘ ‘. N

the floor up to 3 ft below the bolting flange level.
The reactor tank location would be approximately
centered in the 42-ff-wrde, 64-ft-long high-bay ex-
tension directly in line with the ARE experimental
bay. The reactor would be positioned so that the
top of the shield would be at the building floor
elevation,

. To permit use of the experimental pits for instal-

lation of auxiliary equipment, to permit possible
underwater reactor disassembly work after reactor

operation, and to provide a large entry door to the
ART areaq, the south wall of the ARE experimental
bay would be removed, the overhead crane facility
would be revised from 10- to 20-ton capacity, and
the truck door in the nor'rh wall of the ARE Building
would be enlorged.

The double-walled tank described above is
essenhclly the container conSIdered for this instal-
lation. The inner tank, or pressure vessel, would
be approximately 24 ft in diameter with a straight
section about 11 ft long and a hemispherical bottom
and top. The joint between the removable top and
the lower portion of the tank would probably be «
bolting flange, with provision for sealing the joint
with a low-melting-point alloy, Thus the top would
be used only as required by the operating program.
The outer tank would be a right circular cylinder
approximately 27 ft in diameter and about 47]/2 ft
high. About 26 ft of this water-containing tank
would be above floor grade, and this portion of the
tank would be ofl'ached by a flange or weld joint
only when the operating program requrred that the
upper hemisphere of the inner tank be in place.
The inner tank would be set coaxially with the

outer tank, and the bottom of the inner tank would <

be 1. St above fhe bof'rom of the . outer fcnk T

 

' ".would prowde o ]5& cnnulus befween 'rhe fwo\

 

  
    
 
  
 
 
 
  

“dp

 

“ere Pr fype shleld ‘with @ gross W ghr “of about 35
tons, would be mounted on verr:cal col‘umns ‘with
"the reactor center 3 ft off center fromflt e vessel
axis and about 12 ft above an open- groted floor.
This positioning would provide needed space for

PERIOD ENDING DECEMBER 10, 1954

movement of the portable fluoride fuel and sodium
moderator coolant containers to their operating
stations under the reactor. The off-center [ocation
would also serve to minimize the length of NaK
piping that must run from the reactor through a
bulkhead in the pressure vessel to the heat dump
radiators outside the tank.

Quite a variety of shields has been considered
for the ART. The most convenient seems to be
one functionally the same as that for an aircraft
requiring a unit shield, namely, a shield designed
to give 1 r/hr at 50 ft from the center of the reactor.
Such a shield is both the lightest and most com-
pact that has been devised. It makes use of non-
critical materials that are in good supply, and it
will provide useful performance data on the effects
of the release of delayed neutrons and decay
gammas in the heat exchanger, the generation of
secondary gammas throughout the shield, etc.
While the complication of detailed instrumentation
within the shield does not appear to be warranted,
it will be extremely worthwhile to obtain radiation
dose level data at various points around the pe-
riphery of the shield, particularly in the vicinity
of the ducts and the pump and expansion tank
region.

Several arrangements have been considered as a
means for disposing of the heat generated in the
reactor. The most promising is one that resembles
a turbojet power plant in many respects. It em-
ploys radiators essentially similar to those suit-
able forturbojet operation. Conventional axial-flow
blowers would be employed to force cooling air
through the radiators. This arrangement is flexible
and as inexpensive as any arrangement devised.
’WI“ give thermal capacrtles cnd flurd transit

 times essen’rml[y the same as those in a full-scale

ireraft | power plom‘. It will also gwe some very
a!uabie expenence ‘with ‘the operation of high

- temperature hqurd#ro-cur hear exchangers that em-

“"body features of construchon ‘and fabricating

~ ‘techniques suitable for aircraft use, It should be
mem‘loned fhcf ‘the type of radiator currently en-

ed i |s_yone that has been tested’ at ORNL and

    
 

proposed ro “date.  While’ r‘rsb performance leaves
“something to be desired, it is felt that the basic

' cpnflgurotion is sound “and that modifications can
”be made Tater - fo give improved performance.. '

‘As shown in Fig. 2.8, 16 NaK-to-air heat dump

radiators would be mounted in an air tunnel which

39

,
AN
h o= = s

Vg
wend
o
e
T
i

e

 

pedrs "tombe fhe ‘most rehable of any radiator

 

 

 
kiR e B S S0 i cne s

 

 

 

ANP QUARTERLY PROGRESS REPORT

would be placed on a diagonal across the south-

“west corner of the addition. Each radiator core

would have an inlet face 24 by 24 in. and a depth

- of 6 in, These radiators would be located at floor

grade over the NaK pipe lines and the NaK fill and
drain pit. Four 75,000-cfm axial-flow blowers de-
signed to give a head of 10 in, H,O would force
air through the radiators and out through a discharge
stack.

The fuel fill-and-drain system envisioned for the
ART incorporates two shielded dump tanks. One

would be coupled to the reactor with a remotely
- operated coupling, while a second emergency dump
“tank would be welded directly to a discharge pipe
~from the reactor. The remotely operated coupling

to the first fuel dump tank would give flexibility*in

' the operation of the reactor, since it would make it

possible to bring fuel to or remove it from the site

. expeditiously and would keep the footage of pipe
~and the number of valves to a minimum.

For- handling the heavy, shielded fluoride and

sodium containers inside the inner tank, a track

would be installed on the floor and inside the wall.
Wheels would be mounted on both the bottom and
one end of the tank dolly so that the assembly
could be lowered by the overhead crane to the floor
track, with the end wheels on the dolly riding
against the vertical track. Once on the floor track,
each dolly would be moved to its operating station
under the reactor. Each track pair in this area
would probably be mounted on a {ift for raising the
tank connection nozzle to the contact position
within the reactor shield.

In addition to the NaK piping bulkhead mentioned
above, a control junction panel such as that shown
in Fig. 2.7 would be installed as a part of another
bulkhead through the tank below the building floor
grade to pass the tubing and the electrical con-
ductors required for operation, control, and moni-
toring. Two more bulkheads, in the form of man-
holes, would probably be installed in the upper
portions of the inner and outer tanks. One manhole
would be about 3 by 5 ft and located just above the
inner tank flange to allow passage through both

container walls and thus provide entrance to the

- 40

inner tank subsequent to placement of these sections
of equipment. The second opening would be a
manhole about 5 ft in diameter in the hemispherical
top of the inner tank to provide overhead crane
service after placement of the top. Sufficient cat-
walks, ladders, and hoisting equipment would be
installed within the inner tank to provide easy
access for servicing all equipment, as shown in
Fig. 2.6. A -

The control bulkhead in the doubl_é-wdl[ed'tank
would be located so that the associated control
junction panel and the conirol tunnel would be on
the control room side of the tank. The tunnel
would extend to the auxiliary equipment pit (formerly
the ARE slab storage pit), where it would t_erminofé.
The tubes and conductors from the junction panel

‘'would be channeled from the tunnel either to equip-

ment in the pit and basement or to thé control
room (formerly the ARE control room). The pit and
basement equipment would include such items as
the lubricating oil pumps and cooler, borated shield
water pumps, cooler and makeup equipment, vacuum
pumps, relays, switch gear, and emergency power
supply.

The reactor off-gas flow would probably be piped
through the NaK piping bulkhead to a disposal
facility outside the building., Such a system would
probably consist of an activated-charcoal absorption
bed contained within a pipe which would be long
enough to provide the required delay period to
bring the activity of the krypton (which would not
be adsorbed by the charcoal) to a tolerable level.

Field maintenance and laboratory facilities
would be installed in the area east of the ART test
bay and south of the low bay of the ARE. This
area and the ARE experimental bay would be
partitioned from the ART test bay with about a

16-in.-thick shield wall of stacked solid concrete

block, This wall would not be erected until after
placement of the upper sections of the double-
walled tank, The only other major rework of the
ARE facility to accommodate the ART would
probably be that of modifying and equipping one
of the ARE experimental pits for underwater dis-
assembly work on the reactor after operation.

B e ey

£ o A
3 N oy

- ¢ ST a5

fe
R T
 

g

ey

“loop test results,

PERIOD ENDING DECEMBER 10, 1954

3. 'E'XPERl_MENTA_L' REACTOR ENGINEERING

H. W. Savage

Alrcrcft Reactor Englneerlng D|v15|on

-Design work is under wcxy on an in- pt|e test
[oop for operation in a horizontal beam hole of
the MTR. This work is being done in cooperation
with the Solid State Division and Pratt & Whitney
Aircraft. The specifications for the loop have
been established. Tests are under way on a
horlzonfal shaft sump pump for the loop, and a
salt-to-air heat exchanger has been fabricated.

‘The use of ARE-type sump pumps for high-flow
heat exchanger tests is being investigated. The
K-25 loop test of ARE components is being con-
tinued as a life test of the ARE type pump, and
over 3000 hr of operation hcs been [ogged De-
velopmental work is under way on pumps for the
ART, with particular emphasis being given to
impeller fabrication and performance.

Three Inconel forced-circulation loops have been
operated with NoF-ZtF;‘-UFI‘.' The first two loops
tailed after 48 and 3 hr, respectively, but the third
loop has operated for 150 hr with a Reynolds
number of 10,000 and a 200°F temperature gradient,
A third beryllium-sodium-Inconel mass transfer
test has been completed and has been submitted
for metallographic examination, and components
have been fabricated for tests of sodium in multi-
metal loops. o

Developmental work is under way on components,
such as a pump and a gas-furnace heat source,
for a loop for testing fused salt-to-NaK heat ex-

changers. A leak test has indicated that smq[l ;
|eaks of fluoride fuel into NaK are self-plugging. ~

IN-PILE LOOP COMP

   

ature conditions within the range of out-of-pile
forced c1rcu|otlon exper:ments, data should be

     
  
  
 
 
  
 
 

   

By Teshng ot flow cmd 'remper-

obtoined on Which to base extrapolations of the
more readily obtainable out-of-pile test data. Of
course, space limitations of the reactor beam
holes somewhat limit the conditions ob’ramctb!e
in a safe and reliable experiment.

The proposed test unit will consist of an 8-ft
loop of Y%-in. schedule-40 seamless pipe with a
little over 1 ft in the active zone of the reactor.
The ratio of the total volume to the volume in the
high flux zone, known as the dilution factor, is
approximately 10. Flows in the turbulent region,

'Reynolds number approximately 4000, with a

maximum temperature of 1500°F have been pro-
posed. A temperature differential of 300°F and
heat transfer of 30 to 50 kw are expected with a
power density of approximately 2 kw/cm®. These

" conditions probably can be achieved with a fuel

composed of f\lcfl“:--Zrl:A-L'H:4 (53.5-40.0-6.5 mole %).

Design work is proceeding on this loop and its
components to meet the above specifications.
Handling equipment is planned which will fa-
cilitate isothermal operation of the loop prior to
insertion into the reactor and which will also
receive the radioactive materials after in-pile
operation,

Horizontal-Shaft Sump Pump
J. A. Conlin ~ D.F. Salmon

Aircraft Reactor Englneenng Division

--Two, ldenhccl horizontal-shaft, centrifugal-type
sump pumps fabrlcc'red cccordmg to the design

"descrlbed prev:ously ‘were tested with the fluo-

ride mixture. ‘NaF ZrF (50-50 mole %Y at 1350°F

dnd a shaft speed of 6000 rrpm for 500 and 1000 hr,
; .fespectlvely. Both pumps showed consuderable
T WEGr “on the Grcphltar lapped ‘steel face pla'res,"
b_vbut nelther ‘seal had failed complefely.

   

 

 

p, F. Salmon, ANP Quar. Prog. Rep. June 10, 1954,
ORNL-1729, p 19.

41

 

.>ome m! '

the shaft arrangemen’r':' o
been altered to prowde '

    

i
! L
L

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

a lower seal temperature; also, the shaft will be
operated at a lower speed. The pump is a
“‘regenerative-turbine’’ type, but it differs from
‘conventional design in a number of points. There
is no positive shaft seal, but, rather, there is
close clearance between the shaft and the impeller
housing. Some fluid leaks along the shaft laby-
rinth into the small combination sump and ex-
‘pansion chamber and is then drawn back into the
impeller through a hole connecting the sump to
the pump suction. To accommodate thermal ex-
pansion, total impeller side clearances have been
~increased to 0.060 in., as compared with the usual

UNCLASSIFIED
. ORNL—LR-DWG 4522

 

64
o IMPELLER SIDE CLEARANCE (in.)

60 SYMBOL | FRONT [BACK|TOTAL
. 0030 |0031 1006/

a 0022 0041|0063

 

 

 

 

 

 

 

 

 

 

 

 

. b6 ]
. \\ o 0043 0020|0063
52 M- &
1S \\OOO
IR
48 LN \

 

N
44 NN

N

 

40 NN

 

 

36

2 N
T~

 

HEAD (f1)

 

28 \0§\\¢
X

 

 

 

%0 NN
24 <90, e
~ ",o,b .
20 \%i}\&\
=
5 ;;\\
12 \'\\

 

 

 

 

 

 

 

 

 

 

 

 

 

-0

 

O 02 04 06 08 10 12 14 16 18
' "FLOW {gpm)

Fig. 3.1. Performance Characteristics of a
RegenerativesTurbine Type of Horizontal-Shaft
Sump Pump Tested with Water.

42

0.005 to 0.010 in. The size of the pump envelope
has been reduced by locating the inlet and dis-
charge ports axially rather than radially at the
periphery and by decreasing the annular passage
around the impeller.

A Lucite model incorporating the changes made
as a result of the tests described above has been
tested with water. The curves of Figs. 3.1 and
3.2 give pump performance characteristics at dif-
ferent speeds and impeller side clearances. As
may be noted, the total impeller end clearance
is a primary factor in determining pump per-
formance, while the distribution of clearance, front
to back, is relatively unimportant. This is fortu-
nate, since the end clearance can be fixed in
fabrication, but the impeller position is affected
by thermal expansion and bearing alignment. ’

UNCLASSIFIED
ORNL-LR-DWG 4523

 

 

 

 

 

 

72 {
o8 IMPELLER CLEARANCE {in)

POINT [FRONT |BACK [TOTAL |
1 0024 [0020[0044

64 ; 2 |0030 (0030|0060
. 2 10030 (0031|0061

60 N a |ooz2|0041]0063 -
\ 5 |0.043 0020|0063

56 f"oo 6 |oo019 |oosi|oos0
7 |0058 0021|0079

 

 

 

 

5
\A%

52

48 \ 2

44 N\

3 3
L N
40 <
4

FLOW: 0.92 gpm

 

 

 

HEAD (ft}

 

 

” \\
32

28 D

 

 

 

 

 

 

 

 

 

 

 

 

1 .
4
24 o0, T~
\ Ly : &
.
20 \ 7
\2 4
o
.
\Z.
t2 6'
8 .
0.040 0050 0.060 0070 C.0BO

TOTAL IMPELLER END CLEARANCE {in}

Fig. 3.2. Effect of Total Impeller End Clearance

on Performance of the Regenerative-Tuthine Type
of Horizontal-Shaft Sump Pump Tested with Water.

TPy e ,‘;‘\ “*’ ¢
LT S 3R ()
 

 

 

»)

LY

The pump has unusual ability to remove gas.
In water tests, air was injected into the impeller
at a rate of several cubic inches per minute and
was ~ continuously removed with only a slight

“increase in the amount of gassing. Under normal
operation there is usually no more than 1 vol %
‘of gas in the fluid.

The pump will satisfactorily prime and fill the

test loop in which it is installed if fluid is added

to the sump while the pump is running and the

total impeller side clearances are mamfalned at

0.040 in. or less. However, priming becomes
errafic with the greater clecrcnces required for
operation of the pump at high temperatures.
Several mefhods have been demonstrated for ob-
taining a posmve 'dlsp]acemenf of fluid into the
foop with Iarge pump clearances. Each method,
however, requires either complicated freeze valves
or pressure controls. Vacuum filling appears to
be the most atiractive solution to the problem but
has not been ‘rested

Heot Exchanger
D. F. Salmon' L. P. Carpenter

Aircraft Reqcfor Engmeermg D|V|5|on

A 5|ng|e-tube, longitudinal ly fmned, salt-to-gir
heat exchanger has been fabricated for use in an
in-pile loop. Since the staggered-fin arrangement

required to ensure adequate heat transfer with air

resulted in a very high pressure drop, it seems
imperative to use a lead shielded recirculating
system with a secondary gas-to-water heat ex-
changer for safety in the event of a salt break
into the gas stream. The use of helium would

simplify both heat exchangers, ‘dnd therefore a
" ““ticipated in the orlgmal desagn, ’rhat IS, above |

2200 rpm. The crn‘lcal speed rcmge was found to

'sult-’ro-hehum exchanger is no bemg desng

chcnges and to evaluate the probable trouble-free
life based on actual use.

 

      
 
  
    
 
    

"’estabhsh'\performunce ranges without ma|or deSIgn

PERIOD ENDING DECEMBER 10, 1954

The test loop, set up at K-25 for eS:omining

ARE fuel-tube hair-pin sections and an ARE heat
exchanger, was stopped after 2047 hr of nearly
trouble-free operation for removal and examination
of the ARE test components. At the time of
removal of the ARE test components, only a
cursory inspection of the pump was made because
operation had been trouble-free except for the
plugging of gas nozzles with zirconium fluoride
vapor condensate. An inspection of the inside
surfaces of the pump tank cover revealed sta-
lactites of salt that covered nearly all riser and
nozzle skirts and bolts and protruded beneath the
lower heat radiation baffle intoe the pump tank
gas space. These stalactites are not deleterious
to pump operation. The loop was modified to
consist only of the pump, loop piping, throttling

valve, and the previously used venturi element,

and the pump has now been operated for an ad-
ditional 1000 hr at 1350°F, 40 gpm, and 1500 rpm.
Slight noises were noticed at the pump bearing
housing after a total of 2847 hr of operation.
However, these noises had not greatly increased
when the total operating time had reached 3000 hr.
The gas-line-plugging problem which was en-
countered after 2047 hr of operation was eliminated
completely by a newly designed vapor trap. Oper-
ation of the loop is being continued as a life test
of the pump. |

Pump performance data (Fig. 3.3) were obtained
over the speed range 600 to 1400 rpm with the fluo-
ride mixture NaF-ZrF,-UF, (53.5-40-6.5 mole %)
at 1300°F., Other tests of ARE-scale pumps
were made to determine critical speeds and per-

formance dcta at speeds higher than fhose an-

 

be bef ween 2850

NO SO ':olcm’r pumps
th copacmesé of“l430'gpm each at a ‘1’25 -t heod
“ahd two or more NaK primary-coolant pumps with
capacities of 2800 gpm or less at a 280-ft head.

 

The problems with all these pumps will not be

43

;nd 3250 rpm “and is fhus well

cbove The speeds requtred for “heat exchanger '

 

b
i

T

 
 

 

 

 

ANP QUARTERLY PROGRESS REPORT

 

 

 

 

 

 

 

STONFOENTIAL
. ORNL-LR-DWG 4524
€0 60
50 1400 rpm >0
—'——_—-—-.__—-_-—_~
3 \ 20
\EFFJCIENCY AT 4200 rpm

{200 rpm 63
— —'_—'_-'—-—_______‘~~ —
ht " >

< 30 0
: / T~ 08
= \ 5
L
/ w

 

/ ——___'000 rpm
\

 

/ 800 rpm

 

‘\ "’

 

 

 

 

 

 

 

 

 

 

 

 

 

/ *——'_‘-—-.\
/ 600 rpm \
"o 0
. o 10 20 30 40 5

0o 60 70 80 20

FLOW (gpm)

Fig. 3.3. Performance of an ARE-Type Sump Pump with NaF-ZrF ,-UF , (53.5-40-6.5 mole %) at-'l-300°F.

unlike those encountered in developing ARE
pumps, but they will be magnified by the higher
capacity requirements, by the resulting higher

-power requirements, and by a number of special

restrictions imposed because of the proposed
locations of the pumps.

The location of the fuel pumps imposes problems
in impeller and in inlet and discharge housing
design, since it will be necessary to match the
fluid dynamic requirements of passages leading
from heat exchangers into the pumps and from the

pump discharge housings into the core. Various

impeller des‘igns"rure being investigated and will
be used to evaluate cavitation characteristics.
Previous experience with ARE impellers indicates

that casting is not a promising means for fabri-

cating Inconel impellers in the present state of
the art, and therefore a brass impeller has been

'made for studying fabrication problems and ob-

taining initial performance data.  Preliminary
planning has been started for the fabrication of

.44

models for study of pump discharge and entry
conditions, -

Seal design, bearing structure, cooling, lubri-
cating, and driving means are other problems being
studied, and the reliability of these components
and operations is being ascertained through a
succession of mechanical shakedown, water per-
formance, and high-temperature tests.

DESIGN AND OPERATION OF FORCED-
CIRCULATION CORROSION AND
MASS-TRANSFER TESTS

Fused Saits in Inconel

.. A. Mann W. B. McDonald
7 W. C. Tunnell
Aircraft Reactor Engineering Division
The Inconel forced-circulation corrosion test
loop previously described? as loop No. 2 was put

 

2y, C. Tunnell, W, K, Sfcir, and J. F, Bailey, ANP
Qua{. Prog. Rep. Sept. 10, 1954, ORNL-1771, Table 3.1,
pdi.
RN . e Myt S Pl

 

.

s

)

)

‘reducing the power input into the heater section,

fuel being used is NaF-ZrF4-UF4 (53.5-40-6.5

el Reocer Enginaaring Division
A third test® has been completed in which molten
. “sodium was pump : :
" taining a beryllium insert. Approximate test con-_
. lemperstire, 1300°F; minimon flowing. sediom <14
_temperature, less than 1000°F; Reynolds number =~ _
" at minor diamefer of beryllium insert, 190,000, D. R. Ward L. A. Mann

 

PERIOD ENDING DECEMBER 10, 1954

into operation with a Reynolds number of 10,000
and a temperature gradient of 200°F in the fuel
mixture NoF-ZrFA-UF4 (50-46-4 mole %). The
design of this loop is illustrated in Fig. 3.4. A
failure occurred in the heater section of this loop
when the pump was stopped after 48 hr, without

UNCLASSIFIED
ORNL-LR-DWG 4525

  

to take a faulty instrument out of operation. This
loop was constructed of 0.5-in.-0D, 0.020-in.-wall
Inconel tubing. Three unsuccessful attempts have
been made to get the loop back in operation, but,
in each case, failures of the tubing occurred during
the cleaning operation with barren fuel mixture.

Another loop (loop No. 3) constructed of 0.5-
in.-OD,  0.040-in.-wall Inconel tubing was suc-
cessfully started and a Reynolds number of 10,000
and a temperature gradient of 300°F were attained.
After operation for 3 hr, the test was terminated
because of failure of tubing near a welded joint.
The data from the startup revealed conclusively
that the loop could be operated without a heat
economizer; therefore construction of the loop can
be simplified.

A third loop, without the economizer, was built
with 0.5-in.-OD, 0.045-in.-wall Inconel tubing.
This loop included a sump pump (model LFB),
a cooler, and a heater. The loop has been in
operation over 150 hr with a Reynolds number of
10,000 and a 200°F temperature gradient. The

 

 

e , HEATER
mole %). The maximum fuel temperature of 1500°F

is produced by a power input of about 72 kw. This
loop is scheduled to operate for 1000 hr.

Beryll i'qusbd ium-Inconel Ma s_s' Trdnsfg‘r

W BM‘cDonald ) :Fi‘g. 3.4. Fdrced-Circuhfio_n 'Cr_ollfrbsii:ilh Test

" l.oop.

 

 

‘ah an Ineanal leas cons and other sections of the loop have been sub-
ough an Inconel loop con- " LS B SUR EER TENEORETE
ST et oo 2o mitted for metallographic examination, A fourth
ow rate, 3.5 gpm: berglliom test loop of similar construction has been fabri-
Do o P oo cated and is being instqlled for testing.

 

    

s

Sodium in Multimetal Loops

 

   

i .. W.B.McDonald -
... . Aircraft Reactor Engineering Division
. The design of a |oE>p4 fqr feéfihQ éofn_bincfions
of structural metals in contact with high-velocity

 

 

3L. A. Mann and F. A. Anderson, ANP Quar. Prog. 4. A Mann, ANP Quar. Prog. Rep. Sept. 10, 1954,
Rep. Sept. 10, 1954, ORNL-1771, p 32. ORNL-1771, p 41.
45
Y, T e
Vo i o]

 

 

 

 

 

 

 
 

 

 

'ANP QUARTERLY PROGRESS REPORT

turbulent liquid metals under high temperature
differentials has been completed, and components
have been fabricated. The first series of tests
will include combinations of Inconel and type 316
stainless steel as loop materials, with sodium as
the circuvidted liquid metal.

Nutural Gas Heui‘ Sources for Forced-Cnrcuiahon

Loops
L. A, Mann
Alrcrafi Reactor Engmeermg Division
R. Curry 7 ER Dytko

Profi & Whl'rney Alrcraf’r

Several ’rests of burners were made in v order to
obtom prellminary data for design of a natural-gos
heat source for forced-circulation corrosion test
Ioops. It was found that exhaust gas temperatures
of between 2800 and 3300°F could be obtained
without initial air preheating, and, by using air
at a rate of 65 to 650 cfm, the equivalent of 100
to 1000 kw of heat was released. Maximum system
thermal efficiencies of 20 to 60% appear to be
feasible.

Tests with eleciric-resistance heated units indi-
cate a maximum wall temperature of 1750°F with
the tubing lengths required by pressure-drop and
flow considerations. Somewhat higher tube-wall
external element temperatures or proportionately
longer heater sections would be required with gas
heating. An economic study indicates that there
is little choice between providing electrical power
supplies and providing large-capacity air supplies
for gas heating. Therefore, since development of
a gas-heated unit specifically for this purpose
would be expensive and since there is no present
proof of reliability of components and operation
of gas-fired loops, further study of gas firing for
these test loops appears to be unwarranted.

HEAT EXCHANGER DEVELOPMENT
‘ Heat Exchanger Tests
R. E. MacPherson
- Aircraft Reactor Engineering Division

R. D. Peak
Pratt & Whitney Aircraft

A number of tests of fused salt-to-NaK heat
exchangers are to be made for determining the
fabricability, operability, operating parameters,

7.and reliability of heat exchanger tube bundles

such as those proposed for the ART. The test

46

loops are to include reliable pumps, reliable heat
sources, and adequate heat sinks.

A preliminary heat exchanger test reported previ-
ously® indicated reasonable structural integrity
of the exchanger (1682 hr of operation), but little
data on operating parameters were obtained. For
the next experiment, a regenerative system is to
be used in which the heat exchanger will be an
economizer that will transfer about 4 Mw of heat,
while a heat input and release of only 1 Mw will

_be reqmred

Header Leak Test
R. E MacPherson

A:rcraft Reuctor Engmeerlng DIV!SlOl"l

R. D, Peak
Pratt & Whitney Aircraft

It has been postulated that small leaks which
develop in metal walls separating NaK and fluo-
ride fuels may be self-plugging as the result of
the precipitation of high-melting-point reaction
products. A program of tests is currently under
way to check this possibility. To date, one test
has been run, which indicates that self-plugging
can occur with NaF-ZrF,-UF, fuels. The leaks
are installed between three cells as shown in
Fig. 3.5. The center cell is pressurized to leak
into the end cells under the pressure differentials
shown, 5 and 50 psi.

The test rig was filled with NaF-ZrF ,-UF , fuel
in the center cell and NaK in the end cells. The
leaks were about 0.002 in. in diameter. After
12 hr at 1300°F, there was no apparent leakage,
and x-ray pictures taken at intervals during the
test showed no changes in liquid levels.

Pump for Heat Exchanger Tests

W. G. Cobb W. R. Huntley
A. G, Grindell R. E. MacPherson

Aircraft Reactor Engineering Division

A series of tests is being made in order to
determine whether an existing model DANA pump
(ARE moderator-coolant-type sump pump) is suit-
able for the high-speed operation which will be
required in the intermediate heat exchanger test
loop. Critical shaft-speed determinations made
by K-25 personnel indicated that the pump could
be operated saofely at the proposed speed of ap-
proximately 3700 rpm.

 

SA. P. Fraas, R. W. Bussard, and R, E. MacPherson,
ANP Quar, Prog. Rep. June 10, 1954, ORNL-1729, p 22.
 

i, L

 

¥

’?

»

-

iy

 

"flnveshgatlbjns of gas fired furnfl‘aces E:cpable Qf
transferring h_ea’r to [quId mefal‘uf a hlgh ‘rate }

 

reoson.

PERIOD ENDING DECEMBER 10, 1954

UNCLASSIFIED
PHOTO 22735

: . = CENTER
END CELL f  CELL

END CELL

 

Fig. 3.5. .H..eai E.x(:hanger-l'!e;dder Leak Test,

' The rotary element of the pump wdsr'rfhen oper-  combined in forthcoming ‘ex_perimenfs. The problem
ated at 3700 rpm in the cold shakedown test stand, is also being explored with vendors of gas-fired
and the bearings and mechanical oil seals were  equipment.

found to operate satisfactorily. The pump is now .
being observed for performance qnd mgassmg at Sodium-Cooled 100-kw Furnace Tests
hrgh speeds in fhe wo’rer test stand. R, E MocPherson o R W, Bussmd

  
 
 

 

nay 'be requrlred H’&Vé’hgweh'—n : gas-f:red sodwm heat source wds operaf@d wwh
: an ou?put of up to 85 kw at sodlum oufle'r temper-
coson gtures of ]500°F The mstaliatton 1!8 “shown in

 

 

 

SN

 

_.,.‘ A LT AT

nlf" volfm'\e‘wnthout exfenswe heat stor

   

: F|g 3.6. A compcrlson of &emg,n cgnfltf_q_qns w:fh

   

    
  
 

al ¢ ) hr of e was ac pmu-
lated uyhd‘ér ‘the condlhons “given in Table 3.1.
_ “"Since the efficiency was lower than desired, the
Two approaches have been investigated experi- system became limited by gas flow because of

mentally, and the best features of both are to be high pressure drops on the gas-supply system af

 
  

47

   

 
 

 

b et

 

 

s
4

o e e

 

 

NATURAL GAS
MANIFOLD

S £

 

 

Fig. 3.6. Gas-Fired Heat Source Test Installation.

48

 

 

 

 

 

UNCLASSIFIED .
PHOTO 22793
¥
v .
»~
 

2

Ky

[

  
 
 
 
 
  
  
 
 

h 5""'furo|
" weakening rather than by sample melting. The
weakening of the walls was probably caused by

 

TABLE 3.1. GAS-FIRED FURNACE OPERATING

 

 

CONDlTIONS
De;ign Aéqul
Sodium outlet femperofure, °F | o 1500 1500 )
VSOdIUm inlet femperofure, °F | 1100 1215 |
Sodium AT, °F 400 285
 Sodium flow, gprfi: S ‘ 7.3 8.7
Heat output, kw 100 85

o heat removed
Efficiency :

heat relecsed

X 100), % 66 55

 

the 85-kw level
and test operated at lower femperatures (1100 to
1400°F), higher efficiencies were obtained, not
only because of the increased temperature differ-
ential between-the flue gas and the sodium but
also because of the clean condition of the inside
surfaces of the furnace. " Over 100-kw output at
essentially design effucnency was attained through-
out operation in the lower temperature range, but,
because of an unsohsfoctory flame-failure .safety
system, sustained operation was impossible. After

the unsuhsfoctory condmon had been corrected,

continuous operahon ‘at 1500°F was resumed, and
it was found again that 55% was the best ef-
ficiency obtainable.

The furnace performed exceptionally well as a
compact, high-capacity heat source. At the
maximum operating condition of 166 kw total heat
released, the release rate based on the combustion
chamber volume was ~over 1.5 Mw/#t3, or over

5, 000 OOO Btu/hr/ff3

 

When 'rhe unn‘ was f!rst fired

It is fe[f that ’rhls heat-
'Vrelease rote is hlgh smce ull combushon was
probably not accompllshed |n “the 0.1-#3 volume“:“
‘ of ’rhe combushon chamber, qs ie"\l:denced by 'rhe“

 

 

PERIOD ENDING DECEMBER 10, 1954

local boiling of the sodium coolant, by accumu-

lation of a gas pocket in the line, by external

corrosion in the hlgh—temperature exhaust gases,
or by higher-than- predlcfed local gas- -side heat
transfer coefficients. The coolant tubes were
arranged in ten pora“ei passes that were mani-
folded into two ‘banks of five passes each. It
was observed in operation that the sodium outlet
temperature from one bank of five tubes was as
high as 1550°F, while that from the other was
only 1480°F. Temperature variation from tube to
tube within each bank of five tubes could con-
ceivably have been equally bad and have resulted
in hypothesized sodium temperatures of 1600 to

- 1650°F. The bmlmg point of sodium is 1620°F

at utmospherlc pressure and about 1800°F at
2 atm, and thus local boiling of the sodium,
together with consequent reduction in the local
heat transfer coefficient, is certainly a con-
ceivable possibility. Although it seems possible
that small amounts of gas were trapped in the
spirally wound tube coils d‘uting initial filling
with sodium, the location of the tube failures
makes it difficult to visualize the accumulation
of gas in stable gas pockets as contributing to
the failures.

Some external corrosion and erosion of the tubes
were observed, but there was no strong evidence
that they were advanced enough to have been the
cause of failure. The gas-side local maximum
heat " transfer coefficient could have been con-
siderably higher than the predicted value of 60
Btu/hr.ft2.°F because of nonuniformity of gas
flow and combustion, and burning could have taken
place on or very near to the tubes in question.

__This seems p055|b!e in view of the location of
_the fcn[ures relative fo the oufle‘r bends of the

tube c01| ‘mcxmfolds. These outlet tubes bend

_‘ichrectly across the chamber perpendlcular to and
__jns:de the lnner ‘row ‘of wound tubmg. In this
"_7‘_)4|ocaf|on, ’rhey could have’ acted ‘as flame stabi-
~ lizers for the uncombusted gas’ mixture “and ‘thus
__,_;_',_:caused hngh local coeffncuent_s :
V'T‘the first row behmd them. |'r_seems‘"probcfl:le that

 

“the tubes in

   
 

combma’rlon of non f!ld‘w ‘through

_fhe' ten parallei flow tubes and a hlghpr thcn-

pred:cfed local heat trqnsfer coeff:menf was the

-~ “cause of failure. ~The “pertinent ‘parts “of the as-

sembly are being examined by the Metaliurgy
Division in order to obtain more information re-
garding the failures,

49

 

 

c

4

 
s e

ot bt

b Bl i il

 

 

 

 

ANP QUARTERLY PROGRESS REPORT

Natural-Gas Burner

L. A. Mann

Aircraft Reactor Engineering Division

R. Curry
Pratt & Whitney Aircraft

The exploratory tests on gas burners, described

previously,® were used as a basis for new burner

developments. Designs for obtaining better mixing
of air and gas and higher pressures in the burning
chamber have been investigated, and some tests
have been made in which preheated air and gas
were used. The test units were based, in general,

~on designs developed by the Esso Marketers in
the past few years. |In these units, the burning

chamber serves also as a flame holder and a

" burning-rate controller. A schematic diagram of

the basic design is shown in Fig. 3.7.

The first burner that was built was operated
at approximately 3100°F and 400-kw heat release
without. a pressure-forming barrier at the exit of
the burner. It was then operated at about 3200°F
and 250-kw heat release at about I/3 atm pressure
drop from burner to air to simulate a heat ex-
changer after the burner.

A second burner was designed on the basis of
data from the first burner, plus specific provision
for preheating the air and gas. Tests were con-
ducted in which a temperature of about 3300°F
and 1-Mw heat release were obtained by using
unpreheated air and gas. Auxiliary tests indicated
that premixed air and gas fuel could be preheated
to 1600°F or more without serious preignition.

TRAP FOR FLUORIDE VAPORS

W. G. Cobb A. G. Grindell
W. R. Huntley

Aircraft Reactor Engineering Division

In operating a high-temperature system with a
molten fluoride salt containing appreciable quan-
tities of zirconium fluoride and blanketed with
inert gas, it is difficult to prevent plugging of
the gas lines leading to or from the vessel. A

.. __.vapor-trap device has therefore been developed
b".-_rlv\'rhkich provides a sufficiently large vapor con-
" :denser to prevent plugging in an operating period
~“i.of several thousand hours with reasonable gas
" movement (no excessive continuous bleed) and to
'f_e:_rix'a‘B'le adequate heating of the gas line between

 

6. A, Mann and L. F. Roy, ANP Quar. Prog. Rep.

" Sept. 10, 1954, ORNL-1771, p 41.

50

UNCLASSIFIED
ORNL-LR-DWG 4526

BURNED GAS

 

S Ss SY ]
e —
—_——
—_——————
e Ul

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

GAS-AIR MIXER

 

 

 

Fig. 3.7. Natural-Gas High-Intensity Burner,

the trap and the fluoride-containing vessel.

Gas line plugging with zirconium fluoride vapor
condensate occurred frequently at the pump tank
of the ARE component test at K-25, Plugging
was eliminated by heating the gas line to 1300°F
and by installing the vapor trap so that no cold
surfaces for vapor condensation were available
until the gases reached the large cross-sectional
area of the trap. This type of unit was first tested

Dyiv ey m
Py . 74
w NP F i

&

5
 

while connected to a small tank of relatively
. stagnant NaF-ZrF -UF, (53.5-40-6.5 mole %), at
1400°F. The unit operated for 500 hr, during
which time approximately 125 t3 of helium was
. bled into the tank and out through the vapor trap.
No plugging of the trap or of the Z—in.‘ gas lines
had occurred at termination of the test. Exami-
nation of the vapor trap and heated tube revealed
that all heated surfaces remained cledr and that

w?

C by

o

 

 

| PERIOD ENDING DECEMBER 10, 1954

condensation had occurred on only the cooled
surfaces. A second unit was then installed in
the test loop at K-25, and approximately 70 3
of vapor-laden helium was periodically vented
from the pump tank without plugging. The trap
has since operated over 900 hr with more than
100 periodic vent operations without plugging.
Similar traps installed in the ARE fuel pump were
also successful.

51

 

T T T T T ey

 

 
 

 

 

[T

"ANP QUARTERLY PROGRESS REPORT

4. CRITICAL EXP ERIMENTS

A D Calhhcm -

" Physics Dwns:on

REFLECTOR MODERATED REACTOR

ot B. L. Greenstreet
Alrcmff Reacfor Engmeermg Division

~{'-:?:'-"7 J J Lynn D. V. P. Williams
o F’hysms Division
, R. M. Spencer
_ Unn‘ed STafes Air Force
J S. Crude]e E. V. Sandin
.70 40 W. Noaks

o Pratt & Whitney Aircraft

The second of two critical assemblies of the
reflector-moderated reactor was constructed as a
part of a program presented previously.! The
reactor is composed of a beryllium reflector and a

" fuel region of enriched uranium metal and Teflon

(CF2)n. The first assembly, an essentially spheri-
cal fuel core surrounded by the reflector, was
described in the previous report.2  Since it con-
tained no extraneous structural materials, the
experimental results could be compared with the
results of multigroup reactor calculations. The
results of the second experiment were also to be
correlated with the calculations, since the second
assembly, too, was unpoisoned by structure, The
assembly has a central polyhedral core, or isfand,
of beryllium about 10 in. in diameter enclosed by a
4.5-in.-thick uranium-Teflon fuel annulus. The
fuel is surrounded by an effectively infinite be-
ryllium reflector.  Aluminum shells, 0.065 in.
thick, separate the fuel from the island and from
the reflector and provide structural stability, The
assembly is completely described elsewhere,® and
a summary of its composition is given in Table 4.1,
Figure 4,1 is a view of the mid-section.

The calculated? critical fuel loading was 7.16 kg

of U235, or 0,142 g of U235 per cubic centimeter

of fuel annulus. The first critical experimental
array of 7.39 kg at a density of 0.146 g of U233

 

ID. Scott and B. L. Greenstreet, ANP Quar. Prog.
Rep. Mar, 10, 1954, ORNL-1692, p 45.

2D, Scott et ale, ANP Quar, Prog. Rep. Sept. 10, 1954,
ORNL-1771, p 44.

'3B. L. Greenstreet, Reflector-Moderated Critical

' Assembly Experimental Program — Part 1I, ORNL CF-54.
i ,10-119 (Oct. 19, 1954).

W.V_E. Kinney, private communication, Oct. 25, 1954,

52

per ‘cubie centimeter of fuel cnnulus confomed

about 7.5% excess reactivity, In ‘order to com-
pletely fill the fuel region without excess control
rod poison, it was neceéssary to decrease the
loading to 4.66 kg of U235 at a density of 0.092 g

of U233 per cubic centimeter of fuel annulus. This
latter loading has been shown, by calibrated control

rods, to contain 0.92% excess reactivity, equivalent
to about 0.31 kg of U235, which gives 4.35 kg of
U235 as the critical mass of the unpoisoned
assembly, '

A series of measurements of neutron and fission-
rate distributions and of reactivity coefficients
will be made prior to the incorporation, in the
structure, of additional materials to simulate
reactor components,

TABLE 4.1. COMPOSITION OF SECOND
REFLECTOR-MODERATED REACTOR
CRITICAL ASSEMBLY

Beryllium Island
VYolume, ff3 0.37
Average radius, in. 5,18

Fuel Annulus {exclusive of aluminum
core shells and interface sheets)

Volume, 2 1.78 (50.4 liters)
Average inside radius, in. 5.18
Average outside radius, in, .57
Mass of (kg)
Teflon 99.38
Uranium 5,00
U235** 4.66
Uranium coating material 0.05
Scotch tape 0.12
Core Shells and Interface Sheets
Mass of aluminum, kg 5.85
Reflector
Volume, ft° 22,22
Minimum thickness, in. 11.5
Mass of (kg)
Beryllium 1155.0
Aluminum 29.2

*Density, 1,97 g of Teflon per cubic centimeter of fuel
annulus.
**Density, 0,092 g of U235 per cubic centimeter of fuel

annulus,

@
PERIOD ENDING DECEMBER 10, 1954

PHOTC 22782

 

 

 

 

-

 

 
 
 
 
 

ical Assembly.

i
™
o

 

deiutéd Reactor Cr

$Second Reflector-Mo

 

 

 

 

Flgo 4.10

 

-
™
*
<
5
&

 

 

o
 

 

 

 

 

 

e

 

 

 
 

 

G-

 

Part 1l

L

 

 
®

“t e

.

4

 

T TR

v

TN T

 

 

 
 

X e

 

 

=»

o fe,

A

Oua
1954 ORNL- 1771, Flg.5'| p55.

0RNb-1976

5. CHEMISTRY OF MOLTEN MATERIALS
~W. R, Grimes

7 Mdteriols Chemistry Division

Study of the NaF-ZrF ,-UF, system by the use
of various techniques, mcludmg quenching, thermal
'onalysm, and filtration, has continued. Improve-
ments that have been made in quenching methods
are expected to accelerate the availability of data
by this method. Phase studies of the UF,-bearing
fluoride systems have continued to be emphasized
because of inferest in their favorable corrosion
characteristics, It is believed that problems
attendant fo handling the easily oxidized UF,
will be alleviated to some extent by extendlng
phase studies to include LaF;, an isomorphous
stable substitute for UF,.

Investigations of the chemlcc[ reactions in molten
salts are conhnumg in an effort to understand the

" mechanisms mvolved in the purlflcahon of fluoride

mixtures and in ‘the reduction of UF, to UF3 in
fluoride melts. Slmulcr studies of the corrosion
products of Inconel in fluoride mixtures and of the
reduction of UF, by structural metals are under
way.

A sufficient stock pile of purified zirconium
fluoride-base fluoride mixtures will have been
‘prepared by the end of this calendar year to allow
termination of operation of the large-scale (250-1b
capacity) production foculr'ry. Careful control of
operating conditions in the pilot-scale focrln‘y has
allowed production of sufficiently consistent UF,
content material to permn‘ its release for corros;on
testing. Elecfrolyflc purn‘ncat;on of NaF- -ZrF ,

mixtures has been adopted, ‘and sugnlflcanf cuts_
_in processing “time have resulted, Howe

ot been posmble fo. CUf the processmg time Of»--m allows a temperature gradient to be imposed upon

a nickel capsule opproxumafely 4 in, in length that

   

 
  

77<UF -bearmg mlxtures by Using 'rhe elect OlY'”C:f
g "'purlflcaflon method,

 
   

ln‘:the fu,ndamental studies of fused salts,

 

ep.

P. A. Agron and M. A, Bredig, ANP Quar. Prog. Rep.
June 10, 1954, ORNL-1729, p 48.

- solutions.

    
  
 

     
   
  
 
 
  
   
  

values

fff»'.w\The apparatus consists of a round nicke
i Iengfh ‘and ]4 in. in dlametgr, wh|ch is )

ppotted in a MarshoH furnace by a mclcel tube
welded to the top. The capsules are supported

 

PHASE EQUILIBRIUM STUDIES

C. J. Barton
Materials Chemistry Division
H. Insley, Consultant

Solid Phase Studies in the NaF-ZrF -UF ; System

C. J. Barton R. E. Moore

L. M. Bratcher R. E. Thoma

Materials Chemistry Division
H. Insley, Consultant

Additional quenching data obtained with a number
of compositions in the NaF-ZrF -UF, system have
essentially confirmed previously reported findings.!
There was some indication from thermal analysis
of mixtures along the Na,U.F, ;-Na;Zr.F;; join
(formetly referred to as the NagUgF , +NaZrgF 4,
join — the 7 to 6 ratio was adopted on the basis
of structural considerations?) that these com-
ponents may not form a complete series of solid
Preliminary petrographic and xeray
diffraction studies of the slowly cooled melts lend
support to this view, but quenching data will be
obtained in order to assure that the crystalline
phases observed represent equilibrium conditions.
The perspective diagram shown in Fig. 5.1, pre-
pared by R. M. Freestone, gives a picture of the
liquidus relationships in this system,

Quenching studies in this system are expected

to proceed at an accelerated pace because of the
use of a new quenching apparatus similar to one

Fhas recently developed at Mound Laboratory (private
A%+ communication from E. Orban), The new appardtus

m‘ams a ‘number of compurfments, in each of

',whlch isa small quanhfy of the pan‘lcu!ar compo-
sition bemg mveshgafed Experlments wn‘h a_‘_w _
'50°C temperafure gradlent olong a capsule con~
S approxlmctely 20 compartmenfs have been
. carried out .

 

inside the block in a /16""‘ hole drilled from the
bottom to within 2 in, of the top. The capsule

 

 

 

 

   
    

| block

57

 

 

oot ol A ot b ol

 

 

 
 

ik,

 

ANP QUARTERLY PROGRESS REPORT

’ SEenET
ORNL-LR-DWG 3398

 

 

 

 

     

 

 

Koo
LoE
N >
N M
SN Y
T o O
v SN 3
= NN
o N
= P
. =2

Fig. 5.1. Phase Diagram of the NaF-ZrF -UF
System,

rests on a nickel wire support, which, in turn,
rests on a movable nickel platform at the bottom
of the hole. Temperatures are determined at 11
points along the sample length by means of thermo-
couples and can be varied by adjustment of the
various furnace windings. By moving the platform
to one side with a handle above the furnace, the
capsule and wire support are allowed to fall into
an oil quenching bath,

Pfidse-Rélutioqships in UF3-Bearing Systems
R. E. Moore R. E. Thoma
~ Materials Chemistry Division
H. Insley, Consultant

Study of UF;-bearing systems has been con-
tinved on NaF-ZrF ,-UF,, NaF-UF,, KF-UF,, and
Zrl::“-UF‘,-UF3 mixtures, In addition, work was
begun on NaF-LaF; and KF-LaF; systems, since

LaF, ‘is known to be a stable *'stand-in'’ for the

more difficultly handled UF,. Work on the KF-UF,
mixtures has revealed that a red phase, formerly
designated as K3UF,, may have as much as 75%
of the uranium in the tetravalent state; its apparent
homogeneity over a wide range of UST/U%Y is
possibly due to a solid solution of K;UF, and

58

 

K;UF,. The existence of the latter has yet to be
demonstrated. It is now thought that NaUF ,, rather
than Na,U,F,, as previously postulated, is the
empirical formula of an incongruently melting
compound in the NaF-UF, system.

UF, in ZrF -Bearing Systems. Studies of ZrF ,-
UF ,-UF, mixtures were reported earlier.3*4 The
more recent of the reports? indicated a probable
composition of 50 mole % ZrF4-.-25 mole % UF4—-25
mole % UF, for a brownish, slightly birefringent
phase noted in this system. Recent studies of
mixtures in this system that were prepared by
adding uranium metal to ZrF ,-UF, mixtures and
stitring the partially reduced melts until they
solidified indicated that the mixture with the above
composition consists of two phases, while a
mixture containing 33.3 mole % ZrF ~33.3 mole %
UF,-33.3 mole % UF, is essentially a single-
phase brownish crystalline material that is be-
lieved to be a compound. Another phase observed
at higher ZrF, concentrations, usually described
as olive-drab, appeared to be the only phase
present in a mixture containing 75 mole % ZrF ,—
12.5 mole % UF,-12.5 mole % UF,. All these
compositions are theoretical compositions based
upon complete reaction between UF, and uranivm
metal in the molten material. The meager analyti-
cal data available for such mixtures indicate that
the reactions probably did not go to completion in
all cases, and so the percentage of UF is likely
to be higher and the percentage of UF; lower than
are shown by the formula.

In the NaF-ZrF -UF; system, four compositions
wete prepared along a join connecting the minimum
melting compositions in the NaF-ZrF, binary (42
mole % ZrF ,; mp, 500°C) and the NaF-UF; binary
{29 mole % UF,; mp, 715°C). Liquidus tempera-
tures determined by thermal analysis for the
mixtures containing 5, 12, 20, and 25 mole % UF,
were 650, 775, 865, and 775°C, respectively., The
25 mole % UF, mixture contained free UF,, in
addition to the NaF-UF; complex and Na,ZrF .

KF-UF; System. Earlier studies of KF-UF,
mixtures5 indicated the existence of two com-
plexes. A red crystalline phase observed as the

 

3V. S. Coleman, C. J. Barton, and T. N. McVay, ANP
Quar. Prog. Rep. June 10, 1953, ORNL-1556, p 41.

4C. J. Barton et al., ANP Quar. Prog. Rep. Sept. 10,
1953, ORNL-1609, p 57.

3¢C. J. Barton et al., ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL-1771, p 59. '

 
 

 

. principal phase in mixtures containing 75 mole %
2 KF was believed to be K;UF, while a blue phase
; " with optical properties snmllar to those of the blue
NaF-UF, complex was assigned the formula
: K3U,Fg. It was also previously reported® that all
| KF-bearing reduced-uranium mixtures that were
analyzed chemically contained varying quantities
of tetravalent uranium‘, usually in a form that could
hot be recognized as such either petrographlcolly
or by x-ay diffraction analysis, Some progress
was made durmg the past quarter in gaining a
better undersfondmg of phase relationships in this
KF system, Investlgahon of the KF-UF, system
has been considerably aided by begmn:ng the
study of the KF- -LaF, system, The compound
LaF, is isomorphous with UF,, has almost identi-
cal lattice dimensions, and, furthermore, has only
one stable valence state (3+) whereas UF
maintained with dn‘flculty. Dergunov’ published
partial phqse diagrams for the alkali fluoride~LaF,
sysfems, and Zocharncsen8 has studied mtxtures
of KF and NaF with LaF; by an x-ray diffraction
technique. Both w°rkers indicated only a single
compound, KLaF,, formed by these two compo-
nents, According to Dergunov, it melts incongru-

Y

“+

& ently ot 770°C, It seems probable, therefore, that
the compound that was called K3U,Fy was actually

S . KUF,, with an incongruent melhng point of
- 820 i 10°C, Tests performed duting the past

quarter demonstrated that the red crystalline phase
formerly designated K,UF, can have as much as
75% of the uranium in fhe 'retravalenf state. Since
this phase is usually cubic, as is K3UF7, and
has a refractive index near that of K 4UF,, the

most hkely explcnahon of the h mogeneity of the
“3+/U4+_ ratios
77560 to 500°C and helps to explain why it is dif-
* ficult to detect ||qu1dus temperatures in mixtures
" of this type by thermal cnqums., It is probable
that the solublhfy of UF,
éutectic is htgh enough fo provrde a usable fuel
"";"""*sysfem for a reflector-moderated c:rculcflng-fuel
reocfor. '

 
  

rred phase over a wi

 
  
  
 

 

  

- K,UF,, a <ompound no
‘s _fOfmed TIR YRR ER

  

  
 
 
  

B

 

Ibzd p 57 )

o TEL P Dergunov, Doklady Akad. Nauk S.5.8.R. 60,
o 1185 (1948).

- 8W. H. Zachariasen, J. Am. Chem. Soc. 70, 2147

- {1948). ‘

 

“mole %) were reported earlier.

__seems to be that a _solid s‘oluhon oz(:l(:{]UF7 andf"w_r
‘tg____gxisf, o
' reported a Cs. LaF;s com~
spdiidmg potasswm compound" o

PERIOD ENDING DECEMBER 10, 1954

phases or a recognizable UF, complex in addition

to the KUF

NaF-UF; Systems. A partial phcse diagram for
the NaF- UF system was presented prewously.
No addltlonai data have been obtained since that
time, but available information on the NaF- LoF
sysi‘emIo 11 indicates that the incongruently melf-
ing compound in this UF, binary system is probably
NaUF , rather than NasU,Fo.  Since UF, is the
primary phase that separates from the melt at the
50-50 composition and since UF, has a higher
density than the liquid, it is likely that the earlier
identification of compound composition was in
error because of lack of equilibrium conditions on
cooling.

Solubility of UF, in NaF-KF-LiF Eutectic
R. J. Sheil

Materials Chemistry Division

Thermal analysis data for UF, dissolved in the
NaF-KF-LiF eutectic composition (11.5-46.5-42.0
12 5 view of cutrent
interest in alkali fluoride fuels, it appeared desir-
able to check the earlier data with newer tech-
niques. A purified mixture containing 4 mole %
UF, was heated in a plastic inert atmosphere box
s:mllar to the one described previously.1® The
liquidus temperature for this mixture was deter-
mined by visual observation to be 560 £ 10°C, The
mixture was cooled to 500°C, and a sample of
liquid was filtered through a nickel filter stick,
Chemical analysis showed that 17.6 wt % uranium
was in solution at 500°C as compared with 18.3
wt % in the unfiltered material. This shows that
only 0.2 mole % UF4 precipitated in cooling from

in the NaF- KF-LiF

e T SO S

" Ibzd, p 59, Fig. 5.4.

10 Dergunov, op. czt

 

Zacharlusen op czt

125 p. Blakely, L. M. Bratcher, and C. J. Barton,
ANP Quar. Prog. Rep. Dec. 10, 1951, ORNL-1170, p 80.

By, s. Grim, ANP Quar. Prog. Rep. Sept. 10, 1954,
ORNL-1771, p 56,

59

 

 

 

 
 

 

 

. A L S S

ANP QUARTERLY PROGRESS REPORT

UF; Stability
W. C. Whitley ~~ R. J. Sheil
- Materials Chemistry Division

Du‘_ring' the past quarter studies have been under

‘way on the stability of UF, separately and mixed

with KF. Disproportionation of UF; in KF in

Inconel at 1000°C has been established. When

't:‘c:;h’rd'ined alone in Inconel, UF, does not detect-
ably disproportionate below 1250°C.,
Stability of UF; at Elevated Temperatures. The

'-dlsproporflonqtlon of UF; as a function of temper-

ature was studied eorher under high vacuum.

‘Recent availability of high-purity UF,; made de-

sirable a study of the high-temperature stability
of UF; in a helium atmosphere. For these experi-
ments 25-g samples of UF, that contained 99.4%

of the uranium in the trivalent form were sealed

by welding in Inconel tubes containing '% atm of
helium. After being heated at the desired temper-
ature for 1 hr, the tubes were removed from the
furnace and were dair quenched to minimize re-
combination of UF ; and uranium metal. After the
UF, samples were cooled to room temperature and
ground in an inert atmosphere, they were examined
petrographically, by x-ray diffraction, and by
chemical analysis. The sample heated to 1200°C
showed no evidence of disproportionation, the
sample heated to 1250°C showed very slight
evidence, and the sample heated to 1300°C showed
definite evidence. [n addition tfo the increase in
UF_4 content noted Jn the petrographic examination,
an increase in U4* was shown by chemical analy-
sis (from 0.6 to 1.6%), and the presence of uranium
in the tube walls that had been in contact with the
UF3.was established. These data show that UF,
is more sfable at elevated temperatures than
estimates of its thermodynamic properties !4 12
had indicated. [t is interesting to note that the
sample of UF; heated to 1300°C was practically
insoluble in b0|||ng 10 N HCI, whereas samples

" heated to lower temperatures dissolved readily in

this solvent. This could indicate that either a
sintering action occurred, which takes place in
oxide systems well below the melting point, or

 

]4L. Brewer et al.,, Thermodynamic Properties and

- Equilibria at High Temperatures of Uranium Halides,

Oxides, Nitrides, and Carbides, MDDC-1543 (Sept. 20,

1945, rev. Apr. 1, 1947).

15A Glassner, A Survey of the Free Energies of

- Formation of the Fluorzdes, Chlorides, and Oxides of
the Elements to 2500°K, ANL-5107 (Oct. 22, 1953).

60

that the small amount of liquid UF, formed at the
high temperature exerted a cementing action of the
UF, crystals on solidifying.

Stability of UF; in KF. An attempt was made to
determine the stability of UF in molten KF as a
function of temperature by starting with dry KF
and high purity UF, (99.4%). Mixtures of KF and
UF, in the molar ratios of KF to UF, of 3.0, 1.5,
aond 1.0 were mixed, sealed in Inconel capsules
containing ]é atm of helium, and heated for 2 hr
at various temperatures before air quenching. A
period of two weeks elapsed between the time of
the experiment and the time that the samples were
submitted for chemical analysis for total uranium
and U3* and thus there is some doubt of the
significance of the U3* determinations. However,
both the total uranium and the trivalent uranium
values exhibited a downward trend with increasing
temperature in the temperature range 750 to 1050°C
for mixtures having a 3 to 1 KF-UF, molar ratio.
An average of 7 moles of UF. disappeared per
gram-atom of uranium lost, presumably by alloying
of the uranium metal with the Inconel capsule
walls, Mixtures having 1.5 to 1 and 1 to 1 KF-UF,
molar ratios were heated to 850 and 1050°C, Analy-
ses of these melts also showed the downward trend
in U3* and total uranium values with increasing
temperature. In these samples, however, the ratio
of moles of UF, (or gram-atoms of trivalent ura~
nium) to gram-atoms of uranium lost averaged about
5 as compared with the theoretical value of 4. The
capsules used to contain these materials were not
analyzed for metallic uranium; however, a similar
experiment was performed in which an 85 mole %
KF-15 mole % UF, mixture was heated to 1000°C
in an Inconel capsule for 90 min, The melt was
sampled by drilling after the top and bottom of the
capsule had been cut off. The remainder of the
melt was dissolved out, and the walls were care-
fully cleaned mechanically; then, approximately
7 mils of the tube wall that had been in contact
with the melt was removed by drilling. Flashes
of light observed during drilling gave a qualitative
indication of the presence of uranium in the tube
walls which was confirmed by chemical analysis,
which showed 17.9 wt % uranium. The total of
0.28 g of uranium that was recovered from the tube
walls represented 3.6% of the total uranium con-
tained in the KF-UF; mixture. Analysis of the
melt showed 70% of the total uranium to be in
the trivalent state. It seems certain that dispro-

5
7
N
 

“)

*

*

portionation of UF, occurred at a lower tempera-
ture than would have been expected in the absence
of KF. Further experimentation is planned to
determine the stability of UF, in KF and other
solvents as a function of temperature and of con-
tainer material, Preliminary experiments indicate
that UF; may be more stable when dissolved in
KF in a plchnum container than in a mckel cone
tainer,

-Differenii-al» Thermal Analyfsis”
C. J. Barton

Materials Chemistry D|V|S|on
D. L. Stockton

Merck and Company, Inc,

As a result of the high frequency of failures ex-
perienced with nickel capsules containing NaF-UF ,
samples, a new system was designed for containing
samples during differential thermal analysis.
Graphite containers constructed so that the thermo-
couple junctions can be immersed directly in the
melt are now being used, The new system is
designed to permit evacuation before mfroduc’non
of helium as the inert atmosphere,

A cylindrical graphite block (378 in, in diameter
and 6 in. high) serves as a heat sink inside a
nickel reaction vessel (4 in, in diameter and 7 in,
deep) with a flanged head. Six graphite thimbles,
l/2 in, iD, 3'/4 in. OD, and 3 in. deep, are spaced

on a hexagonal artay inside the upper half of the

graphite block. Each thimble is filled to a depth
of about 1 in., with either sample (about 6 g) or
reference material, Al,0,; the sample depth is
reduced by about 50% upon melting.

A /-ln.-dia nicke!l tube which contains two |
Chromel-Alumel thermocouples contained in a 7
single four-hole porcelam insulator extends m'fo_': '_-“where P is expressed in atmospheres and C in
each scmple ‘thimble,  The fhermocouples possfl

through a Kovar seal to allow evacuation.

. The. six nic

- 5',"":_”nected to t}ie flonged top by qugelok flflmgs.
in oddl’rlon, “the head 'has three /-|n. Swagelok
| flttmgs f'or vacuum-Venhng hehum, oncl a pressure
indicator, Three posfs support a circular plashc:'aw")
_ring, whlch serves as a 24-pomf gcck-ponel ‘The
'cenfer sechon__of the posfs is /-m. fublng 'ro"’

 

mainfained’ af room temperufure. Prehmmory tests -

with this CIppCIl’(fl‘US ‘indicate that it may be dIHICU”;wM‘ReP Sept. 10, 1954, ORNL-1771, p 64.

to obtain a sufficiently vacuum-tight assembly to
avoid oxidation of UF ,. It has shown very high

 

 
   

-'prewously descrlbed procedure. |

PERIOD ENDING DECEMBER 10, 1954

sensitivity as compared with the nickel capsule
technique and may be very useful if the oxidation
difficulty can be overcome. Further expetimen-
tation with this apparatus and with simplified
equipment that will permit the use of smaller
samples is in progress.

CHEMICAL REACTIONS IN MOLTEN SALTS

F. F. Blankenship L. G. Overholser
W. R. Grimes

Materials Chemistry Division
Reduction of Fer by H, in NaF-ZrF, Systems
C. M. Blood

Materials Chemistry Division

~ The reduction of FeF, with hydrogen is of par-
ticular significance in considering the time neces-
sary for purification of NaZrF . mixtures, as
discussed previously,1® and has therefore been
studied in considerable detail. The reaction

FeF, + H,= Fe® + 2HF

was previously explored by a dynamic method and
by an equilibration method that actually determined
equilibrium conditions of the system. During the
past quatter the dynamic method was employed
further,

Measurements of the equilibrium HF pressures
from the reduction of FeF, in NaZrF 5 by hydrogen
gave, as a preliminary result, an equilibrium
constant, K, of 0.2 at 600°C as compared with the
value of 2.0 at 800°C previously reported.}® The
equilibrium constant is calculated from the equation

Pir
K. = ——MmMm
' Plz'i g * CFeF

- mole frachons.

Reducflon of UF to UF in Fluoride Melss
H A Frledmcn

Materlqls Chemtsfry Dw:snon

Determ:nahons of the effect of reoct:on con-

- ditions on the extent of reduct:on of UF4 to UF
< in fluoride melts hQVe continved by the use of a

17 To defermme

 

'6C. M. Blood and G M Watson, ANP Quar Prog

e, M. Bload et al., ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL-1771, p 77.

61

 

 
 

ANP QUARTERLY PROGRESS REPORT’

whether o greater percentage of UF, could be
reduced to UF, in NaF-LiF-KF (11.5-46,5-42.0
- mole %) at a Iower temperature than previously
used, a trial was made in nickel at 500°C, After
2 br of equilibration of the alkali fluoride with
sufficient UF, to equal 25 wt % uranium as UF,
' _'and 100% excess uranium metal, the mixture could
" not be filtered, and no quadrivalent uranium was
S deitec_t_able by petrographic or x-ray analyses, Thus
complete reduttiofi‘ of UF, to UF, was indicated.

‘A similar experiment mvolved equnhbrcfmg the
NOF LIF eutectic (40-60 mole %) in nickel for 2 hr
at 725°C with added UF, and an excess of ura-
nium metal, Chemlcal cmoly5|s revealed that
all the uranium present after the test (25 wt %)
was in the frivalent form, When the same type of
experlment was catried out with the LiF-KF
eutectic (50-50 mole %) at 600°C, only 60% of
uranium was found to have been reduced. Thus
KF appears to differ from NaF and LiF as a
_solvent for UF, and UF,.

A trace of volatilized alkali metal was the only
evidence of reduction when purified KF was heated
at 900°C for 2 hr with finely divided vranium metal.
A similar result was noted when NaF-KF-LiF

(11.5-42.0-46.5 mole %) was heated to 700°C with
uranium metal for 4 hr with gas stirring, Dispro-
porfionation of UF, is therefore seen as a reason-
able explanation for the quadrivalent uranium found
in UF ;-KF mixtures.

Heating of high-purity UF; with the NaF-KF-LiF
mixture at 800°C yielded material which, when
filtered, was found to have 73% of the uranium in
the filtrate, as well as all the uranium in the
residue, present as U3*, Evidence of slight vola-
tilization of the alkali metal precludes a definite
conclusion as to whether reduction of KF or dispro-
portionation of UF, was responsible for the de-
tected UF,; however, the presence of uranium
metal also in the residue indicates strongly that
some disproportionation took place.

Electrochemistry of Fused Salts

L. E. Topol
Materials Chemis’rry Division

Decomposition potential measurements of molten
KCIl and of chromous and chromic chlorides dis-
solved in KCI have been continued. These elec-
trolyses, reported in Table 5.1, conclude, for the
present, the decomposition study in fused KCI,

“TABLE 5.1. DECOMPOSITION POTENTIALS OF VARIOUS CHLORIDES IN KCI AT 850°C

 

 

 

 

Anode: Carbon
~ Salt in KCI*
Salt @it in i Container Cathode Atmosphere E (observed) (v)
e {mole %)
KCl** 100 Platinum Crucible He 3.25 t0 3.30
100 Platinum Crucible He 2,90 to 2,75
CrCi3 3.0 Morganite Al,0, Platinum He 0.55 fo 0.65, 0.30 to 0.36, 0.90 to
1.05, 1.651t0 1.80, 0
33.3 Morganite Al,O;  Nickel He 1.60, 0.70 to 0.80, 1.25 to 1.35
33.3 Morganite Al,O,  Platinum He 0.45 to 0.50, 1.04 to 1.15, 1.25,
1.45, 1.57, 0
33.3 Morganite A|203 Platinum H2 0.98 to 1.03, 1.22 -
 GeCly (old) 3.0 Morganite Al,O,  Nickel ) 1.06, 1.45 to 1.50, 1.75
 (frosh) 3.0 Morganite Al,O,  Nickel ) 1.70 to 1.80, 1.15 to 1.25
(o!d) ) © 3.0 Morganite A1203 Nickel He 1.20 to 1.45
{fresh) 3.0 Morganite Al,0,  Nickel He 1.16, 1.67 to 1.75, 1.35 to 1.65
. ) 33.3 _ Morganite Al 295 Nickel He 0.90 to 0.98

 

*These concenfrahons were made by mixing together the appropriate weights of the salts.

. **Th:s run ‘was made with a slotted Morganite crucible inserted between the two electrodes as a dlcphrugm

62

3
™
g
5
5

L
o
o,
o
S
o
.
il

v %
 

 

#

In the elecfrolyses of KCI the vormbie in the
two experlmem‘s was the insertion of a “‘dia-
phragm’’ between the electrodes in one of the runs,
The diaphragm could produce the following effects:
first, aHHough the surface areas of the electrodes
involved were the same in both runs, the interpo-
sition of an msulahng crucible with a small hole
through which all the current must flow could
result in a smaller effective surface area for the
cathode. Thus, despite the maximum current in
each cell being the same (1 amp), the current
density of the unseparated cell might not have
been large enough to extrapolate to a true value
of E, the decomposition potential. Second, and

- more likely, the lower voltages obtained in the

cell without the diaphragm were due to interaction
of the electrode products at one of the electrodes.

1t should be noted also that the decomposition

po'fenhols measured in this case decreased with
time. In other runs with KCl for which small
cothodes were ‘used, for example, rods or wires,
decomposmon pofenhols of around 3.25 v were
found, with or without the use of a dlaphragm.
This anofnaly may be due to a greater evolufion

~of gaseous potassium at an electrode of small

surface area, whereas a thin layer of pofassium
formed over a large area may go into solution more
readily. | | -

The chlorides of chromium may undergo the fol-
lowing elecfrochemlcal reactions at inert elec-

 

trodes:

(1) CeCl, = Cr + Cly, E° = 1.36v,
(2) 3CeCly = 2CeCly + Cr, E° = 0.84 v,
(3) 2CrCl, = 2Cr + 3Cl,, E° =~ 1.08 v,
M)ZQCI 2ocb.%c5f'"*“““”%

  

   

'18

  

 

and for scmples ’rho} had been‘contomed m hghfly;-::‘::&
. : o Rep Sept. 10, 1954, ORNL-1771, p 63.

 

 

18 E. Topol ANP Quar Prog. Rep Mar 10, 1954,
ORNL 1692, P 63.

   
   
 
 
  
 

 

  
  

. trivalent chromlum

(K NaCrF ) are observed in 1F\e cooied mel'rs.
] CrCI3 soluhons lng'

The compound.; ;_omques /very similar fo those_descrlbed prenously

  
 
  

 

 

 

PERIOD ENDING DECEMBER 10, 1954

sealed containers for several days. Although there
are differences in the measured decomposition
potentials, it is difficult to explain them and the
wide ranges of values found. Thete is no evidence
of dilute solutions of CrCl, undergoing reaction
2 electrolytically, although thermal disproportion-
ation cannot be ruled out, There are similar varied
and divergent values of E for both dilute and
concentrated CrCl, solutions in an inert atmos-
phere. In this case there is evidence of reaction
4 occutring in solutions of various concentrations.
At the end of most of the runs, current-voltage
curves intersected the origin and indicated the
anode and ccthode reactions to be equal and
opposn'e

Cr3* + e= = Gt .

In every case a deposit of metailic chromium was
found on the cathode.

Although decomposition potential measurements
have been terminated, electrochemical studies in
fused salts are to be continued and will include
emf measurements, primarily in fused fluorides,
with special emphasis to be placed on finding a
reversible fluoride electrode.

Stability of Chromous and Ferrous Fluorides
in Molten Fluorides

J. D. Redman C. F. Weaver
Materials Chemistry Division

The compound ‘Cer or some complex compound
of divalent chromium is quite stable in NaZrF,
melts. When CrF is added to such melts, it is
reduced to the dlvalenf state even by metals as
noble as mckel this reduction is, apparently,
qo'onflfahve.w These observations are in accord
with the known fact ’rha’r CrF is the corrosion

pro-ducf wh 'N”F;ZrF -UF ‘melts are tested in

 

    
  

 

_ ,\_olloys of chromium.

when NaF KF LlF UF "melts are

il Vc:l[oys, ‘the reactlon product is

and complex fluochromates

However, 7

 

 

Recent sfudles made wnth uppqrc’rus ‘and tech-

in _connection ‘with F,-Cr® reaction?? have
l‘ndlcoted ’rha’r ’rhls behovnor must be expected

  

 

19.] D Redman and C, F, Weaver, ANP Quar. Prog.

L G. Overholser, J. D. Redman, and C. F, Weaver,
ANP Quar. Prog. Rep. Mar. 10, 1954, ORNL.-1692, p 56.

D507 N7 63
! . s 2
it £ P

v 3 (RGN i

 

e L AR T e e

 

 
b i e i P i

b, i

e Lol e T L

ANP QUARTERLY PROGRESS REPORT

In these experiments, pure preparations of CrF
or CrF, were added to the NaF-KF-LiF eutectic

in apparatus of nickel, maintained at the indicated

" temperature for 5 hr, and filtered. Data obtained

by qndlysis of the filtrates are shown in Table

52

TABLE s, 2 STABILITY OF CrF AND CrF
' MOLTEN NnF LiF- KF

 

Found in Filtrate

 

' ‘Ter‘npe‘ramr‘er
.; A_ddl-tfv,e, (°C) et o3t N2t
e o (ppm) (%) (ppm)
ch equnva[enf 600 120 1.08 260
to %8 wi % Cr3+ | 520 0.74 40
' 90  0.91 60
370 1.00 820
800 600 3.63 15
o 430 4,22 30
480 4.22 30
440  3.60
ce2t
(%)
CI‘F equ:vaienf 600  0.39  1.12 40
to 5 8 wt % Cr2t 0.33 0.86 50

S ' 0.32  0.87 30
' .10 1.1 50

800  1.05 2.97 40
0.94 3.30 30
0.99 3.24 90
0.98 3.35 20

 

 

 

These data indicate that CrF, in NaF-KF-LiF
eutectic is quite stable toward reduction by
nickel. In the most favorable case, less than 10%
of the CrF; appears as Cr**. The solubility of

~trivalent’ chromium seems to be at least 10,000

ppm af 600°C and at least 40,000 ppm at 800°C;

the solid phases observed in the cooled melts
were complex fluochromates.

When CrF2 is the additive, however, the situ-
ation is entirely different. Most of the dissolved
chromium is trivalent; complex fluochromates are
observed in the cooled filtrate. The amount of
material added was such as to exceed the solu-
bility of CrF, ot 600°C; so at that temperature
no material balance is possible. At 800°C, how-

ever, the solubility limit was, apparently, not

exceeded, and the concentrations found corre-
sponded extremely well to those to be expected
if chromium metal were not soluble in the melt
and if the reaction

3CrF,==2CF, + C°

had occurred. Published figures for AF for CrF,
and CrF, show that the standard free energy for
the recctlon, as written, is +14 kcal and suggest
that

a2C F .
r . -
K = —_2 - 10-3 at 800°C .

eq 3
a CrF2

This disproportionation must be considered to be
surprising. f K is to be reconciled with the
observed value, the activity coefficient of CrF
in the melt must be less 'rhan_]O" . Addmona?i
data at various temperatures and concentration
levels to be obtained in the near future should
afford a more accurate estimation of the equi-
librium constant for this reaction.

When ferrous fluoride is added to the NaF-KF-
LiF mixture in equipment of nickel, no appreci-
able reaction is observed. The solubility of FeF
appears to be 12 wi % at 600°C and 19 wt % at
800°C, and it appears that FeF, is dispropor-
tionated very slightly, if at all.

If ferric fluoride is added to NaF-KF-LiF in
equipment of nickel, considerable quantities of
NiF, are observed in the filtrate, and thus re-
duction of the Fe3* by nickel metal is indicated.
However, the experiments made to date have not
afforded reasonable material balances between
NiF,, Fer, and FeF3 in the system. Additional
data will be required before this reaction can be
evaluated.

Reduction of UF , by Structural Metals

J. D. Redman C. F. Weaver
Materials Chemistry Division

The apparatus and techniques for experimental
determination of equilibrium constants for the
reactions

Cr° + ZUF4 = CrF2 + 2UF3
and
Fe® + 2UF4 = CrF, + 2UF3
% ey vy '\m
UEdT 068

LMy
RGeS g, Shiis

 

i

have been described in previous quorterly re-
ports.2!  Equilibrium data have been shown for
the case in which the pure metals have served as
the reducing agents and the solvent was NaZrFs.

By application of identical techniques, ad-
ditional data have been obtained on the UF ,-Fe°
reaction, for reduction of UF, by Inconel, and
for reactions of UF, with the pure metals in the
NaF-KF-LiF eutectic.

The results of some additional studies on the
system . : -
Fe® + 2UF, = 2UF, + FeF,
in NaF-ZrF, at 600 and 800°C are given in Table
5.3. In fhese experiments, 2 g of hydrogen-fired
iron wire was reacted with UF, in about 40 g of

NaF- ZrF4 The values for K K, reported are defined
as o
| c2. C
UF, “FeF,
K = ,
x Cz

where fhe concentrutlons are given in mole frac-
tions with the NaF- ZrF content considered as
equal to 2 moles. The vaiues for the UF; concen-

 

2].1 D. Redman and C. F. Weaver, ANP Qfiar. Prog.
Rep Sept. 10, 1954 ORNL.- '|77'i, p60

TABLE 5.3. EQUILIBR]UM DATA FOR THE SYSTEM
CFe® + 2UF, = 2UF; + FeF, IN MOLTEN NuF-ZrF4
. AT 600 AND 800°C

' UF4 Ad'd"ed.: 0.360 mole/kg ‘o'f melt |

 

- Concentration* of
Temperqfure Fett in Filtrate K

Lemps - x
€ oo (epm) i

 

  

 

PERIOD ENDING DECEMBER 10, 1954

trations have been calculated from the Fe'f

values found for the filtrates, since a reliable
method for determining UF, in this mixture is not
available.

It will be noted that the values given for K
at 600°C are larger than those at 800°C, which
is in agreement with all previous measurements.
The values at the respective temperatures are
also in fair agreement with those given earlier.

The reaction between UF, and chromium in
Inconel with NaF-ZrF, as solvent has been in-
vestigated at 800°C. In these experiments, 2 g
of Inconel {hydrogen fired at 1200°C) was con-
tacted with the melt for 5 hr prior to filtration.
The results are given in Table 5.4.

The values given in Table 5.4 for the equi-
librium constant are lower by about a fivefold
factor than those obtained when chromium metal
was used. |t is possible that the activity of
chromium metal in Inconel is appreciably less
than its mole fraction. It appears more likely,
however, that these experiments represent ‘‘equi-
librium'' with Inconel whose surface layer is
depleted of chromium metal. Additional data will
be obtained by using longer equilibration times
and varying surface-to-volume ratios.

A study of equilibria for reactions of Fe° and
Ce® with UF, in the NaF-KF-LiF eutectic has
been started. It has been observed that when
Fe® is equilibrated with the NaF-KF-LiF mixture
alone about 100 ppm of Fe (presumably Fe**) is
found in the filtrate. This is very close to the

"TABLE 5.4, DATA FOR THE REACTION
Cr (INCONEL) + 2UF, = 2UF; + CrF, AT 800°C IM
"‘MOLTEN NaF-ZrF

UF 4 Added: 0.360 mole/kg of melt

 

_ Concentration™ of

 

 

   
 

o cr™ in Filtrate K, **
s (ppm) '
B I
6QSfl 4,_;2 ¥ 10_5 :
= e e s < 10° 5 :

 

”“'\'*Blank of 190 ppm

 

*Blank of 100 ppm to be subtracted from determined
values in caleulations,

 

  
 

cted from determined
"values in calculations.

**Activity of chromium in Inconel assumed to be equal
to mole fraction (0.2).

;”E ‘:-‘f 5*"*[? sy »*»\ (»\ 65
- i

 

 
 

 

 

 

~ ANP QUARTERLY PROGRESS REPORT

'- 'vulue obtcnned by equdlbrctlon of Fe® with NaZrF

“in similar equipment. However, equilibration of
Cr° with the alkali fluoride eutectic produces
- about 500 ppm of Cr in solution, and, based on

" the fmdlngs reported in the preceding section,

i this is, presumably, a mixture of Cr3* and Cr++.
_ This high concentration of soluble chromium in
- the NaF KF L|F sys’rem is surprising, since about
- ]00 ppm of Crt* is obtcnned in a similar equi-
[tbrchon with chZrF5 The reaction responsible

| ff'_for ‘I‘his h|gh' “blonk" value is not yet known.

Prel:mlnary data for the reaction of UF, with

Cr_d in 1[’1[5 solvent indicate that at 300°C about

© 2500 ppm of Cr is presem‘ in the equilibrium
filtrate; this is nearly the same conceniration
that resu[’rs in Nc:zZr]""5 At 600°C, however, about
1200 ppm ‘of Cr is recovered in the filtrate. I
,'appears, ’rherefore, fhaf fhe complex reaction in

N NaF- KF LiF is much more femperature sensitive
B fhan i

s

    
 

lmple reacflon in NaZrF .
PRODUCTION OF PURIFIED MOLTEN
FLUORIDES

F F. Blankenshlp G. J. Nessle
" L. G. Overholser
“Materials Chemlsfry Division

Elecfrolytuc Purification of Zirconium-Base

Fluorldes
C M. Biood H. A. Friedman
F. W. Miles F. P. Boody

Materials Chemistry Division

Interest in the extremely short process time and
high efficiency obtained in the electrolytic purifi-
cation of molten NaF-ZrF, mixtures led to an ex-
ploration of the limitations of electrolysis as a
means of purification. At cathode current densities
up to 0.3 amp/cm?, the presence of UF, in a salt
mixture was found to greatly reduce the efficiency
of electrolysis because the reduction of UF
UF, at the cathode was counteracted by the
oxidation of UF; to UF, at the anode. A steady
* state was approached wi\en about 5% of the UF
* had been converted to UF ;. Apparently the rapid
':reduchons obfomed with NaF-ZrF | mixtures con-
k:'ralmnrg no UF4 were due, in large part, to the

- electrolytic precipitation of zirconium metal on
- the cathode and to the removal of fluorine as HF
.‘,‘_‘-f;ff.‘-:when 'H, was used as the stripping gas or as
jffluorocorbons when He was used. The reduced
“--77,,-_7,:metall|c |mpurmes were precipitafed in the

 

cathode deposit. Any structural metal ions which
were not deposited as a primary electrode process
were precipitated by reaction with the zirconium
metal. |

Electrolysis was most effective in purification
when two advantages were achieved: first, a rapid
rate of over-all reduction of the melt, or a good
current efficiency, as measured by a high rate
of removal of HF or fluorocarbons, and second,
the deposition of impurities as an adherent cathode
coat which could be removed from the cell. In
NaF-LiF-KF mixtures the electrolytically reduced
potassium was slightly soluble in the melt, and
therefore oxidation of dissolved potassium at the
anode nullified reduction at the cathode and the
current efficiency for over-all reduction was low.
Also, an adherent cathode deposit was not
regularly found.  Presumably, both these ad-
vantages could be regained, not only for the

NaF-LiF-KF system but for UF ,-containing melts

as well, by alterations in cell design, electrode
construction, and current densities.

Some of the incidental points noted were that
0=~ and SO,~~ could be removed by electrolysis
in NaF-ZrF4 melts, and hence HF treatment is
not necessarily required in the purification of salt
mixtures. Platinum was unsuitable as an anode
material. The NaF-LiF-KF mixtures were par-
ticularly prone to show an ‘‘anode effect,”” or gas
polarization of the graphite anode, when there
was a slight contamination by air. The UF -UF,
oxidation-reduction system showed no poiurrzc’rlon
at graphite anodes or nickel cathodes.

The electrolyses were carried out in apparatus
of the type previously described.2?2 The graphite
anodes were 1.12 em in diameter and were im-
mersed to a depth of about 10 ¢cm to give an area
of about 25 cm?2. The cathodes were constructed
of a ¥-in.-dia cylinder of nickel gauze having a
calculated immersed area of about 70 cm2. The
wet areas of the pot and of the probes which were
occasionally used as electrodes were 500 and
10 cm?, respectively. Ordinarily, 10 amp was
passed during any continuous electrolysis step.

Preliminary results with SO,~~ removal were
sufficiently encouraging that an electrolysis trial
with helium as the sweep gas was carried out on
a 3-kg batch of NaF- ZrF, (53-47 mole %) which

had received no prewous purification other than

 

22¢, M. Blood et al.,, ANP Quar, Prog. Rep. Sept. 10,

1954, ORNL-1771, p 73.
i

 

 

o

i

»;

the drying which occurred on heating to 600°C.
When the current was applied at 600°C, 30 v was
required to obtain 1 amp. The anode was removed
for examination, found to be in good condition,
and reinstalled. Apporenfly, this was an ‘‘anode
effect,”” and it was probably due to the presence

of oxygen, which is frequently encountered in the

electrolysis of fluoride melts with carbon anodes.
By using a helium sweep.rate of 200 em3/min,
,CF4 was found to be the major constituent (0.33%)
in the effluent gas. Also, CO was present to an
extent of 0.07%. Cn raising the melt temperature,
the anode effect diminished and then suddenly
disappeared at 710°C. The response to temper-
ature was reversible; the anode effect was pro-
nounced at 700°C and absent at 725°C. The
electrolysis was continuved at 725°C.

- At low current densities (1 amp), where easily
reducible impurities were presumed to be carrying
the . current, the predominant constituent recog-
nizable by mass spectrometry of the. effluent
helium was CO, which was produced at an ef-
ficiency of about 0.2 equivalent per faraday.

At slightly higher current densities (1.3 amp),

~ where a plateau in the E-I curve occurred because

the current was controlled by rate of diffusion
of easily reducible ions and where very little

" increase in current resulted from an increase in

applied voltage, -CF‘# began to appear in the
effluent gas.

‘At still higher current densities, the decom-
position potential of ZrFF, was exceeded, and CF
became more prominent, along with small amounts
of higher fluorocarbons. For 6.0 v and 3.6 amp

PERIOD ENDING DECEMBER 10, 1954

the efficiency of production of CO was 14%, and
that of CF, was 16%.

As shown in Table 5.5, impurities (940 ppm of
Fet* and smaller amounts of Ni*t and Cr++)
theoretically equivalent to 0.15 faraday were
initially present. Additional elecirolysis beyond
five times the theoretical amount apparently had
little effect on the Fe analysis. The experiment
demonstrated that NaF-ZrF, mixtures may be
effectively purified by electrolysis under helium
without any preliminary HF treatment. No notice-
able corrosion of the electrolysis apparatus oc-
curred.

When H, was used as the sweep gas while
electrolyzing NaF-ZrF4 (53-47 mole %) at 800°C,
HF could be removed ot a rate proportional to
the current used (0.32 equivalent of HF per
faraday). These results were obtained with ¢
purified melt from which zirconium metal was
being deposited.

In view of a poor efficiency found for the elec-
trolyses of 250-Ib batches containing UF,, an
experiment on 3 kg of NaF-ZrF -UF, (53.5-40-6.5
mole %) was carried out by using H, and elec-
trolysis simultaneously. A filtered sample showed,
prior to electrolysis, 740 ppm Fe, 90 ppm Ni,
40 ppm Cr, and 46 ppm S. The HF level when
the filtered sample was withdrawn was 1 x 10-3
mole/liter,  Simultaneous electrolysis and H
stripping were started. Periodically the elec-
trolysis was stopped and the HF concentration
in the effluent H, was measured to oktain an
index of the degree of reduction achieved. The
HF concenirutions prevailing during electrolysis

 

  

 

" Composition (wt %)**

 

 

 

 

Na F
‘_”'_121 443
n 8 azs

 

*A raw. batch was eiectroiyzed wn‘h no preiimmcry HF or H treatment,
**Theoretical caleulated composuhon (wt %) Zr = 42.5; No = 12.5; F = 45.4,

 

67

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

~were also followed so that the rate of reduction

could be determined. The results are presented .

in Figs. 5.2 and 5.3. After 2.65 faradays had

. passed, the HF level was below 10~% mole/liter,
corresponding to fairly complete purification, and
the trial was terminated. Analysis of the filtered
melt showed 120 ppm Fe, 40 ppm Ni, and 13
ppm Cr. The results given in Fig. 5.2 show that

- the rate of HF evolution with electrolysis is about
three times as great as without electrolysis during
the early stages of the purification. This factor
decreases until there is hardly any difference at

-the end of the run, The decrease is considered

to be due to the gradual accumulation of UF. and
is reflected in Fig. 5.3, which shows the apparent

current efficiency in terms of equivalents of HF
per faraday as a function of the number of faradays

passed. The apparent efficiency varies from 3.5

to 0.5% for the UF ,-containing melt as compared

 with 32% in NaF-ZrF,. The cause of the low

 

 

 

 

 

 

 

 

 

 

 

) eGSR
. ORNL—-LR—-DWG 4527
8
] i ! |
INITIAL HF CONCENTRATION: 10_3moies/li’rer
7 I CURRENT: 10 amp ]
GAS SWEEP RATE:0.33 liter/min
<
g0
o
= _—WITH ELECTROLYSIS
o
E
4
2 o
@ ™
= Y
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— \ \o
t ® Se
w \ | ——WITHOUT "Ne
£ 2 C— ELECTROLYSIS —Seg
L .
® - 9
] D .
. "'-..__‘.\\
0

 

 

 

 

 

 

 

 

0 05 40 1.5 2.0 2.5 30
o FARADAYS PASSED

Flg. 5.2. Rate of HF Removal During Electroly-

~ sis of NaF-ZtF -UF, (53.5-40-6.5 mole %) Under

H, at 800°C.

68

efficiency was apparent from the current-voltage
cutves obtained,.when UF , was present.

After electrolysis had proceeded for a short
time, decomposition potentials and ‘‘knees’’ were
not encountered. Plots of current against voltage

showed a sharp increase from the origin and thus -

indicated that no overvoltages or decomposition

"potentials need be exceeded to obtain large

currents. This increase was interpreted as being
due to the presence of UF, and a small amount
of reduced UF,. The primary electrode reactions,
reduction of UF, at the cathode along with oxi-

dation of UF3 at the anode, caused the over-all

changes in the melt to occur very slowly. No
appreciable amounts of precipitated metals were
found on the cathodes, and it was presumed that
the bulk of impurities such as Fe'! must have
been reduced in the melt as secondary reactions
rather than being deposited on the cathode. Such
behavior might result in metal ‘‘fogs,’’ which
would be most undesirable.

Fluoride Production Facility
F. L. Daley J. P. Blakely

Materials Chemistry Division

A total of 1119.6 kg of processed fluorides was
produced in the 250-1b facility during the past

SRR
ORNL~LR~DWG 4528

 

3.5

 

3.0 \. GAS FLOW RATE: 0.330 liter/min

2.5 \\

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Z
o
P
o
&
xr
L
Q.
o
o —e——f——3.0-amp GURRENT
- o
~ N Y
w 3.0 &
£ \\{\
- o
q 15 S
o // ®
E 10-amp CURRENT e
510 T~
. Y
T S
m '~
Z 05 .
: |
T TOTAL FARADAYS PASSED, 2.646
0

0 0.5 1.0 1.5 20 2.5 3.0 3.5
: FARADAYS PASSED ' '

Fig. 5.3, Apparent Current _Ef.fici.en'cy-' for Re-
moval of HF While Electrolyzing NaF-ZrF ,-UF,
(53.5-40-6.5 mole %) Under H,, at 800°C.

.

X,

-
23

=y,
3
M"‘"“S

xe
 

 

 

"

Y.

.y

-e:-

  
 
 
 

AL

A m RN

1
%
\

in approx1motely 2/4 hr and remained below fhlS/

-+ of iron; how

  
   

Of the c:bove quantities, Prot’r & Whl'mey Alrcmf’r
mole %) and 163.2 kg of NaF-ZrF, (50-50 mole %).
NaF-ZrF ,-UF ,

ZrF -UF

W.-’

“vessel

fhe des:red HF_concen’rrahon' o
had beén reached in 5 hr, or less. Analysis of a

quarter, The various compositions and the amount
of each processed are listed below:

c T ~Amount
omposition
P B Processed (kg)

 

NaF-ZrF -UF ; (50-46-4 mole %) 456.4
NaF+ZrF, (50-50 mole %) 339.4

‘NaF-ZrF -UF ; (50-43.5-6.5 mole %) 323.8
| o 1119.6

T TR BN RS 5

e

Asen,
R R T ey A S R R TR R BT

’%

received 6824 kg of NaF-ZrF .UF, (50-46-4
Battelle Memorial Institute received 23.2 kg of
(50-46-4 mole %), and the Naval
Research Laboratory received 11.3 kg of NaF-
(50-43.5-6.5 mole %). The remaining
ma’rerlol WCIS distributed to various requesters in
the ANP program. s A

S R AT

R B A ey A T

P

e,

While processmg fluorides for general distri-
bution, an attempt has been made to convert to an
electrolytic purification process for removal of
metallic impurities from fluoride melts. The first
trial of electrolysis was made on a non-uranium-
bearing mixture NaF-ZrF, (50-50 mole %) and
proved to be very successful. A 250-lb reactor
was modified so that it could contain a
carbon cathode and a nickel anode, After a 250-Ib
batch had been subjected to the usual preliminary
hydroflucrination, electrolysis of the melt was
carried out. For the electrolysis, a current of
40 amp at 3 v, a melt temperature of 1525°F, and
a helium atmosphere were used, and the melt was

cgnfofed The melt reached fhe acceptable out-gas

concenfrqtlon (1 x 10=4 mole/liter of H2)

HF

~limit for 45 min “before eledrolysns was terminated.

AnalySIs of an unf:ltered sample ylelded HO ppm:'
| 27 ppm Ni, At this pomt fhem-"}
' purlfled ma’rerm]" was spiked wn‘h |ron to givea”"
' 2500 ppm and was agam electro-_
- fimated that about 10 hr of elec- "~ ~

T froiy5|s would be requ:red to remove thls amoum‘“_
“at the end of 6}’ hr the ouf-gos";-’rw
readmg ‘was approx:mdtely _7 x ]0 =5 mole of HF = f

 

 Fe, 40 ppm Cr, a

 

7 Tt'is belleved th
(1 % 10'

 

filtered samp[e from fhls batch yielded 40 ppm
F_e, 60 ppm Cr, and 355 ppm Ni.

   

- uncertainties
uramum-bearmg

  
  

mole/liter)

PERIOD ENDING DECEMBER 10, 1954

Under ideal conditions, the use of hydrogen
stripping to remove metallic impurities requires
12 to 16 hr. With normal production, that is, reactor
vessels containing metallic impurity holdup from
previous batches, hydrogen stripping may take as
long as 50 hr. Therefore it is apparent that elec-
trolytic purification of non-uranium-bearing com-
positions is the better process, and it is to be
used as soon as permanent facilities can be in-
stalled.

The electrolysis of uranium-bearing compositions
is still being investigated. Two batches of NaF-
ZrF ,-UF , (50-43.5-6.5 mole %) have been subjected
to electrolytic purification. The first baich was
given 12 hr of electrolysis, and it was found that
the HF out-gas concentration appeared to level
off at 1.6 x 10~4 mole of HF per liter of effluent
hydrogen. = Consequently, electrolysis was dis-
continued and the batch purification was finished
by hydrogen stripping., Analysis of a filtered
sample from this batch yielded 225 ppm Fe, 20 ppm
Cr, and 40 ppm Ni. A second batch of the same
composition was electrolyzed until the desired HF
out-gas concentration was reached, after 27 hr of
purification, during which period 19 hr of elec-
trolysis was employed. Analysis of a filtered
sample from this batch showed 90 ppm Fe, 55 ppm
Cr, and 40 ppm Ni.

The large increase in electrolysis time required
for the uraniumebearing composition has led to a
more detailed study of the effect of uranium on the
electrolytic purification process. Also, the de-
sirability of obtaining alarge stockpile of processed
fluorides by the first of the year and the present
in the electrolytic purification of
composmons have led to the
temporary cbandonmenf of this method of removing

metallic lmpurltles “from 250-1b batches of fluoride
' composn‘lons confcuning uramum

Alkall Fluorlde Processmg Facnh'ry

J P Blakefy C. R, Craff
erEa ) Trulff |
Materscis Chem:sh'y D'V'Ston

 

   
  
  

 

s under way

Developmenml work ma process -

 

An 1ncreusrng'r6‘re""of eqmpment fallure in
the processmg facility was believed to be due to
the highly corrosive character of alkali-metal
fluorides which have been saturated with HF.

.' 3 69

 
'belew 100 ppm.
~batches showed no detectable differences. The

 

 

s ik T o

RN o

i

‘the reprodumblhfy of this method.
‘aftémpts the U3
"6 wt %. The new series of preparations, listed in
.. :Table 5.6, shows a fair degree of consistency, and

 

ANP QUARTERLY PROGRESS REPORT

Consequently, experiments were run with mixtures

- prepared without the preliminary HF treatment, and

the resulting batches were studied closely for

- such contaminants as oxides, sulfur, iron, chro-
_mlum, and nickel. In all cases the oxide content

was  not defecfable, and the sulfur content well
Comparison with HF-treated

iron, chromium, and nickel concentrations were,
in general, lower than those resulting from HF-

_treated batches. In an attempt to produce the

NaF-KF-LiF eutectic (11,5-42,0-46.5 mole %) plus

14 wt % UF, without the preliminary HF treatment,

it was found that the previous difficulty of in-
consistent results still remained.

A study of the reaction
' U + 3UF4—-—->4UF

in 'rhe presence of cfkah-metal fluorides has shown

~that potassium fluoride has a strong inhibiting
effec'f on the complehon of the reaction. It was
':"found that with the KF concentration of the eutectic

mixture (NaF-KE-LiF, 11.5-42. 0-465 mole %), a
little better than 50% conversion of the UF, to
UF3 could be expected. With this in mind, new
attempts were made to produce consistent batches,
and it soon became evident that the handling
technique was quite important. After several
changes in processing techniques were tried, the
most promising method was chosen as a standard
procedure. Briefly, this method utilizes separate
purification steps for the main constituents in-
volved. The eutectic (NaF-KF-LiF, 11,5-42,0.46.5
mole %) is first heated and stripped with H,. The
melt is cooled and the UF4 is added, after which
the melt is again heated and siripped. The melt
is then cooled, and uranium metal is added. The
final heating and stripping are then carried out,
and the batch is transferred to a storage can.

In this manner, drying and purification of the
alkali metal fluorides are accomplished without

-danger of hydrolyzing or oxidizing the UF ; also,
. thorough mixing of the UF is provided before the
““uranium metal is added. A further advantage is

that the batch is well purified and miscellaneous

- _side reactions of impurities with the uranium metal
.. _are kept to a minimum,

A series of batches was processed to demonstrate
In previous
concentration ranged from 0.5 to

TABLE 5.6. RESULTS OF ANALYSES OF
NaF-KF-LiF EUTECTIC*.UF,
PREPARATIONS

 

 

 

Uranium Centent impurities

(wt %) (ppm)
u3* Total U Fe Cr Ni -
5.39 8.09 155 70 45
4.83 10.9 95 30 85
6.63 12,5 220 30 . 75
4,92 12,3 60 25 35
5,60 10.5 145 5 1510%*
5.02 1.5 140 30 45
5.86 1.2 80 25 25
5.08 1.4 150 30 60
5.70 10.6 430 28 18
5.47 10,9 215 30 135

 

*(11.5-46,5-42,0 mole %).

**Note: to be rechecked.

such material will be accepted for corrosion studies.
Since it was known that the reaction

U + 3UF, —> 4UF,

<

did not go to completion, the presence of uranium
metal in the reactor vessel after completion of a
processed batch was expected and was verified
by examination of the heels left in the vessel.

A rigid program of equipment preparation was also
initiated. All reactor vessels and receiver vessels
are flange-topped to allow access for cleaning, and
the reactor vessels are equipped with nickel liners
to allow easy removal and cleansing. Receiver
vessels are presently equipped with Inconel liners
because attack by HF gas is no longer a problem
since HF was eliminated from the process. Each
unit is completely disassembled after each run,
and all par'rs are cleaned or replaced when neces-
sary. *

Purification of KF ond RbF '
C. M. Blood

Materials Chemistry Division

Several batches of KF and RbF have been proc-
essed to meet requirements for these materials in

purified form. The RbF, when received, ;entdined

. .e-\-;-j
s

¢
T
f

3

. "“"‘55
o
S

L “-sf

i

o
 

PERIOD ENDING DECEMBER 10, 1954

5

as its chief impurity about 0.2 wt % sulfur. Purifi-  Fluorination of NiF,, CrF,, and FeF, has been

s ~ cation of the RbF has proved to be difficult due to  discontinued because the analytical = methods
> attack of the nlckel purification equipment by the  available are not able to establish the fluorine-to-
' v _ sulfur. It has been found possible by rigorous HF metal ratios accurately enough to determine the

% and H, treatments to reduce the sulfur content to  effectiveness of the BrF, ftreatment in removing
" an ccceptcb!e volue, but in the process a large small amounts of oxides that may be present.

amount of nickel is picked up by the melt. Con-
siderable treatment time is required to reduce the
resulting nickel fluoride so that it may be removed L. G. Overholser F. Kertesz
from the product, Materials Chemistry Division

CHEMISTRY OF ALKAL!I HYDROXIDES

Preparahon of Vatious Fluotides Purification of Hydroxides

- B, J. Sturm ) E. E. Ketchen‘ | ' | ~ E.E. Ketchen

T, D Materlals Chemls‘rry Division Materials Chemisffy Division

' : ~ Approximately 5 kg of NaOH was purified either
by filtering a 50 wt % aqueous solution through a
fine sintered-glass filter or by decanting a 50%
solution to remove insoluble Na,CO,.  In either
case the material was then dehydrated at 400°C,
Both methods produced material of desired purity

The preparcflon “and purn‘lcahon of various
structural metal fluorides and several complex
fluorides derived frbm these and alkali fluorides
have been conhnued These materials cmd hydro-
fluormuted UF4 are being utilized at increasing
rates for various investigations of interest to the
ANP program. Chemical analysis, supplemented (~0.1 wt % each of H,0 and Na,CO,). About
" & by x-ray and petrographic examination, has been 1.8 kg of high-purity KOH (~0.1 wt % each K,CO,,
. - utilized to determine purity and ldenhty of 1he H,0, and Na) was prepared by reacting pure po-
" tassium with water, It is not planned to prepare
more hydroxides in the near future because the
material on hand should supply the present limited
requirements. '

N
.y

. o _muterlais. _ o

,. ‘ Uranium tetrafluoride, as received, may contain
small amounts of oxides, moisture, and higher-
valence uramum._ Hydrofluormahon at 600°C is
being carried out as a satisfactory method of re-

* o

Effect of Additives on Hydrogen Pressure

s moving these impurities, During the quarter, 11 kg Over NaOH-Ni System
' . of the purified material was supplied for studies ' | y
requiring high-purity UF ,. Additional batches of F. A. Knox

NIF were prepared by the hydrofluorination at Materials Chemistry Division

600°C of partially dehydrated NICl,6H,0. Ane  The effects of small additions of Ni, Na,0,
hydrous AlF; was prepared by Te thermal de- and mixtures of the two on the hydrogen pressure
_composmon Of (NH4) 3AIF at 500°C. The latter e the NaOH-Ni system have been investigated

v o was synthes:zeg by heafmg Ang 3/2H O w”h - by using a modlflco’non of fhe prevaously described
oo NHy HF2 at 125°C. Two pounds of K N’F was appardfus. 28 It had been found previously that

V_w"w"fprepared by reac'rmg fhe apprc;prlcn‘e qucnhttes large additions of NIO resulted in the formunon of
| r°f KHF2 and N'F 4H O at 800 C exfracfmg the ‘water vapor with no detectable hydrogen. A smaller

 
 
  
 
 
 
  
   
 

     
  

'

was synthemzed by -heafm
9 and CrF 3/H O

»_mckel capsule ogcnn gcve evtdence of ‘water vopor,
“at 800°C the” pressure was only 31 mm, with water

:‘f‘prepar'éd by hem‘mg an | excess‘ Of NH HF2 with ™ When 0.5 mole % was added, a hydrogen pressure
CrF 3/H ,0 at QOOOC Pcm‘ of fhe (NH ),CrF

in excess of soturated water vapor pressure re-
.\..-;.;”suh‘ed and the gas collected from the system was

to Y'eld C"F3: “and cnofher P°”'6"r “aft found to be ‘about” 95% hydrogen. “Therefore, the

   

 

- decomposed to CrFy at 600°C, qufl_r‘tre.qfed W”‘h _-_,;;'_Iorger addmons of NlOGre seen as bemg suffimen’r
" hydrogen at '800°C 1o give CrF a o - e
<  Three b“"rChes. of FeFy We“" ””"””f”ed as a 23ANP Quar. Prog. Rep. Sept. 10, 1953, ORNL.-1609,
means of removing small amounts of impurities. p 84.
o ey . 71

- -\,:-l(
v
e ?

  

_-,addmon of 3 mole % NIO to purlfled NQOH ina |

'droplets found in fhe cool porhons ‘of the appcrotus.'

 

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

fo cause oxidation of the hydrogen from the NaOH-

Ni reaction to water.

Addition of 2 mole % Na,O also resulted in
oxidation of the hydrogen to water. At 800°C the
pressure with this addition was found to be only

33 mm, while the hydrogen pressure of the NaOH-Ni
- system was 126 mm at this temperature.

A 2 mole % addition of an equimolar mixture of

Na,0 and NiO to NaOH gave a pressure of 33 mm

at 800°C. Evidence indicates that a substantial

: frachon of fl-us pressure is due to water vapor,

FUNDAMENTAL CHEM]STRY OF FUSED SALTS
Solublhty of Xenon in Fused Salts
R. F. Newton

Research Director’s Department

Smce the tenfohve va[ues for the solubility of

- xenon in the NaF KF- LlF eutectic and in the

KNO -NaN03 ‘eutectic were reported, 24 experi-

"-'menfs have shown that leakage of xenon through
the frozen seal was possible. To minimize this.

possibility the apparatus was redesigned to have
at least 10 cm of liquid above the frozen section
on the xenon side and to be about 6 cm long. The
new design also permits returning the fused salt
by adjusting the pressure difference and melting
the frozen seal,

While the metallic apparatus for use with the
fluoride eutectic was being rebuilt, the KNO,-
NaNO, eutectic was reinvestigated in a glcss
apparatus that incorporated the new design features
described above. The nitrate mixture can be used
without significant decomposition up to about
450°C; in 6 hr, ot temperature,about 0,01% of the
nitrate was decomposed at 410°C and about 0.3%
was decomposed at 500°C,

With the new apparatus, values of § x 10~% and
9.3 x 10=% mole of xenon per milliliter of melt

at 260°C and 8.9 x 108 and 9.6 x 10-8

450°C were obtained. These values are essentially

in agreement with the previous ones, namely

'-‘8,5'><. IQ"‘S at 280°C and 10 x 10~8 at 360°C.
“The change in solubility with temperature appears
1o be very small,

. X-Ray Diffraction Studies in Salt Systems
P. A. Agron M. A, Bredig
Chemistry Division

MF.XF, Binary Systems. The renewed interest
in the binary systems of alkali fluorides with

uranium trifluorides was emphasized this past
quarter. In studying these systems, difficulties
arise from partial oxidation of U(HI) to U(IV)
and/or through disproportionation of the uranium
in solution in the molten alkali fluorides. It thus
appeared to be worth while first to obtain un-
equivocal x-ray data on the corresponding binary
fluoride complexes of some of the ‘‘4f rare earth’’
metals,2? especially of lanthanum, which may be
considered as a good “*stand-in’’ for U(H1).

Zachariasen reported2é several trifluorides of the
“‘Sfrare earth’’ group as having the same ‘“tysonite’’
structure as those of the ‘‘4f'' group, previously
determined by Oftedal.?” For the trifluorides,
the average La-F and U-F distances are given as
2.50 and 2.56 B\, respectively. The lattice di-
mensions of these hexagonal cells differ by only
0.05%. Thus it is not surprising to discover in
samples prepared here?® that the structure of the
double fluorides?® of NalaF, and KLQF4 had
their counterpart in analogous NGUF4 and KUF,
structures,

According to Dergunov, in MF-LaF, systems
all alkali fluorides, with the possible exception
of CsF, form 1:1 compounds, and only CsF forms
a 3:1, CS3LGF6, congruently melting compound.
These data were obtained by visual observation
of crystallization from melts, Compounds below
the eutectic that are stable and polymorphous
transitions would not have been discerned.

In the present work, fused mixtures of the binary
systems MF-LaF, (except 3NaF-LaF, and LiF
mixtures) of the molar compositions 3:1 and 1:1
were made available from thermal-halt measure-
ments.2?  X-ray diffractometer patterns were ob-
tained on these melts, The phases found in the
NaF and KF systems are indicated in Table 5.7
and are the phases expected for these binary com-
positions., Excellent agreement is shown for the

7

 

24%. F. Newton and D. G. Hill, ANP Quar. Prog. Rep,
Sept. 10, 1954, ORNL-1771, p 70.

23y, H. Zachariasen, Acta Cryst.
Am. Chem. Soc. 70, 2147 (1948).

1, 265 (1948); J.

26y, H. Zachariasen, Fluorides of Uranium and Tho-

rium with the LaF3 Type of Structure, MDDC-1153 (date

of manuscript, June 1946; date of declassification,

July 18, 1947). |
27\, Oftedal, Z. psysik. Chem. 5B, 272-291 (1929).
28\/ S. Coleman, W. C. Whitley, and C. J. Barton,

Mcferlqls Chemistry Division,

L. M. Bratcher and C, J. Barton, Materials Chemlsfry
Division.

"
 

L] y

™

L3

e

o

:The relcmve thermcl

& 32y, S ‘Coleman and W. €. Whitley, ANP Quar. Prog.

PERIOD ENDING DECEMBER 10, 1954

TABLE 5.7. PHASES FOUND IN MF-LoF; BINARY MIXTURES

 

Hexagonal L m‘c\;ice

 

 

.Scmple Molc‘r. Structure Type Dimensions {A) o

Number Composition Observed Repor‘l‘ed3]
L MB-108 INaF - 1L aF, BNl aF a = 617 a=6167 %

' o : c = 3.82 c = 3.819
-109 3KF-1LaF, By-KLaF 4 (and KF) a = 6.52 a = 6.524 %

' c = 3,79 c = 3.791
110 3KF~ILaF, B1-KLaF 4 (and KF) 4 = 6.52 a= 6524 %

‘ B ' S e = 3.79 c = 3.791
-107 IKF-1LaFg B1+KLaF 4 (and unknown a = 6,52 a = 6.524 %

' ' phase) c = 3.79 c = 3.7%91

131 & .‘.]kF-—TLOF:; One phase, p.redomina'nfly

{same as unknown cbove)

 

*Zcxchc:ficxsen;3‘l the quantity given is the number of molecules per unit cell.

lattice rco'hs'tdri"rs";”t(-'l"abie 5.7). The expected
complex phase is found in the NaF- LaF, system,

The x-ray analysis of the KF-LaF, fuswn (Table

5.7, sample LMB-107) indicates the presence of
B1 -KLaF, and an unknown second phase. A
second fusmn at this composition (sample LMB-

131), with an lmposed slower rate of cooling, gives

predominantly a single phase which corresponds
to the unknown structure found in the previous
preparation. For the 3KF- -LaF , composition, only
the 5,-KLaF, complex has been observed. Petro-_
graphic exummahons?’ indicate that the new phase
has optical’ prOperhes snm:lar to fhe B -phase.

rmal stabi "tles of the»'e,_’rWO phases

“ted fo be E:ubic,3'_ remom

   

 

Rep. Sept. 10, 1952, ORNL-1375, p 79 V. S. Coleman,
W, C. W |tiey, and C. J. Barton, ANP Quar Prog Rep.
Dec. 10, 1952, ORNL-1439, p ]09

 

'rly.ie]ded the data shown in Table 5.8.

" “tration of Na; possubly, Na:K

 

 
   

= fo"300°K At 298.15°K; c

The in-
terpretation was aided greatly by comparing these
diffraction patterns with the corresponding LaF,
salts, The Greek letter assignment is adopted to
correspond with the lanthanum compounds as given
by Zachariasen.

" The ternary salt mixture 1 KF-1 NaF-2 UF,
gives the phase 3, “(K,Na)UF,. Here the ordering
requires a doubling of the ¢ axis observed in the
B,-phase of the binary salt. The extent of the
. By~KLaF  present in this melt indicates that the
former phase is stabilized by the greater concen-
= 2:1 in the lomce.

Physncal Chem'sfl'y
E . R. :Van Arfsdoie” o

 
   
  

  

HO)/T were fobulqted at‘ 10- deg mterva!s up
= 22.83 cal/mole- °K,

 

 

Defcl[s of this work will be published in separate
reports and articles by the Chemistry Division.

73

 

 
R g L

i i‘ b s i Lo

 

ANP QUARTERLY PROGRESS REPORT

TABLE 5.8. PHASES FOUND IN MF-UF; BINARY MIXTURES

 

 

Molar Composition? Ph Structure T Lattice zb
olar Composition ases tructure lype Dimensions (A)

72NaF ]UF3 HexqgonalAphase and fiz-NaUF4 = 6.17 3/2
3NaF—2UF3 also NaF = 3.78

]KF ]UF3 Hexagonal phase and B]-KUF4 a = 6,51 3/2
: : cubic phase® c = 3,76
Fluorite type ag = 5.9

IKF-INaF-2UF, Two hexagonal phases By “(K,Na)UF,, a=621 3
c =771

B-KUF, a = 6.51 ¥

c = 3.76

 

a.Sampfes prepc':lred by V. S. Coleman and W. C. Whitley and examined petrographically by T. N. McVay.

The number of molecules per unit cell.

Thls phose may be the cubic phase of KUF4 or the compound UOF with the fluorite structure.

and S° = 29,90 e.u., of which 0.53 e.u. was ob-
tained by extrapolation below 20°K.

The heat capacity of a sample of pure MoS,
was measured from about 15°K to room temperature,
The heat capacity and the derived thermodynamic
functions mentioned above were tabulated at 10-deg
intervals up to 300°K, At 298.15°K, C, = 15.62
cal/mole *°K, and §° = 15.34 e.u. The heat ca-
pacity of MoS, follows a T2 dependence between
15 and about 70°K of the sort previously reported>4
for MoO,. This behavior of MoS, is consistent
with its lcyer structure, which is so pronounced
that MoS, actually is a good high-temperature
fubricant (Liqui-Moly).

Density and electrical conductance measurements
were made of molten mixtures of KCl and Kl across
the entire composition face. The molar volume, as
calculated from the density, is nearly additive

for this system, although it shows deviations close
to the Kl side. In contrast, the equivalent con-
"'_”ductance shows pronounced negative deviations
- from addlflwty for all mixtures, with maximum
,—_’:devmhon at high KCl concentration, These data
“‘again show that it is inadvisable to interpret
‘maximums and minimums in electrical conductance
~as the consequence of formation of compounds in

 

34g. R. Van Artsdalen, ANP Quar. Prog. Rep. Sept.

. 10, 1954, ORNL-1771, p 72.

74

the melts. The specific conductivity of molten
Kl is expressed by the equation

k = =1.7100 + 6.408 x 10~%
- 2.965 x 10~%:2 mho/cm

between 725 and 925°C, and the density is gwen
by the equation
o= 3.0985 - 0.9557 x 10~3 g/cm®
between 680 and 910°C. The temperature, ¢, is
in °C.
The density and electrical conductance were

determined for pure molten NaBr and are given
below for the range 750 to 960°C:

k = —0.4392 + 5,632 x 10=3; ‘
- 1.572 x 10~%:2 mho/cm

and

p = 2.9518 — 0.8169 x 10~3; g/cm3.

The self-diffusion coefficient of sodium ion in

molten NaNO, has been determined by a radio-
chemical tracer technique. The heat of activation
for self-diffusion is approximately 6 kcal/mole and,
therefore, higher than the heat of activation for
electrical conductance. Similar work with thallous
chloride®5 has shown that the heat of activation

35E. Berne and A. ‘K.lemm, Z. Naturforsch, 8a, 400
(1953).

 

L2

.
LR
™0
-4

o
wn, 3
L

A

)
o

 

Xt

 

for self-diffusion of the thallous ion is greater than
that for electrical conduction. These results are
in accord with the concept that the application
of an electrical field lowers the potential barrier
restricting migration of charged ions in the ap-
propriate direction of the applied field. The self-
diffusion coefficient of a given ion can be used in
conjunction with the Nernst-Einstein equation®®
to calculate the contribution of that particular ion
to the total equivalent conductance of the sait.
When this is done, it is observed for sodium nitrate
and thallous chloride that, in each case, the cation
accounts for about 95% of the equivalent con-
ductance of the respective molten salt at temper-
atures about 25°C above the melting point, This
indicates that the fransport numbers of these
cations relative to thelr anions in these two salts
are opproximately 0 95.

Vlscosfly Measuremenfs B

F. A, Knox F Ker'resz
N Y. Smith

Mcn‘erlals Chem:s’rry Dlws:on

The automatic coplllary wscometer, which was

used previously®? for determining the viscosity of

 

36/\ = Dz F2/RT where A is the equivalent cone

ducfance D the self-diffusion coefficient, z the charge
of ions, 'F the Faraday constant, R the gas constant,
and T the absolute temperature.

-

PERIOD ENDING DECEMBER 10, 1954

fluoride melts, has been modified and placed in
operation for viscosity determinations c¢f alkali
nitrate melts. Effort has been concentrated on
determining viscosity changes immediately above
the melting points as an adjunct to the work being
done by Van Artsdalen to establish the signifi-
cance of ion size on physical properties (see
section above). In view of the low melting points
of the salt mixtures being studied, a glass instru-
ment is used, but it is planned to use an all-metal
apparatus when the investigation is extended to
chloride and fluoride melts.

The salt mixtures studied included pure sodium
and potassium nitrates as well as mixtures of the
two in 12,5 wt % steps. The values obtained for
the pure salts were about 0.3 centipoise above the
data reported by Dantuma,3® who determined the
viscosity by the logarithmic decrement method.
The viscosities of the mixtures nearly overlap at
higher temperatures, being located between the
viscosity curves of the pure components. A plot
of recnprocal temperatures vs the fog of viscosity
gave an apparent deviation from linearity near the
melting point of most of the mixtures, while the
pure salts showed a direct linear correlation.

 

376, A. Knox, N. V. Smith, and F. Kertesz, ANP
Quar. Prog. Rep. Sept. 10, 1952, ORNL~1375, p 145,

38F2. S. Dantuma, Z. anorg. u. allgem. Chem. 175,
33-4 (1928).

 

Yo umn ~
N ;‘ wi f Q 75

 

 

 
b, il ce 1

B bt

 

N T

gk e s ey . . N

s b i o

RN -

 

ANP QUARTERLY PROGRESS REPORT

6. CORROSION RESEARCH

W. D. Manly

G. M. Adamson

Metallurgy Division

W. R. Grimes

F. Kertesz

Materials Chemistry Division

‘Studies of the corrosion of Inconel, type 316

-‘sfamless steel, Hastelloy B, and several special
qlloys were continued through the use of thermal-
~convection apparatus. Additional information on
‘the compatibility of the sodium-beryllium-Inconel
‘system was obtained from static, whirligig, and

'fhérmql-c;oh‘\)ecfi_on loop tests. In an effort to find
a brazing alloy that has good resistance to sodium
and to NaF-ZrF - UF4, tests of several brazes on

type 304 stmnless steel T-joints were made. Simi-

far joints of A-nickel brazed with various alloys
were also tested in NaF- -ZrF ,-UF, at 1500°F and
in sodium hydroxide at 1100 and 1500°F. Addi-
tional screening tests were made of carbides of
boron, titanium, zirconium, and chromium in static
s'o‘dium, lithium, and NQF-ZrF4-UF4 at 1500°F for
100-hr periods. An 80% Mg—20% Li alloy being
considered as a crew-compartment shielding material
was corrosion tested in water at various temper-
atures; the weight losses were found to be negli-
gible.  The resistance of Inconel to attack by
molten rubidium was studied in a dynamic loop
apparatus. Fundamental studies are under way of
the mass transfer in liquid lead and of fused hy-
droxides as acid-base analog systems, and further
mformahon has been obtained of the flammability
of sodium-bismuth alloys. In further chemical
studies of corrosion, information was obtained on
the effect of temperature and of chromium additions
on the corrosion of Inconel by fluoride melts with

NiF, additions.

THERMAL-CONVECTION
LOGP CORROSION STUDIES

G. M. Adamson A. Taboada
Metallurgy Division

Inconel Loop Containing UF, in
- Alkali-Metal-Base Mixtures

- Additional tests were made with alkali-metal-base
fluoride mixtures containing UF; and UF , in Inconel

o __"'thermal-convecnon loops, but the results obfomed
- were not so favorable as those reported previously.!

Cons:derable difficulty was encountered in making

76

the mixtures, in controlling the total uranium con-
tent, and in determining the ratio of UF; to UF,.

Two supposedly duplicate Inconel loops were
operated for 500 hr with the UF;-UF, mixture in
NaF-KF-LiF (11.5-42-46.5 mole %) at a hot-leg
temperature of 1500°F (AT of ~200°F); however,
the rough surface of one loop showed maximum

penetration to a depth of 2 mils, whereas the other

loop was attacked to a depth of 13 mils. In three
other supposedly duplicate loops operated for 1000
hr, attack to depths of 8, 13, and 13.5 mils de-
veloped. Metallic-appearing layers were found in
the cold legs of all these loops.

Several loops filled with NaF-KF-LiF with and
without UF, were operated to obtain comparative
data for use in evaluating the results of these
tests, The data obtained are presented in Table
6.1.

Type 316 Stainless Steel Loops Containing
UF, in Alkali-Metal-Base Mixtures

Several batches of the eutectic NaF-KF-LiF (11.5-
42-46.5 mole %) containing mixtures of UF, and UF
were circulated in type 316 stainless steel loops.
However, the results obtained cannot be used for
comparative studies because the UF ,-to-UF ratios
in the fluoride mixtures were unknown and the mix-
tures were of doubtful purity. One loop plugged after
886 hr of operation at a hot-leg temperature of
1500°F; examination showed heavy, intergranular,
subsurface voids to a depth of 11.5 mils in the hot
leg and some intergranular attack in the cold leg.
Another similar loop operated its scheduled 1000
hr, but a large deposit of dendritic crystals was
found in the cold leg.

A similar loop in which all the uranium (12 wt %)
was present as UF, was operated to obtain com-
parative data. This loop plugged after 570 hr of
operation, and metallic crystals were found in the
cold leg. The hot-leg surface was very rough, and
there was penetration to a depth of 5 mils. The

 

'G. M. Adamson, ANP Quar. Prog, Rep. Sept. 10, 1954,
ORNL-1771, p 96.
 

 

"

PERIOD ENDING DECEMBER 10, 1954

TABLE 6.1. RESULTS OF INCONEL THERMAL-CONVECTION LOOP TESTS
OF NaF-KF-LiF (1. 5-42-46.5 mole %) WITH AND WITHOUT UF,
CIRCULATED AT A HOT-LEG TEMPERATURE OF 1500°F

 
  
   
  

' Hoste”oy B thermal C >0
‘& o _;F -UF, (50- 46-4- mole %),

 

Period of Uranium Maximum

Metallographic Notes

 

 

Test Content Attack :

(hr) (wt %) {mils) Hot-Leg Appearance Cold-Leg Appearance
500 11.9 1§ Heovy,"infergronulor, subsurface voids Metallic deposit

500 12* 13,5 Hequ, general, intergrufiulur, subsurface voids Metallic deposit

1000 12* 20 Heavy, general, intergranular, subsurface voids  Metallic deposit

500 o 7 Mederate, general, intergranular, subsurface No deposit

voids
Heavy, general, intergranulor, subsurface voids  Thin deposit in one section

500 o 75

 

‘*Estimated; analyses not yet available.

same fluoride mixfure without uranium was circu-
lated for 1000 hr without signs of plugging. The
hot leg of this laop showed heavy attack to a depth
of 10 mils, and some grains were completely removed.

Effect of Induced Potential

Welding rods for use as electrodes were attached
to the upper hot legs and lower cold legs of Inconel
thermal-convection loops that were operated to
defermlne the effect of a small induced emf on
corrosion, Bah‘ery chargers were used to apply
direct currents of 5. 4 amp at 0.45 v. The loops
circulated NaF- ZrF UF (50-46-4 mole %) at a
hot-leg temperature of ]500°F

The direction of current flow was expected to
have an effect on 'rhe attack, but no such variation

could be observed in 'rhe four Ioops that have been
vever, the depths of ch‘ack were less
e _fhan the depfh of ctfcck found m a Ioop filled from“"

the same bcn‘ch of fluorlde mleure cmd oper'd'red/
'w:fhouf ctn mduced potenhol ’

  

exam med However ,

 

loy B Thermol-Convechon Loops

 

fluoride mixtures, but there is some evidence that

 
  
   

-convection loops have been

these foops are p‘i'eseh'red in Table 6 2. "“The perlod'fi—,;_, |
, " lated NaF- ZrF UF (50 464 mole 7) at a hot-leg

- of operation of the |oop has not had a 'no’rlceable
‘effect on corrésion in the loops that circulated the

the small attack that occurs takes place in a fairly
short time and is a function of the condition of the
original surface.

The loops operated with sodium were known to be
covered with an oxide deposit and were not cleaned
before filling., Additional loops that were cleaned
with the use of dry hydrogen are now being operated.

Special Alloy Thermal-Convection Loops

Several loops were fabricated from tubing of
modified Inconel specially drawn at Superior Tube
Co. from billets that were vacuum cast at ORNL,
The loops were standard size, but the tubing was
I/2 in. in diameter with a 0.065-in, wall rather than

- the standard 3/ in.-IPS schedule-10 pipe; they were
- filled with NCIF ZrF ,-UF, (50-46-4 mole %) and
operated with a hof-leg tempercfure of 1500°F,
. Some of these loops ruptured when attempts were

... made to restart them after the two power failures,

.. The data obtained from these loops are given in
' nTc_xb!e 6. 3 I’r mc:y be noted 'rhat a considerable
fir'educhon in depth of attack was found in both
loops in which the chromium content of the alloy
had been reduced to about 5%,

;llntergrcnulor, subsurface-vmd type of attack to a
~ depth of 6 mils in ]000 hr. ThlS loop also circu-

“temperature of 1500°F. The attack was less than
that usually found with Inconel.

77

“One’ loop of Hustelloy ’C d'eveloped a heavy,"

 

 

 

 
 

 

 

 

- ANP QUARTERLY PROGRESS REPORT

TABLE 6.2. RESULTS OF OPERATION OF HASTELLOY B THERMAL CONVECTION
LOOPS WITH VARIOUS FLUIDS '

 

“ . Hot-Leg Temfiérafure

 

 

g ‘ L . Period O'F
Circuloted Fluid Temperature Differential Test Metallographic Notes
| | (°F) (°F) (hr) '
NdFr_-.Z'r'F“-'UF4 | 1500 325 1000 rRough surface; mfergrdm..ll'd} penef?c’ridné; subsurface
7 . voids to 3 mils
1500 325 2000 Shghfly rough surface; voids to a depth of 2 mils
1650 400 1000 Rough surface; most voids to depths of 2 mils or
less, but occasionally to 3 mils
1650 400 1000 Maximum attack to o depth of 2 mils; slight deposit
. in hot leg
NaF-KF-LiF-UF4 1500 325 1000 Subsurface voids and intergranular penetration to a
' : . depth of 2 mils; thin deposit in hot leg; intergranu-
lar attack in cold leg to a depfh of ].5 mils
Sedium 1500 400 1000 Fine metallic crysfuls adhered to cold Ieg when loop
drained while at 1500°F
1500 400 1010 Plugged by mass of fine dendritic crystals in hot leg
that probably formed in cold leg
TABLE 6.3, RESULTS OF OPERATION OF SPECIAL convection loops with beryllium inserts in the top

ALLOY THERMAL-CONVECTION LOOPS WITH
NaF-Z:F 4~UF 4 (50-46-4 mole %) WITH A
HOT-LEG TEMPERATURE OF 1500°F

 

 

 

Alloy Composition {wt %) Operating Maximum
Time Attack

Ni Fe Cr . Mo (hr) (mils)

76 7 17 1000 13

76 14 10 823* 8

76 14 10 1000* 12

76 19 5 647 3

76 19 5 460* 3

83 7 10 1000 13

83 7 10 1000 15

74 10 6 10 1000 3

74 10 6 500 3.5

10

 

*Leak developed in loop.

SODIUM BERYLLIUM-INCONEL
" COMPATIBILITY

E. E. Hoffman
C. F. Leitten
‘A, Taboada
Meto”urgy Division
As one phase of the sodium- berylllum-lnconel
‘compatibility studies, a series of Inconel thermal-

of the hot legs were operated. The inserts were
hollow cylinders about 6 in. long with inside
diameters the same as those of the Inconel tubing
of the loop. The outside of the beryllium was pro-
tected by an Inconel sleeve and was separated
from it by a 0.050-in. annulus filled with slow-
moving sodium. The loops were operated for 500
he with high-purity sodium at hot-leg temperatures
of 900, 1100, and 1300°F. In none of these loops
was any attack found on the inside surface of the
beryllium, and the outside surfaces of the inserts
in the loops operated at 900 and 1100°F were
similarly unattacked. However, in the loop oper-
ated at 1300°F, the outside surface of the beryllium
insert showed widely scattered voids to a depth of
3 mils, All the beryllium specimens were darkened,
but no deposits were found by metallographic
examination. _ ,

A summary of compatibility tests conducted by
two different methods is presented in Table 6.4.
In none of the tests was a layer deposited in the
cooler portions of the test system that could be
detected metallographically. in all the tests, how-
ever, beryllium was found to be present on various
sections of the Inconel tubes. Figure 6.1 shows
the distribution of beryllium around an Inconel
whirligig loop following a 270-hr test at a hot-zone
temperature of 1200°F and a cold-zone temperature
of 1060°F, Since the velocity of the sodium bath

-,

&

O
.4

 
 

¥

2

 

d
e L

in this apparatus was 10 fps, the bath temperature
was practically isothermal, The temperatures given
are the outside tube wall temperatures. The be-
ryllium distribution shown appears to be temperature

~ dependent; however, in other dynamic tests, this

was not so obvious. The outside and inside sur-
faces of a beryllium insert from the hot leg of an
Inconel thermal-convection loop are shown in Fig.
6.2. As in nearly all the other compatibility tests,

PERIOD ENDING DECEMBER 10, 1954

the heaviest attack was on the outside surface of
the insert. The attack was due to dissimilar metal
mass transfer between the beryllium and the Inconel
sleeve which surrounded it. The result of direct
contact between Inconel and beryllium in a system

‘containing sodium at elevated temperatures may be

seen in Fig, 6.3, The Be-Ni intermetaliic layer
which formed was approximately 20 mils thick and
was extremely hard and brittle, Tests are presently

TABLE 6.4, RESULTS OF B.E'RYLLIUM-SODIUM-INCONEL COMPATIBILITY TESTS

 

 

Coneentration of

Test o
Test T?s Temperature (7F) Be on Surface Metall hic N
me
Type (hr) Hot Zone Cold Zone of Inconel Tube etallographic Notes

. (pg/cmz)

 

- Whirligig 3.00 1100 !.sothermal- 5.3 to 7.9
300 1200  Isothermal  0.05
300 1300 lsothermal 0,07

300 1400 Isothermal  39.6 to 127

270 1200 1060

- Convec on

1300 1130

          

16,5 to 39.6*

  
   
 
    

Two dark deposits on Inconel which analyzed high in
beryllium; layer on Inconel sleeve; beryllium insert
‘showed no attack on inside and 1 mil of attack on

outside surface

No deposits on Inconel tube; 0.4-mil layer on Inconel
sleeve, beryfl:um insert had no attack on inside
surfcce and 1 m:l of attack (volc{s) on outside

‘surface

Inconel sleeve which enclosed beryllium insert had
1-mil layer on surface; beryllium insert had no
attack on inside surface and subsurface voids to

a depth of 5 mils on outside surface

Deposit of 2 to 3 mils on Inconel sleeve; beryllium
insert attacked to a depth of 1 mil on inner surface

and to a depth of 3 mils on outer surface

Deposfi' of 0. 2 mll on Inconel s!eeve, bery|||um ine
serf o'rfacked fo a depth of 2 ml!s on |nS|de surfcce

   

 

*See Fige 61,

 

TS T e
el R 79

 

 

 

 

e T OGO T Ty g e

-

 

 

 
 

 

 

 

- ANP QUARTERLY PROGRESS REPORT

UNCLASSIFIED
ORNL-LR~-DWG 4650

 

BERYLLIUM INSERT:
LENGTH — 3.75in.
ID—0.430in.

Flg. 6 '|. Dlstnbuhon of Berylllum Around an

~ Inconel Whlrhg|g Loop Containing a Beryllium
“Insert and’ Sodlum. Tesf period, 270 hr.

UNCL ASSIFIED
3497

 

 

 

 

 

 

 

 

 

 

 

 

to Sodlum for 1000 hr at 1300°F. (a) Surface of
i 'berylllum exposed to sodium in annular space.
- (b) Surface of beryllium exposed to flowing sodium.

~ Unetched. 500X. Reduced 37.5%.

under way to determine the minimum spacing that
can be tolerated between beryllium and Inconel.

STATIC CORROSION STUDIES

E. E. Hoffman .
W. H. Cook C. F. Len‘fen
Metallurgy Division

Brazing Alloys on Stainless Steel

Additional tests have been made in an effort to
find a brazing alloy that has good resistance to
both sodium and the fuel mixture NaF-ZrF -UF,
(53.5-40-6.5 mole %). The results reported in Table
6.5 were obtained from tests of type 304 stainless
steel T-joints brazed with the alloys listed by the
Wall Colmonoy Corporation. Tests are ‘also being
conducted on Inconel T-joints brazed with each of
the alloys listed in Table 6.5. All the brazing
alloys listed had good resistance to the fluoride
mixture, but only alloy C-29 had good resistance
to both the fluoride mixture and to sodium.

Brazing Alloys on Nickel

A series of T-joints of A-nickel were brazed with
the various alloys listed in Tables 6.6 and 6.7 in
a dry hydrogen atmosphere and were then exposed
for 100 hr to static NaF-ZrF -UF, (53.5-40-6.5
mole %) at 1500°F or to static sodium hydroxide at
1100 and 1500°F,. The specimens were examined me-
tallographically in the as-received condition in
order to evaluate the inclusions and the porosity.
The brazed specimens were thoroughly cleaned
before and after testing so that valid weight change
data could be obtained. In order to prevent the
rounding of the fillet edge upon polishing, each
specimen was nickel plated after testing. The
data obtained from metallographic examination of
the specimens exposed to the fluoride mixture are
presented in Table 6.6, and the data on the speci-
mens exposed to sodium hydroxide are in Table

6.7.

Most of the alloys tested appeared to have good
corrosion resistance to the fluoride mixture, except
the braze alloy with the composition 69% Ni-20%
Cr-11% Si. Several of the brazed T-joints, es-
pecially those which included copper, gold, or
silicon as an alloying element in the braze material,
showed numerous voids in the interface between
the base material and the braze fillet., These voids
are not considered to be caused by the attack of

the fluoride mixture but rather by diffusion of a

e TR “A"\.:‘
e i

LR f

 

 
 

 

PERIOD ENDING DECEMBER 10, 1954

 

A

 

 

 

 

 

 

 

 

 

 

 

 

 

x
x
’ o o © o = o o O 1e lo -
: e B kB B & g kB kB B B
.-y | @ INCH |® ® < & > 5 & 5|2 _
= Fig. 6.3. Diffusion Layer of Be Nl Which Formed on Inconel in Dlrect Contacf wn‘h Berylllum Insert
'. in a Thermal-Convection Loop After Exposure for 1000 hr to Sodium at 1300°F. Hardness impressions
:F

md:ca?egrelahv‘e‘hordnesses. e, Nig, 1300 DPH; Inconel, 180 DPH. Etched with oxalic acid. 500X.

constituent from the braze alloy into the base Table 6.7 shows the results obtained from the
material. A T-joint brazed with 60% Pd—37% Ni~  metallographic examination of the brazed A-nickel
3% Si and then exposed to the fluoride mixture for T-joints exposed to sodium hydroxide. Of all the
100 hr is s_h_own in Fig, 6.4, Many voids can be  braze alloys tesfed only the 82% Au—lB% Ni-alloy
“seen along ‘the interface between the base moferlal had good resusrqnce 'ro at’rack ot both 1100 and
- and the braze f[llet but the surface of fhe specnmen ) l500°F The specrmen ’rested in sodium hydrox:de
" dppears to be*freeof ofl'ock - _ L for l00 hr ot 1500°F is shown in Flg. 6.6, '
,.’:ln’order to Ver_lf;/ the poss:bilrfy of cllffuswn"“ TR 'rhese rests, as in the fused fluoride tests,
‘base “voids were found along the interface between the
base materml and the broze f:llef when the broze
c:lloys ‘that contamed copper, gold or silicon as
_an olloymg elemem‘ were ‘used, However, fhe same
specrnfens 'res’red at ll00°F reveoled no such

vo:ds. 7 Tl’ns si 'uahon s Pmbc'bly caused by the

 

  
 

   
 
 

 

‘manner as were the tos'led T-|omts"\un’c~lv placed in
f_‘_on evacuofeclw copsule and whec:tecl “at 1500°F for

  
 

 

A

     
  

 

  
  

   
  

 

« 7 o and 'rl':ef_l; | _
' " “Voids appears to be independent of environment bemg “Fested in sodlum hydroxlde For 100 hr at
and therefore not a result of attack by the fluoride 1100 and 1500°F, respectively. As may be seen

 

o mixture. | in Fig. 6.7, no attack occurred along the braze
81
IR

   

AR RARPTTRUER CORT TN AP TEIRILT Cre T (U T

 
ANP QUARTERLY PROGRESS REPORT

{ TABLE 6.5, RESULTS OF STATIC TESTS OF BRAZING ALLOYS ON TYPE 304 STAINLESS STEEL T-JOINTS
IN NaF-Z¢F -UF , (53.5-40-6.5 mole %) AND IN SODIUM AT 1500°F FOR 100 hr

 

 

Brazing Alloy Weight o Weight

 

Composition Bath Change*  Change Metallographic Notes
(wt %) {9} (%)
Alloy A-16 NaF-ZrF -UF, —0.0003 =—0.042 Braze fillet unattacked
23 P, 77 Ni _ Sodium -0.0009 ~0.135 Subsurface voids in braze fillet to a depth
Lo i . of 5 mils; attack confined to Ni3P phase
:i Alloy C-27 NGF-ZI‘F4-UF4 o 0 No attack
A 9.6 P, 2,75 Cr, 88.6 Ni Sodium —~0.0006 —0.086 Subsurface voids to a depth of 5 mils;
S _ . o Ni;P phase attacked
4 Alloy C-29 NcF-ZrF4-UF4 +0.0011 +0.180 No attack '

 

; 10.2 P, 13 Cr, 76.8 Ni Sedium +0.0004 +0.054 Less than 0.5 mil of small subsurface
: ' - ' voids '
Alloy B-11 NaF-ZrF4-UF4 ~0.0006 -0.082 Braze fillet attacked to a depth of 1 mil

in several areas

10.8 P, 9.2 Si, 80 Ni
: o Sodium 0 0 Subsurface voids in fillet to a depth of

4 mils

Small subsurface voids in braze joint to a
depth of 0.5 mil

Subsurface voids to a depth of 7 mils;

Alloy J-10
9P, 15 Fe, 4.5 Cr, 71.5 Ni

Nc«F‘-ZrFA-UF4 +0.0007 +0.153

Sodium —0.0004 -0.067

Ni3P removed from fillet zones

Maximum attack was 0.5 mil in the form of

Alloy H-10
10 P, 4.3 Mo, 86.7 Ni

Na F-ZrF4-UF-'4 +0.0036 +0.730

small subsurface voids

 

Sodium —0.0006 -0.108 Subsurface voids in braze fillet to a depth
_ of 19 mils
Alloy 1-10 Nc:F-Zer-UF:4 +0.0008 +0.112 No attack
11.6 P, 6.25 Mn, 82.15 Ni  Sodium —0.0018 —0.346 Subsurface voids in braze fillet to a depth
| ' of 11 mils
f Alloy D-11 NaF-ZrF,-UF, +0.0015 +0.304 No attack

9.9 P, 11.3 Fe, 78.8 Ni Sodium +0.0006  +0.108 Subsurface voids in braze fillet to a depth
| of 25 mils

 

*Weight change data include brazing alloy and base material of joint.

 

fillet of the specimen tested at 1100°F and very
few diffusion voids were present. However, as
may be seen in Fig. 6.8, the specimen tested at
1500°F showed a é-mil surface ottack and also
several small interfacial voids.

Additional static corrosion tests are being con-

ducted on many of these same brazing alloys in
sodium in order to evaluate their suitability for

"_‘usv'e as aback braze for the sodium-to-fused fluoride
heat exchangers.

gual

Screening Tests of Carbides

Specimens of carbides of boron, titanium, zir-
conium, and chromium were tested in static sodium,
lithium, and NaF-ZrF ,-UF, (53.5-40-6.5 mole %)
at 1500°F for 100.hr periods. The measured densi-
ties .and apparent porosities of the specimens and
the results of the corrosion tests are given in
Table 6.8. The data indicate that TiC and Cr,C,
had the best resistance to corrosion in each of the
mediums; Figs. 6.9 and 6.10 show the typical

. )
Sama? -

E }
2

X
a0
 

A ARG L M SRS

 

 

 

e ARG, S

PERIOD ENDING DECEMBER 10, 1954

 

 

 

 

 

 
  
 
   
   
  
    
 
    

   

 

  
  
      

 

, TABLE 6.6. RESULTS OF STATIC TESTS OF BRAZED A-NICKEL T-JOINTS IN
. NaF-Z¢F,-UF, (53.5-40-6.5 mole %) AT 1500°F FOR 100 hr
s, Brdiing Aiioy 7 _ ‘Weight | Wévighfr
- Composition Change Change Metallographic Notes
(wr %) (@) "
2 _ .
69 Ni—20 Cr-11 Si ~-0.0017 ~0.055 Surface attack to a depth of 16 mils along entire braze fitlet
100 Cu ~0.0006  -0.019  Surface attack to a depth of 0.5 mil along entire braze fillet
’ 82 Au—18 Ni . Z0.001 —0.036 Braze fillet appeared to be unattacked
80 Au—20 Cu  —0.0007 0,026  No attack on braze fillet surface
T, 90 Ni-10 P ~0.0004 w --0.013 " No attack on braze fillet surface
60 Pd—40 Ni ;0._0016 ' -.-0.06 . No attack on braze fillet surfu;:e
60 Pd—-37 Nil——-3 Si +0.0008 | +0.0'2‘7. " No attack on braze fillet surface
TABLE 6.7. RESULTS OF STATIC TESTS OF BRAZED A-NICKEL T-JOINTS IN
SODIUM HYD_ROXIDE FOR 100 hr AT 1100 AND 1500°F
v _ ' VBrcrxzihg Alloy - Weight Weight Test
= _ Composition ‘Change Change  Temperature Metallographic Netes
. (wt %) () (%) (°F) |
69 Ni—20'Cr—1l_ Si - o 1500 Braze failed completely
o +0.0024 +0.093 1100 Braze attacked completely
* ’ ' A
100 Cu ~0.0118  -0.408 1500 Braze attacked completely; large voids appear
s throughout
-0.0011  -0.038 1100 U_ni'form surface attack on braze to a depth of 3 mils
82 Au-18 Ni _0. 004 7 1500 'Nonuniform surffite .a'fl'dck on braze to depth of 1 mil
' ‘--0 0007 1100 No afiack on surfoc'e of braze
80 Au-20 Cu_-‘ V —0-0]06 ’ Umf?rm surfccelfl aflack on enhre brqze to a 2 depth of 3
. . e e mils :
_ 7 - ' “attack . ze to a depfh of T mlf
: - 60/Pd-40N| i y\ cler.m wuth cn‘tcck in fhe form of
5 oo smoli{sfrlngers runmng to a depth of 4 mils
Surface-aflqck toa depth of 0. 5 m|| A
; 'in brittle
. : N|3P phase ' ‘
. B'raze atfacked complefely, attack centered in |\l|3
s phase , e ,
- 60 F’d-—37N|—3S| +0 0023+0.083 o 1500 L Surface aftack on braze fillet fo a depth of 6 mi ||s
e R 028 1100 No cftack ";ésen'r ‘On braze surfoce o
=, oo U0 Sl f
v

ey ' 83

   

 

ER Ll . < i

 
i

 

 

 

 

 

 

ANP QUARTERLY PROGRESS REPORT

appearance of the as-received and tested TiC and
Cr,C,. These corrosion data are not selective
enough to designate whether TiC or Cr,C, is the
superior material, If the lower strength and/or
more brittle nature of the Cr,C, are objectionable,

 

 

0.03

INCH

0.0

~

 

 

 

Fig, 6.4. A-Nickel T-Joint Brazed with 60%
Pd-37% Nl—3% $i and Exposed to NuF-ZrF4-UF
(53.5-40-6.5 mole %) at 1500°F for 100 hr. Nofe
voids at the interface between the braze material
and the base metal. Unetched. 100X. Reduced
38%.

UNCLASSI FIED
¥-135

 

 

 

Fig. 6.5. A-Nickel T-Joint Brazed with 60%
- Pd-37% Ni-3% Si and A#néaled for 100 hr at
-1500°F in an Evacuated Capsule, Note voids at
interface between the braze material and the base

metal. Unetched. 100X. Reduced 40%.

84

the TiC is superior. The ZrC showed signs of
failing only in the fluorlde mrxture, the fine in-
cipient cracks in the surfaces of the ZrC particles
may be seen in Fig. 6.11. There were indications
that the fluoride mixture had penetrated the pore
spaces throughout the ZtC; the pore spaces were
not en!arged

 

Fig. 6.6. A-Nickel T-Joint Brazed with 82%
Au-18% Ni and Exposed to Sodium Hydroxide for
100 hr at 1500°F. Note slight attack at surface
of braze. Etched with potassium cyanide. 200X
Reduced 39%.

Sy
NCLASSIFIED:
Y1338t -

 

| NICKEL PLATE

 

 

 

Fig. 6.7. A-Nickel T-Joint Brazed with 60%
Pd-37% Ni-3% Si and Exposed to Sodium Hy-
droxide for 100 hr at 1100°F. Note absence of
voids at interface. 100X. Reduced 39%.
 

i

 

3

UNCLASS!F%ED
¥-13348

.04

Nj‘gx!—::i" BLATE

02

INCH

 

 

 

Fig. 6.8. ANickel T-Joint Brazed with 60%
Pd- -37% Ni-3% Si and Exposed to Sodium Hy-
droxide for 100 br at 1500°F. Note surface attack

and small voids at inferface between base material

qndubrdz'\_e alloy. ‘Unetched. ]00)(._ Réflq_éea 39%.

PERIOD ENDING DECEMBER 10, 1954

The reasons for failure of the B,C to withstand
the corrosive actions of the sodium and lithium are
not known-at this time; powder x-ray analyses of
the tested specimens of B,C indicate that B,C is
the primary phase and that a new, secondary,
unidentified phase is present as a result of both
the sodium and lithium corrosion tests, The attack
on the B,C by the fluoride mixture is not definite
or uniform. The 5 mils of attack found on one side
of the specimen was the maximum depth of attack
if the 14-mil-thick unidentified phase found on the
other side can be attributed to fluoride mixture in
a surface imperfection of the test specimen. Fure
ther tests will be necessary to properly evaluate
B 4C corrosion resistance to the fluoride mixture.

These screening, single-run, static corrosion
tests of B,C, TiC, ZrC, and Cr,C, indicate that
these corbldes, except B,C in lithium or sodium,
warrant further and more severe tests in the in-
vestigation of materials to withstand the long-term,

i e AU Wi

 

 

 

 

 

 

 

 

 

 

 

 

Fri:g. 6.9. As:Received TiC,

: 500X.

(a)

   

(») TiC Afler Exposure to Sochum at 1500°F for 100 hr.

 

Unetched.

85

 

 

 

 

 

 

 

 

 

 
 

 

  

 

- ANP QUARTERLY PROGRESS REPORT

"TABLE 6.8. SUMMARY OF STATIC TESTS OF CARBIDES AT 1500° F FOR 100 hr

 

 

A Average Weidht
t
Material Density ppar?r.x Dimensional e's . Attack : '
3 Porosity . Change . Remarks
Tested (g/cm”) Change {mils) .
(%) (%)
(%)
Tested in Sodium
B4C 2.51 3.5 ~6.5 Specimen cracked and fell apart during test
TiC ' 4.81 - 1.3 +0.3 0 0 No attack; particles bonded so well that it
: s was difficult to distinguish individual
_ particles in unetched specimen
ZrC 6.9 o 0.8 0 ‘ +0Q.1 0 No attack; particles well bondeg'jl
CryC,  6.59 2.2 . 0 ~0.4 0 No attack; particles not well bonded
Tested in NuF-ZrF4-U F4 (53.5-40-6.5 mole %)
| B4C ' +0.1 ~0.4 5 Unidentified phase on one side of
- o ‘specimen that had a maximum thickness
of 14 mils '
TiC 0 +0.1 0 No attack; as-recelved and tesfed
7 specimens not bonded nor formed so “well
o _ as those tested in sodlum '_ ' -
ZrC ' _ o +0.3 Ma;or pornon of pore spaces fnlled wnfh
‘ o fluoride mixture '
Cf3C2 _ | +0.2 No attack
Tested in Lithium
B4C | -~12,2 Specimen expanded and cracked; original,
. ' C ' dark, semimetallic luster dq”ed
TiC 0 0 0 - | No attack; mccroscopi.c color slightly
lighter
Z:C +0.2 -1 0 No attack
Cr3C2 +0.2 -1.2 0 No attack; small erack in middle of

_specimen; metallic luster disappeared

 

*The average dimensional change (%) is the average of the width, height, and length chunges; whereas the weight

change (%) represents a single value.

hi gh-temperature corrosion of sodium, hthlum, and

NaF-ZrF UF (53 5 40- 6 5 mole %),

Magnesmm-L:fluum Alloy in Wo‘l‘er
w-gvc‘b"m\'&’mfimm&_«@

An_ 80% Mg-zo% Li auozjwh]‘fi has been proz,

g as a possnble crew~compartment shleldmg
g e S,

ST A oy ez et hflwfl‘fi)‘" ol

 

86

material was corrosion tested in water at various

temperatures, Specimens of this material were

carefully cleaned just prior to testing to remove

surface films, chiefly Li N, which form on exposure
to air. A cleaned specimen was found to gain
0.088 mg/ecm? after standing in the atmosphere for
6 hr. The weight losses of specimens tested at

 

W,

ar
 

 

12

o

~0ne ‘spec1men was_tested for various l_gngths of

 
 
  

 

 

Y-12908

 

PERIOD ENDING DECEMBER 10, 1954

    

NICKEL PLATE

 

.02

 

 

 

 

 

 

 

Fig. 6.10. (a) As-Received Cr3C2. (%) C|'3C2 After Exposure to NuF--ZrFA-UF'4 (53.5-40-6.5 mole %)

at 1500°F for 100 hr. Unetched. 100X.

various temperatures in water for 15 min are tabu-
lated below:

7 Témperature Weight Loés
(°cy (mg/cm )
o e

 

S50 027 e
.80 B L
100 | 0.326

 

   
 

hme in water ‘at ‘100°C and was Found to lose
0. 292 mg/cm in the flrst 2 min of the fes’rs, whlch
was 90% of 'fhe‘\'Nél‘gh'r lost in “the ]5-m|n fest.
~ This decrec;.g in_rate of Welght loss may be at-
' :‘frnbufcb[e to 'rhe deple’r:on of hthlum atoms on the .
surfuce, since no protective fllm could be detecfed
Pure magnesium tesied in water at 100°C for 15

min showed a weight loss of 0.09 mg/cm?.

 

RUBIDIUM INVESTIGATIONS

The results of preliminary tests of the resistance
of Inconel to attack by molten rubidium were dis-
cussed in the previous report.? Static tests at

- 1500 and 1650°F have shown a maximum attack of
from 1 to 2 mils in ]00 hr. The attack has been
" both m’rergranulor and in the form of subsurface
'vmds. Since these tests indicated that under

" static lso’rhermal condltlons corrosion of Inconel

"‘by rubidium was no serious problem, a dynamic

test was conducted, The resistance of Inconel
to corrosion by flowing liquid and vaporized ru-
bidium in a closed system incorpordting a tempera-
ture gradient was determined by means of a dynamic
loop apparatys, somewhat similar in appearance

to the conventional thermal-convection loops., A
2E. E. Hoffman et al., ANP Quar, Prog. Rep. Sept. 10,

1954, ORNL-1771, p 86.

 

 

& TTF ey 87

 

 

 
 

 

 

 

 

* ANP QUARTERLY PROGRESS REPORT

 

 

  

NICKEL PLATE

 

0.001

 

0.002

 

0.003

 

0.004

 

0.00%5

INGH

 

0.006

 

0.007

 

 

-

0.008

 

 

p 4 .9‘5-‘,&‘ - !
RS | a1l ¢

Fig. 6.11. («) As-Received ZrC. (b) ZrC After Exposure to Nc:F-ZrF“-UF4 (53.5-40-6.5 mole %) at
1500°F for 100 hr. Unetched. 500X.

loop sectioned after a test may be seen in Fig.6.12.
The hot legs of the loop were constructed of 0.5-
in,-OD, 49-mil-wall tubing, and the condenser
section was of 0.25:in.-OD, 35-mil-wall tubing.
The loop was loaded with 8 cm® (approximately
12 g) of vacuum-distilled rubidium, which filled
it to the level indicated in Fig. 6.12, A vapor-
phase heat transfer system of this type will elimi-
nate mass transfer in the vapor regions. However,
the temperature gradients present in the liquid
region might conceivably cause crystal deposition,
The maximum attack occurred in the hot leg of the
loop to adepth of 1mil, as may be seen in Fig.6.13.
The attack was intergranular, and one grain ap-
peared to be ready to fall from the wall,

A schematic diagram of the apparatus presently
used to distill rubidium is shown in Fig. 6.14,

| Thus far, 3 Ib of rubidium has been purified. Since

the addition of 8% sodium will reduce the melting
point of rubidium to approximately 20°F, tentative

_ 88

 

 

 

plans call for an investigation of the corrosion
properties of the sodiumerubidium eutectic.

FUNDAMENTAL CORROSION RESEARCH

G. P. Smith
Metallurgy Division

Mass Transfer in Liquid Lead

J. V. Cathcart -
Metallurgy Division

The results of previous tests® indicated that
alloys in which intermetallic compound formation
was possible showed a marked increase in re- T
sistance to mass transfer in liquid lead as com- =
pared with the mass transfer obtained with their
pute components. Comparable behavior ‘was not s T
observed in alloys such as Nichrome V in which -

Y

 

3J. Y. Cathcart, ANP Quar. Prog. Rep. Sept. 10, 1954,
ORNL-1771, p 100.
 

 

A

RO

.

L%

o}

a3

 

T ORNL-LR-DWG 4651

  
 
 
  
 

BOILER AND
VAPOR
SEPARATOR

SECTION

    

 

 

 

ol N OPERATING
DI, N LIQUID LEVEL
O,
>
'
, S t60°F B
AT—!360°F - [.||‘1"|!\if>|;|"l||'||||'E|E|
OPERATING TIME = 312ty t 2 3 4 5

 

 

 

MAXIMUM CORROSION
PENETRATION 0.004 in.

NOTE: EQUIUBR[UM LOOF’ SURFACE TEMPERATURES
" ARE GIVEN.

 

 

Fig. 6.12. Sectioned Inconel Loop After Ex-
posure to ‘Boiling or Vaponzed Rubidium for

100 hr,

this hypofhes:s has been tested further, and, as in
earller work e_xper:m m‘sr were"perform d i Il

 

 

 

S”UCfiJl’e. mGy'eGstIy be seen in flie pho'rogrdph'

 

4), V. Catheart, ANP Quar. Prog. Rep. Dec. 10, 1952,
ORNL- 1439, P 148.

 

   

  
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
  
      
    

PERIOD ENDING DECEMBER 10, 1954

The first loops in which the 50% Fe-50% Cr

specimen was tested in liquid lead failed after

46 hr of operation with hot- and cold-leg tempera-
tures of 805 and 550°C, respectively. The second
loop failed after. only 30 hr of operuhon with hote
and cold-leg . temperatures of 805 and 500°C, re-
spectively. Microscopic - examination = showed
clearly that a phase transformation took place in
the test specimens during the operation of the
loops. As shown iin Fig. 6.16, a narrow layer of a
second phase formed at the outer surface of the
hot-leg specimens, Hardness measurements yielded
values of 162 and 1140 (DPH hardness scale) for
the new and original phases, respectively. These
values correspond closely to those expected for
ferrite and sigma phases. The possibility that
the new phase was ferritic was supported by the
fact that it was magnetic and by a chemical analy-
sis of the plugs formed in the loops., Since the
plugs were slightly richer in chromium thon in
iron, a decrease in the chromium content of the
outer layers of the test specimens was indicated.
The test temperature was close to the sigma-to-
ferrite transformation pomt ‘and therefore even a
slight alteration in the original 50% Fe--50% Cr

‘composition would produce an alloy in which only
ferrite is stable at the temperature in question.

A second alloy, 50% Mo—50% Fe, was also tested.
This alloy was chosen because an intermetallic
compound predominates at the indicated compo-
sition. Circulation of lead in this loop stopped

 after 520 hr with hot- and cold-leg temperatures

o cohpounds are presenf -”Duri‘ng fhepasfquarfer ~ of 805 and 500°C, reSpeéfiver. Examination of

this loop has not yet been completed; therefore,
the results obtained will not be discussed other
han to point out that the pluggmg fime was ap-
ely that to be expected for an c”oy in

/v/vh:ch an |htermefa[|:c compound existed, -
: Specuol comment is requured f_orwthe resui’rs ob-

 

ained with the 50% Fe—SO%

‘ndlg.cafed obove, ‘these specumens had bee-n 'rron-s-;f_'

_7foi'm'ed' almosf complefely to sngma phase pr:or to

1esh"hg, “and” ye’r very shorf operatmg tlmes were

fi'observed for the loops contammg fhem. Comparable '
loops ™ confcmmg pure iron and pure chromlum
T required 275 )
‘pluggmg occurred. On the other hand the pluggmg
“times for other alloys thot form mter ':"'etcillc come

X nd 100 hr, respechvely, before

  

 

fiounds, for example, 45% Cr—55% Co, Hastelloy B

(5% Fe-28% Mo—67% Ni), 25% Mo—75% Ni, etc.,

were all much greater than the plugging times

for loops containing their pure constituents.

89

 
ANP QUARTERLY PROGRESS REPORT

0,

 

 

a _j‘
e

 

e agi
+
\

 

 

 

Fig. 6.13. Surface of Inconel Tube Below Liquid Level in Hot Leg of Loop Shown in Fig. 6.12. Note .
intergranular attack and grain separated from wall of tube. Etched with aqua regia. 1000X.

aormE
PHILIPS ORNL-LR-DWG 4652 *
VACUUM GAUGE

 

 

  
  
    
   

 

 

t2-mm VAPOR LINE

 

 

PIRANI
VACUUM GAUGE

 

; . 3-WAY 10-mm 2 -WAY. 8-m
% . VACUUM STOPCOCK VACUUM s'TOPcrgCK

HEATING TAPE

100-ml ROUND-BOTTOM
PYREX FLASK

  

 

TO MERCURY :
ANOMETER PURIFIED RUBIDIUM .
; : , VMF 20-amp 300-m| ROUND-BOTTOM
2-WAY, 6-mm VACUUM DIFFUSION PUMP PYREX FLASK
. STOPCOCK TO HELIUM -
: SUPPLY :
S , l GLAS-COL HEATING MANTLE

TO MECHANICAL PUMP

&

 

Fig. 6.14. Apparatus for Distilling Rubidium.

90
PERIOD ENDING DECEMBER 10, 1954

 

M AL NG _ UNGLASSIFIED '| 0.00
2 , ol g _ S TN : Y- 13215 0.002
0,003
0,004
0,005
0.006
0.007
0.008
0.009
€.010
.04
0.012
0.013
0.014
0.045
0.016
0.017
0.018
0.019
0.020
.024
0.022

 

g

 

 

o

 

 

 

 

 

 

 

 

 

INCH

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

5

J ' ' .~ Fig. 6.15. A Tréfisverse Section _of a 50% Cr=50% Fe Al.!oy After a Sigma-Phase Anneul. Note ex-
tensive cracks. 200X. '

One " possible explanation for the anomalous  one of the test specimens had been completely or
behavior of the 50% Fe—50% Cr alloy is suggested partially dislodged by the erosive effect of the
by the extensive system of cracks present in the ‘circulating lead, the metal particle couid have
original fest spec mens (Fig. 6 15). The cracks  been caught in elther the hot- or cold-leg specimen
“ho dQubt lncreaéed the surface area o-vo!un:lgrcmo tubes ‘and caused a sufficient slowung down of the
| ¢ Tead flow rate to produce prémature failure of the

" loop. "Because of the excessive brlftlen ss of the

 

 

         

 

 

    
 
 
  
 
 
     
 
 
 
 
   

 

 

 

"

 

est specimens, it was not possnble to remove them
" “from the loops without crackmg them slightly; there-
foré, it is not pésSIb]e to sfcte unequwocably that
this explanation is a valid one. However, the ex-
" "“fensive cracking of the test specimens and their
“extreme friability plus the unknown effect on the
‘phase trdnsformation which ‘took place would ap-
_ “b'éc’r' to |us’rlfythe ‘belief that the r'é'su lts obtained
‘small'pl'eces cxlmosf a "’rhe touch with fifhe. 50% Fe-50% Cr alloy may not be com-
Thls increase in fnublhty was probably associated parable.fo\\\fhose obtained with other alloys in which
wn‘h the penetration of lead info the crack system intermetallic compound formation is possible.

‘referred to above. 1f a small chunk of me’rul from

 

 

R PR e

;sv ‘ot llkely fl]at such “small’ pl_ugs could' have

   

 

 

RN

  
 
 

 
   

NooT Ao - ol

Py
2o
<y

 

   

 

 

   

 

 
! A el

s i i M

 

ANP QUARTERLY PROGRESS REPORT

 

 

 

UNCLASSIFIED
Y-43576

 

0.00f

 

0.002

0.003

 

0.004

 

INCH

0.005

 

0.006

 

0.007 .

 

0.008

 

0,009

 

 

 

Fig. 6.16. Transverse Section of the 50% Fe=50% Cr Specmen from the Hot Leg of fhe Second Loop.

500X.

Fused Hydroxides as Acid-Base Analog Systems
| G. P. Smith
Metcllurgy D|V|5|on

i fundamenfcl research is to make an effective
contribution to the development of hydroxide
technology, one of the primary goals must be the
evolution of a systematic chemistry of these sub-
stances. No such systematic chemistry now exists.
lt is the purpose of this research to establish a
theoretical basis on which such a systematic

__chem:s’rry can be bu:[f.

In the flrst report on this investigation® the

-":vanous types of self-decomposition equilibria in
fused hydroxides were considered with the aid of
~ thermodynamics. In the case of sodium hydroxide

 

36. P. Smith and C. R. Boston, ANP Quar. Prog. Rép.

Sept. 10, 1954, ORNL-1771, p 102.

these equilibria may be represented as follows:

Nc120 + HZO

(?) 2N00H\ Ill_

Nc1202 +- H2

Equilibria 1 and |1 are far to the left except at very
high temperatures. Equilibrium IH favors the oxide
over the peroxide, although somewhat less so at
higher temperatures than at lower temperatures.

This report is devoted almost exclusively to the
qualitative application of two acid-base analog
theories to Equilibrium |.. This is, of course, only
a small part of fused hydroxide chemistry, and the
conclusions reached will not, by themselves, tell
very much about corrosion phenomena. Héwever,'
they are an essential part of the over-a” corrosnon
picture,

WREEE ST

 

"y

W
 

&

 
 
   

 

Tawi
.

- Subsequent reports will be concerned with some
quantitative aspects of subjects treofed qual:'ro-
tively in this report, and with Ethbrlq I and 11,
and, hence, with ox1dor|on-reducf|on processes

R

including mass fransfer. o

A large percenfoge of all chemicol reactions in
liquid systems can be classified as acid- -base analog
reactions or as oxidation-reduction reactions. In
most systems the nature of the solvent controls
the acid-base analog reaction, while in oxidation-
reduction reactions the solvent usually plays a
minor role. On the other hand, oxidation-reduction
reactions in fused ‘hydroxides are related to acid-
base equilibria in at least three ways. First, acid-
base reactions depend on Equilibrium | and oxida-
hon-reducfron reactions depend on Equrl:brlum H,
but both are :nterrelcfed by Equrl;brlum I, Second,
in some cases the ‘solvent can stabilize higher
oxldahon states of merols through solvo[yhc re-
action. Thlrd, in some cases the kinetics of the

'hydro_gen replacemenf reaction in fused sodium hy-

droxide appears to be dependent on acid-base equi-
libria in a way onologous to similar replacement
reactions in liquid ammonia. Thus, a study of
acid-base equrl;brrc in fused hydroxides is a neces-
sary prefude to a ‘more interesting study of the

- ox1dohon-reduct|on reochons which are important

in the corrosion ond moss transfer of me’rcls.

In order to focus attention entirely on the simpler
acid-base analog reactions, only Equilibrium | will
be consrdered
temperatures and in fhe absence of s’rrong omdlzmg
or reducing agents, .

lonization Assumpf'ion. As a basic assumption,

it s proposed thot fhe fueed olkah-metcd hy-‘_' ‘
he de of ’t'._‘__";_defmed as acids. They may be regorded in a
\_‘formal woy, as dlssocm’r:ng into profons ond oxide

This treatment is vohd at lower

 
 
  

PERIOD ENDING DECEMBER 10, 1954

This “‘inert” behavior of the cations in acid-base
processes stems from two factors: first, the large

‘ionic radivs and small charge of i’he'c"crione'dnd

second, the excephonol!y small lonrzahon energy

" of the parent atoms,

Solvent-System Concept. In terms of the solvent-
system concept first introduced by Franklin’ for
liquid ammonia and subsequently applied by
Jander® to a variety of liquids, it is clear that
Eq. 2 fulfills. the formal requirements for a water-
like acid-base equilibrium and, hence, that the
fused alkali-metal hydroxides may be properly
considered as acid-base solvent systems.

The remainder of this report is concerned with

_'a_ determination of the types of reactions which
acid-base theories show to be a consequence of

Eq. 2. The results are given in summary only,

‘together with enough comments to orient the reader
“who is familiar with acid-base theory.

It can be
shown that two acid-base theories, the Brdnsted
protonic-solvent theory and the Lux oxide theory,
are applicable to fused hydroxide systems. Since
the approach in this report is largely qualitative,
it will be convenient to clossfiy reactions as
either neutralization reactions or solvolytic re-
actions. Neufrollzohon reactions are those in
which the covalently solvent species occurs as
a reaction product, while solvolytic reactions are
those in which the solvem‘ species occurs as a
reactant.

Protomc-System Concept. Equation 2 represents
a proton donor-acceptor equilibrium. Hence, the
protonic-solvent theory is applicable and leads
to the following analysis. One-half the hydroxyl
ions in Eq. 2 are proton donors and, hence, are

;'lons os in Eq 3

 

SA. E. Wells, Structural Inorganic Chemistry, The
Clarendon Press, Oxford, 1945, p 350.

 

T Remhold New York, 1935.

@ OH- = H' + 0"

  
   
   
    

 

 

 

Bose |

 
  

Bcse || Acrd ll

  

    

CAcid 1

 

 

-E. C. Franklm The Nitrogen System of

 

8. Jander, Die Chemie in wasserdbnlichen L.osung-
smitteln, Springer-Verlag, Berlin, 1949,

93

4

Ve

 

 

 

 

T R

 
T

 

 

o

 

L e o

s ol s

 

 

 

]
-@
|
4
i

 

ANP QUARTERLY PROGRESS REPORT

The protons, having an exceptional polarization
potential, react almost completely with the sol-
vent, Eq. 4, represented by the other half of the
hydroxyl ions in Eq. 2. Water, the product of this

. reaction, thus may be regarded as a solvated
proton (onium species) in fused hydroxide so-

lution. Using the protonic-solvent theory, the
quantitative extent of reaction is regarded as
being controliled by the relative proton affinities

‘of the two bases involved in any acid-base equi-
~librium,

Protomc-Solvolyus-Type Reoctlons. At least
three types of protonic-solvolytic reactions might
be expected. First, the hydride ion provides a
special case: |

(6)_, - H- +_OH‘#_O“" + H,
'Secon‘d} : | | | o
(7)  OH™ + HB—>B= + H,0

where B~ is taken as a generalized representation

of an anionic species. Third,

(8) OH- + B-——> BH + O™~

The second type of solvolytic reaction requires
that the hydroxyl ion be a stronger proton acceptor
than the B~ ion. There are, of course, many
examples of this, such as the obvious reaction

(9) OH- + HCI—> CI= + H,0

which, in fused hydroxides, is a solvolytic re-
action and not a neutralization. Equation 9 repre-
sents an obvious reaction because of the ex-
ceedingly weak base character of the chloride
ion. There should, however, be a series of such
reactions involving bases with proton affinities
lying between those of CI~ and OH™, such as
HP04"" and HCO3"', so that the reaction

(10) HCO,~ + OH~—>CO,~~ + H,0

might occur; if it did occur, it would be an ex-
ample of Eq. 2.

The third type of solvolytic reaction, Eq. 8§,
requires that the B~ ion be a stronger base than
the O~~ ion. At best, there will be very few

" instances of this.

Proton n:-Neutru ||zcmon-Type Reactions. Neu-

'_i'frdlizcn‘lon reactions in fused hydroxides which
~can be treated by the protonic theory are also of
three types: (1) the reaction of water with the

oxide ion according to the formula

(11) H,0 + 0=~ —> 20H"

where the water may come from the dissociation

of another compound; (2) the reaction of water
with an ion which is a stronger base than the
hydroxyl ion but a weaker base than the oxide
ion, that is, essentially the reverse of Eq. 7,

(12) H,0 + B~—> BH + OH~
and (3) the reverse of Eq. 8§,
(13) 0~ + HB—> B~ + OH-

An obvious example of the reaction given by

Eq. 13 is 7
(14) O~ + HCl— ClI= + OH“

Kmehco“y, this reaction undoubtedly would go
by two steps,

HCl + OH- — CI= + H,0
H,0 + O~ ~—> 20H~

if the oxide ion were present in low concen-
trations, with the second step controlling the rate.
However, the equilibrium would be represented
by Eq. 14. Equation 13 represents a wider range
of reactions than Eq. 8 because the permissible
proton affinity range for B~ goes all the way up
to that of the oxide ion; therefore the reaction

(15 HCO,~ + 0=~ —> CO,~~ + OH-

is undoubtedly an example in which the bicar-
bonate ion behaves like an acid.

Oxidic-System Concept. There are a great many
reactions in fused hydroxides which cannot be
usefully treated in terms of the protonic-solvent
theory. Consider, for example, the familiar re-
action '

(16) €O, + 20H~—> CO,~~ + H,0

The application of the Brdnsted theory here re-
quires a rather complex, formal treatment. The
actual proton exchange is between the bases OH™
and O~~. The weaker base, OH~, serves as the
proton acceptor, and the stronger base, O~ ~,

" serves as the proton donor. This is the reverse

of the neutralization reaction given in Eq. 11 and
would not be predicted by the protonic-solvent
theory. This seeming contradiction may be re-

30
i . .J
g
3

i,

L]

¥
 

"

1]

e

k)

)

B

W

solved in a formal way, but it is much more useful
to apply the following straightforward approach.
In Eq. 16 the greater oxide-ion affinity of carbon
dioxide compared with water determines the di-
rection of the reaction.

Lux? and Flood and Férland!® have advanced
a definition of acids and bases which was de-

signed for tréc'fi_ng the special problems en-

countered in nonprotonic fused oxide systems.
According to these authors, a base is an oxide-ion
donor and an acid is an oxide-ion acceptor; thus

(]7) Bcse—\ Acid + O™~

Under this defmmon the convenhonol classifi-
cation of slags as acidic or basic is preserved.
This acid-base concept has not been applied to
solvent systems, but, by a small extension, this
may be done.

Two competing equilibria of the above type can
be considered:

AC ————-“AH + O--

 

 

o~~ + B“—‘“_ABO

(18)
AO + B ¥ BO + ATt
Base | Acid |l Base Il Acid |

The acid-base éqbilibrium shown by Reaction 18

"may be considered to be controlled by a balance

between the oxide-ion affinities of the reactant
acid and the product acid in a manner that would
be analogous to the balance between the proton
affinities of the bases in protonic solvents,

The self- dlssocu:mon of hydroxyl ions according
to Eq. 2 completely fu[fi“s the requsrements of

_the Lux deflmtion, _Eq 17. It may, Gf f"'S" seem_:__

 

   
 
 
 
 
 
 
 
 
 
 
  
 

 

SUIASENT

o fwo hydroxyl |ons the

 

H. Lux, Z Elektrochem 45, 303 (1939).
10
592, 781, 790 (1947).

 

 

H. Flood and T. Forland, Acta Chem. Scand 1,

PERIOD ENDING DECEMBER 10, 1954

Thus, water, H-O-H, may be regarded as a pyro-
analog compound of hydrogen and the hydroxyl
ion as the ortho-oxyanion of hydrogen. Because
of acid-base relations involving pyroions, such
as shown in Eq. 19, two hydroxy! ions will always
be taken together to serve as a Lux base analog.

Oxidic-Solvolysis-Type Reactions. The solvo-
lytic reactions involving the oxides and oxysalt
solutes can be classified according to the ioni-
zation energy of the atom other than oxygen in
the oxide. This method of classification is
possible because the ionization energy of an atom
is related to its oxide-ion affinity. It is con-
venient to divide the elements into four groups
according to ionization energy.

First, oxides and oxysalts of atoms having high
ionization energy should react with the fused
hydroxides to give a change in coordination. For
example,

(200 SO, + 2CH™ &= SO,== + HOH

(see also Eq. 16).

Second, oxides and oxysalts of atoms of medium
ionization energy should react with fused hy-
droxides either to form or to depolymerize poly-
anionic acids. For example, the reaction

(21) P,0,4= + 20H-==2P02~ + HOH

will probably take place. The possible reactions
of this type are numerous because of the sub-
stantial variety of polyanions, For example, there
could be the step-wise depolymerization of silicon-
oxygen compounds beginning with the three-di-
mensional network in silicon dioxide, passing
through the two-dimensional net polyanions,

" through ‘the one-dlmensronu] chain (and ring) poly-
““anions, the pyrocm:on, cnd funully endmg wrfh
" the orthosilicate anion.

This second type of solvolytic reaction should

““tend toward depolymerization and, for some parent
“\'otoms, go all the way to the ortho-cxyumon be-
““¢ause of the relatively high basic character of
" the hydroxyl ion. However, there is no assurance
“+ that the ortho-oxyanion will be the most stable
<= form in all coses.

Third, oxides of atoms of low lomzcmon energy
shq_qld react with fused hydroxldes to form oxy-

“salts Th:s represen’rs a conversion from a catlon

 

to an cmlon. For example,

(22)  NiO + 20H-==Ni0,~~ + HOH

95

sy
v v
o T3y s
\,,ai i S Q
A ? %.“ ‘-',.}

 

 

 
it i

 

 

ANP QUARTERLY PROGRESS REPORT

 If a meta-oxysalt rather than an oxide is used as

the starting material, in some instances, the meta-
oxysalt may be converted into the ortho-oxysalt.

Fourth, oxides of atoms having very low ioni-
zation energy should not undergo solvolysis at
ail. For example, the inert behavior of magnesium
oxide toward sodium hydroxide is no doubt due to
the low ionization energy of magnesium.

There still remain a large number of potential
solutes nof treated above. Of these, the only ones
which the writer has considered are the relatively

';tomc metolhc ‘salts,  These substances will
“probably be separated into ions by the highly

dipolar hydroxyl ions, and the solvolysis of the
cation and the anion can be considered separately.
The cation should tend to form oxides and oxy-
salts. This reaction should go virtually to com-

- pletion except for metals of very small polarization

potential.  The anionic-solvolytic reaction can

'”vary from very complex to nil, and an adequate
" “treatment will require the development of a some-

what more generalized acid-base theory.

'As an example of the reaction of a metal salt
with a fused hydroxide, a small amount of nickel
chloride was added to fused sodium hydroxide at
400°C. The reaction proceeded rapidly with the
evolution of gaseous water and the formation of
a fine black precipitate. The precipitate was
separated from the hydroxide and found by x-ray
analysis to consist largely of nickel(ll) oxide,
together with a small amount of an unidentified
compound.

This reaction can probably be represented, in
large measure, as

(23) NiCl, + 20H™ <= NiO + HOH + 2CI~

Previous . experiments indicate that at higher
temperatures the reaction represented by Eq. 22
would have followed that represented by Eq. 23.
Oxidic-Neutralization-Type Reactions. The neu-
tralization reactions of nonprotonic, oxidic solutes
can be divided into two classes. First, the
neutralization of a base by water takes place

according to the reaction

(24)' BO + HOH—> B** + 20H-

Thls reaction will only occur for nonprotonic

1ac1ds, B** which are stronger than water.

Second, reactions of the following type, which

“décur in pure oxide melts, would represent a
- special type of pseudoneutralization in which no

%

solvent would be produced

(25) Acid + O~ — Base

The reaction shown in Eq. 25 includes the entire
host of reactions treated in the original Lux
theory.

When reactions of the type given in Eq. 25 occur
in fused hydroxides, the oxide ion concentration
can be varied over a wide range of values. This
is not possible with the pure oxide melts to which
the Lux theory has previously been applied. Ex-
amples of the classes of reactions which are of
the type represented by Eq. 25 may be obtained
by substituting O~~ for 20H~ and deleting H,O
in Eqs. 20 through 23. The essential difference
between reactions of the type of Eq. 25 and those
given in Egs. 20 through 23 is that water is
present in the latter reactions but not in the
former. The presence of the acid of medium
strength should shift the equilibrium considerably
for much weaker acids, that is, those derived from
atoms of low ionization energy, but only slightly
for stronger acids, those derived from atoms of
high ionization energy.

Mixed-Type Reactions. Finally, reactions of
substances of what may be called ‘‘mixed’’ types
should be mentioned. These mixed reactions
might involve protonic-oxidic salts of alkali
metals, and protonic or oxidic salts of nonalkali
metals, or combinations of both. The resulting
reactions should be capable of analysis in terms
of the same considerations used in arriving at
the conclusions already stated.

The Bronsted protonic theory was originally de-
veloped to treat the more useful low-temperature
solvents, such as water and liquid ammonia, while
the Lux theory was originally developed to treat
the metallurgically important, high-temperature,
fused-oxide systems. It is of some theoretical
interest that both theories may be rigorously
applied to reactions in fused hydroxides, although
the types of reactions for which these applications
are useful form two sets of almost mutually ex-
clusive reactions — one set for each theory. This
is symptomatic of the need to develop a more
general acid-base theory along the lines proposed
by Audrieth.!! For many solvent systems, such a
theory would be a luxury; for fused hydroxides,
such a theory is a necessity.

 

TIL. F. Audrieth and J. Kleinberg, Non-Aqueous Sol-
vents, Wiley, New York, 1953, p 2723,

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It should be noted that of the types. of reactions
discussed above only a few examples have been
studied experimentally in a detailed or quantitative
way; most have not been studied at all or else
are represented experimentally by some of the
more obvious cnd less m'rereshng exompies.

Flammublllty of Alkah Metal Sofuhons at
ngh Temperatures

G. P. Sml'rh M. E. Sfeldlltz
Mei_'q‘“‘urgy Division

In the pre'c:‘eding-hirfi‘epwon‘12 data were given for
the flammability of alloys of the sodium-bismuth
system, Additional data have now been obtained
for bismuth-rich alloys that provide a much more
complete picture of the reactivity of this system.
The relative reactivity of sodium-bismuth solutions
with dry air at 700°C as a function of the mole
fraction of sodium is shown in Fig. 6.17. The
relative reochvuty scale is “arbitrary. Pure sodium
was assigned a reactivity of four units. Unre-
active solutions were assigned a reactivity of
zero. No measurements were made for mole
fractions of sodium between 0.6 and 1.0 because
a solid compéund"6"frwébmpositi'o'r'1'-NG;Bi'-p'recipi-
tates from solution at a mole fraction of sodium
somewhat greater than 0.6.

A temperature-composition diagram for liquid
sodium-bismuth soluhons is presented in Fig.
6.18, on which is ploh‘ed a curve fhc’r approx:-

 

2y, E. Steid|ifz, L. L. Hall, and G. P. Smith, ANP
Quar. Prog. Rep. Sept. 10, 1954, ORNL-1771, p 102.

i, UNCLASSIFIED

 

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6.17, Relative Reactivity of Sodium.
Bismuth Solutions with Dry Air at 700°C as a
Function of the Mole Fraction of Sodium.

 

 

    

PERIOD ENDING DECEMBER 10, 1954

mately separates the region of reaction from the
region of no reaction. The circles and triangles
on this diagram represent the temperature-compo-
sition values at which tests were conducted.
Most of the circles represent two to four tests
each. The open circles represent tests in which
reaction occurred, while the black circles repre-
sent those which showed no reaction. The black
triangles represent conditions under which part
of the tests showed no reaction, while the rest
of the tests showed a slight reaction. The two
open triangles represent tests which showed an
exceedingly slight amount of reaction. The broken
portion of the curve is poorly defined, inasmuch
as it may actually pass beneath the open triangles
rather than above them, as shown. It may be noted
that jets of pure bismuth showed appreciable re-

activity at 800°C but not at 750°C,

When air was saturated with water vapor, the
line of zero reactivity was shifted toward lower
sodium concentrations by a small but appreciable
amount for sodium-bismuth solutions. This effect
is shown for sodium-bismuth solutions at 700°C
in Fig. 6.19. It may be seen that, although data
for moist air scatter badly, there is unquestionably
a small shift toward lower sodium concentrations.

Combustion studies have been run on four of the
six alkali metal—alkali halide systems for which

UNCLASSIFIED
ORNL-LR-DWG 4654

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

900
REACTION
g
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600 . ol
NO REACTION .\
500 .
0 0.2 0.4 . : 0.6

  
   

MOLE FRACTION OF SODIUM

Temperature-Composition |D‘|c|gmm
SHmeg Reglons Within Which Jets of Sodium-
Bismuth Solutions Did and Did Not Recct with
Dry Air,

97

 

 

 

 

T R T

 

 
R T

 

 

ANP QUARTERLY PROGRESS REPORT

phase diagrams are available. The results de-

scribed in Table 6.9 represent one test of each

- system at the temperature indicated. The temper-
‘ature chosen lies in the one-liquid phase region
of the phase diagram — generally about 50°C

cb'éve' the two-liquid phase region. As may be
seen, all tests were conducted in the metal-poor
sec'non of the system.

UNCLASSIFIED
" ORNL-LR-DWG 4655

 

 

 

 

 

 

 

 

 

 

 

 

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MOLE FRACTION OF SODIUM

Fig.‘ 6.19. Relative Reactivity of Sodium-
Bismuth Solutions with Moist Air at 700°C. The

curve represents data for dry air.

TABLE 6.9. FLAMMABILITY OF ALKALI
METAL-ALKALI HALIDE SYSTEMS

 

Metal Content Test Temperature

 

 

System (mole %) ©C) Reaction
Na-NaF 2 1050 Mild
5 ‘ 1050 Violent
10 1050 Violent
NaNaCl 2 850 Little
6.6 900 Violent
30 1050 Vielent
Na-Ngal 2 700 Mild
5 800 Mild
10 900 Violent
30 1000 Violent
K-KF 2 900 Mild
5 900 Violent
- 10 900 Violent

 

98

~ CHEMICAL STUDIES OF CORROSION
F. Kertesz o |
Materials Chemistry Division

Effect of Temperature on the Corrosion of
Inconel Melts with NiF , Additions

H. J. Buttram R. E. Meadows
N. V. Smith
Materials Chemistry DlVlSion

In order to sfudy specn‘tcclly the removal of
chromium from Inconel by a known corroding agent
in fluoride melts, additions of NiF2 were made
to NoF-ZrF ,-UF (53.5-40.0-6.5 mole %) in Inconel
capsules. These capsules were subjected to
100-hr static tests in the temperature range 600
to 1000°C.

When the chromium concentration of the melt
after the tests was plotied against the amount of
NiF. added to the melt, a temperature effect at
a relatively high concentration of NiF, (690
meq/kg) was noted. Additions of NiF, up to about
70 meq/kg yielded chromium concentrations which
were independent of temperature to within +100
ppm Cr**. The mixture to which the large addition
of NiF, (690 meq/kg) was made showed, at 600°C,
a chromlum content of 280 ppm Cr +4 that was
considerably below the stoichiometric quantity;
at 800 and 900°C, there were successive increases
to 630 and 840 ppm Cr**; while at 1000°C, 550
ppm Cr** could be found. Metallographic obser-
vations paralleled the chemical analyses. Pos-
sible explanations for the finding of decreased
chromium content at 1000°C were set forth ‘in a

previous report,'3

Effect of Chromium Additions on f|1e Corrosmn of 7

Inconel in Fluoride Melts with NIF ‘Additions
H. J. Buttram R. E. Meodows

Materials Chemistry Division

The behavior of Inconel capsules exposed to
NaZrF, and to NaF-ZrF -UF, (53.5-40.0-6.5
mole %) was studied by using 100-hr tilting-furnace

 

131, J. Buttram, R. E, Meadows, and N. V. Smith, ANP
Quar. Prog. Rep. Sept. 10, 1954, ORNL-1771, p 108.

 

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tests in order to establish the effect of odded
nickel fluoride. In the absence of added chromium
metal, the plot of the chromium concentration after
test vs the nickel fluoride added followed a linear
relationship, although the amount found in the
case of very large additions of NiF, was smaller
than stoichiometric. Metallographic tests showed
a heavy subsurface void formation to a depth of
21 mils at the hot end (800°C) and 5 mils af the
cold end (650°C), while the corresponding blank
capsules (no NiF,) showed only light attack to a
depth of 1 to 2 mils.

A series of Inconel capsules containing the
same two fluoride melts and NiF, with chromium
pellets added were subjected to fi'\e same type of
testing.  The after-test chromium values were
above stoichiometric, and metallographic exami-
nation of the capsule walls indicated that the
NiF., additions had little effect on the amount
of attack. The heaviest attack was to a depth of
3 mils, with the average being 1 to 2 mils. The
cold ends had thin metal depesits and a slight
void formation.

 

 

PERIOD ENDING DECEMBER 10, 1954

Effect of Chromium Valence State on Corrosion of
Inconel

H. J. Buttram R. E. Meadows
Materials Chemistry Division

Additions of chromous and chromic fluorides
were made to NaZrF_ and NaF-ZrF4-UF4 (53.5-
40.0-6.5 mole %) in Inconel capsules to determine
the effect of chromium valence state. Metallo-
graphic examination showed that chromous fluoride
additions had essentially no effect even in the
case of large additions (18,000 ppm Cr**). The
depth of attack was about 1 mil. As would be
expected, when chromic fluoride was added, the
Inconel showed attack to a depth of 4 mils and
some evidence of intergranular penetration. The
chromous ion has been demonstrated as the stable
state in the NcF-Zer-UF4 systems; therefore,
it is reasonable to assume that, when chromic
fluoride is the additive, chromium metal will be
removed from the Inconel to enable the chromium
to be reduced to the chromous state. Chemical
analyses made after these tests were inconclusive
and difficult to interpret.

99

 
 

 

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i Bk ek iz oo S

 

' ANP QUARTERLY PROGRESS REPORT

7. METALLURGY AND CERAMICS

“W. D. Manly

J M. Wordé

Metallurgy Division

The mckel-molybdenum base alloys are bemg

“studied extenswely as possible reactor structural
"4mu'rer|als ‘with qualities superior to those of
_lnconel Efforts are under way to evaluate and

improve the existing alloys and to develop better

.ones. Radjator test assemblies have been fabrie
“cated, and weld stability and mechanical properties
~are being studied. Work has continued on the

preparation of duplex tubing, boron carbide shield-
ing, tubular fuel elements, and other special
materials., The results of oxidation resistance
tests of brazing alloys are presented. In addition,
the first produchon of beryllium oxide ceramics by

‘castmg from a sllp which is basic is described.

- DEVELOPMENT OF NICKEL-MOLYBDENUM
BASE ALLOYS

The investigdtions under way for the evaluation
of nickel-molybdenum base alloys as structural
materials for circulating-fuel reactors include
attempts to improve the ductility and the fabrica-
bility of commercially available Hastelloy B through
purification, and efforts to find another suitable
and improved nickel-molybdenum base alloy that
has the strength and corrosion resistance of
Hastelloy B. Fabrication experiments are being
used to determine the effects of various treatments
on the materials developed,

Fabrication Experiments

H. Inouye J. H. Coobs
Metallurgy Division
One of the difficulties experienced in the fabri-
cation of Hastelloy B is due to the narrow range
of forgeability of the alloy — between 1950 and

'2100°F. The upper limit of forgeability is due to

impurities which cause grain-boundary melting.

“The low ductility at temperatures between 900 and

. 1800°F is less understood, but it is believed to be

" due to aging and impurity precipitation in grain

boundaries.

. to eliminate tfrace (tramp) elements by vacuum

.’___:'me[hng and the addition of elements to neutralize
" their effects.

Therefore efforts are being made

One of the most critical components of a circu-

-r‘rj"."‘rlqfing-'fuel reactor will be the fluoride-to-NaK

100

heat exchanger, for which, because of its com-
plexity and fragility, it will be highly desirable to
have seamless tubing. Commercially available
Hastelloy B tubing, which is now made by welding
strip, is not satisfactory because of cracks in the
weld, checks on the surface, and, in the dbse'nce
of severe working of the weld, nonuniform proper-
ties, Therefore several extrusion experiments have
been performed in an effort to produce seamless
tubing.

The initial extrusion experiments were attempts
to produce rod from air melts of Hastelloys B and
C. When extrusion temperatures above 2100°F
and up to 2350°F were used, the a”oys fractured
severely in all instances. For the subsequent
experiments, the extrusion temperature was lowered
to 2100°F, and a reduction ratio of 6.3:1 was
established. The results obtained for the various
materials are summarized in the following.

Vacuum-Melted, As-Cast Hastelloy B. The
ingot that was extruded had been homogeneized
at 2100°F for 48 hr. The front of the extrusion
fractured severely, but some improvement was
noted in comparison with the air melts previously
extruded,

Air-Melted, As-Cast Hastelloy B Plus 0.2% Ti
as Ti-Mn-Al-Ni Master Alloy. The ingot was
homogeneized 48 hr at 2100°F before extrusion.
The exirusion fractured severely, and no improve-
ment in comparison with the vacuum melt was
noted.

Air-Melted, Commercial, Wrought Hastelloy B.
The extrusion fractured severely and thus indicated
that the cast structure was not responsible for
the poor hot forgeability of the alloy.

Air-Melted, Commercial, As-Cast Hastelloy C.
The extrusion was made through @ cascade die
(cone plus shear) to determine the effect of die
design. Thus far, the alloy has been extruded
through cone dies varying from 45 to 25 deg and
through a shear die. In all exfrusions, the rod
fractured severely. An extrusion was also made
by containing the billet in an Inconel can 0,063
in. thick on the outside diameter and with a %-in.-
thick disk at the nose of the biilet. The Inconel
nose separated from the Hastelloy, but the Hastelloy

0

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SN
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~ the extrusion.

was fractured only on the front 1 in.” The remainder
of the extrusion appeared to be sound. A thin
layer of Inconel was found on the whole length of
This experiment has not, as yet,
been fully evaluated.

' NlckeI-Molybdenum Blnury Alloys. In contrast
to the experiments with the Hastelloys, previous
extrusion experiments with vacuum melts of the
binary alloys in the cast condition showed marked
improvement in the quality and ease of fabrication.
The tendency to fracture was still apparent to a
slight degree and became progressively more
noticeable as the molybdenum content of the alloy
was increased from 20 to 24%. Two extrusions of
tube blanks of the 24% molybdenum alloy were
unsatisfactory for further reduction to small-diameter

tubing because of the large fractures and deep-

defects. 1t was found that the defects could be
eliminated by homogeneizing the cast billets for
48 hr at 2100°F, The minimum homogeneizing time
was not determined. The data obtained in the
extrusions of the nickel-molybdenum binary alloys
are presented in Table 7.1.

PERIOD ENDING DECEMBER 10, 1954

Since Hastelloy C could not be extruded with
any degree of success until it was canned
Inconel, it is concluded that lubrication is an
important factor in these experiments, At the
temperature of extrusion, although the alloy is
probably notmolten, the grainboundary is weakened
sufficiently by the friction of the flowing metal
against the die to cause fracture. By providing
a layer of Inconel between the die and the alloy,
as by canning, the friction on the alloy is removed.
The friction could also be reduced by increasing
the extrusion temperature, but, to increase the
extrusion temperature, it will be necessary to
eliminate the lowsmelting-point impurities from the
alloys. An alternative would be to reduce the
extrusion temperature and the extrusion ratios.

The following tentative conclusions have been
drawn: (1) Vacuum melting has little effect on the
extrudability of the alloys. (2) The extrudability
of the alloy does not depend on whether it is in
the wrought or the as-cast condition, (3) Homoge-
neizing has a beneficial effect, as evidenced by

alloys of the24% Mo~76% Ni composition. (4) Poor

TABLE 7.1. EXTRUSION DATA FOR NICKEL-MOLYBDENUM BINARY ALLOYS

Atmosphere: Houghton’s salt bath No. 1550

Heating Time: tube blanks, 30 min; rod, 45 min

Billets: as-cast, 3 in. in diameter and 3 in. long, homogeneized 48 hr in helium plus hydrogen

Die: 30- to '45-deg cone die, olloy CHW
Mandrel: 1-in. stem, straight, olloy LPD

Lubrication: glass wool in container, Fiberglas sleeving on mandrel stem

Dummy Block: 70-30 brass coated with Necrolene

Pressure 'R'equiremems (on 3-in. ram}: maximum, 700 tons; minimum, 487 tons

 

Composlflon R T I R AT SOy

- Temperature

~ Heat No. (Wf %)

 

 

Extrusion

Remarks
Ratio

 

 

 

' '_':'Two fube ex'rruswns, good surfuces
“HIOne rov ext.rus:on, good surfoces o
B f'One fube extrusnon, good surfoces |
: Two fube extrus:ons, minor f|ows '

One rod exfrus:on, good surfoces
VEOne ’rube extrus:on, shqttered o

One tube extrusion; minor defect 2 in.

 

R FR A A from front
24 32 68 2150 6.3 :1 One rod extrusion; failed to extrude
101
.{"2 Qo i o
v g"

 

 

 

;,E

 

 

 
 

 

 

 

ANP QUARTER[Y PROGRESS REPORT

fdbr‘icabilify of the cast alloy is associated with
a second phase which persists after long-time heat
treatments at 2100°F.

New Alloys

H. Inouye J. H. Coobs
Metallurgy Division

Several new alloys arebeing studied in an attempt
to find an ulloy that is superior to Hastelloy B,

“and efforfs are ‘béing made to improve the existing

Hastelloys. Prelammary results of experiments

o ”:f.w:’rh aHoys of various compositions are presented
“in the following.

Hastelloy B Plus 0.03% Cerium. A 3-Ib vacuum
melt was rolled at 2100°F, and there was con-
siderably less cracking than would have been
found with Hastelloy B. Numerous small defects

. _were present, however, that may be eliminated by
- rincreasing the cerium content of the alloy.

" Molybdenum-Columbium-Nickel Alloy. A 20%
Mo—-5% Cb-75% Ni alloy could be cold rolled from
the cast structure to sheet, The structure is two-
phased at 1500°F and single-phased at a solution
annealing temperature of 1950°F. The tensile
strength of the alloy at 1500°F is 36,800 psi
(elongation 3.5%). A creep test at a temperature
of 1500°F and a stress of 8000 psi is in progress.

Molybdenum-Aluminum-Nickel Alloy. A 20%
Mo-2% Al-78% Ni alloy could be rolled at room
temperature from the cast condition. At 1500°F
the tensile strength was 43,500 psi with an elonga-
tion of 5%.

Molybdenum-Nickel Alloy, A 100-g arc melt of
a 24% Mo-76% Ni composition was hot rolled
60% at 2200°F, Following a I-hr anneal at 2100°F,
the alloy was cold rolled 72% without cracking.
The work hardening data for the alloy are listed
below:

Cold Work (%) Hardness YPN (20-kg load)

Annealed 200
25 347
29.6 | 401
405 423
s 432
0.4 483
720 470

R

A 3-1b vacuum melt of this alloy has been rolled
to sheet and will be screened for strength and
ductility in creep and tensile tests. In addition,
an extrusion of the alloy has been rolled to sheet
for obtaining more detailed data,

A 100-g arc melt of a 32% Mo-68% Ni composition
was successfully hot rolled 60% at 2200°F, The
alloy was further cold rolled 72% without cracking.
The work hardening data for the alloy are listed
below:

Cold Work (%) Hardness V PN (20-kg load)

Annealed 231
21 386
31 444
41 471
54 513
62 502
72 516

A 3-lb vacuum melt of the alloy has been rolled
to 0.063-in. sheet and is being machined into
creep and tensile test specimens.

A 100-g arc melt of a 38% Mo~62% Ni compo-
sition was rolled 60% at 2200°F. The alloycracked
when cold rolled. In the annealed condition the
hardness of the alloy is 396 VPN, The micro-
structure of the cast alloy shows a considerable
quantity of eutectic, which appears to be similar
to that found in Hastelloy B.

The following additional arc-melted alloys were
evaluated:

Alloy Results

75% Ni—20% Mo—5% Al

70% Ni-20% Mo-10% Cb Slight edge cracking during
hot rolling, further cold
rolled to sheet without
difficulty

Cracked during hot rolling

78% Ni~20% Mo—-2% V Successfully hot rolled and
cold rolled to sheet with
no visible flaws

78% Ni~20% Mo—2% Zr Moderate amount of cracking
during hot rolling that be-
came progressively more

severe during cold roelling

A
 

D e ok i B

 

Ay

Oxidation and Oxidation Protection

H. lnouye J. H. Coobs
Metollurgy Division

Temporary cychng tests from 1500°F to room
temperature in air have been performed on Hastelloy
B and other nickel-molybdenum base alloys to
determine the behavior to be expected in a NaK-to-
air heat exchanger such as will be required for
aircraftreactor application. The severest operating
condition that could exist would be a thermal
cycle from operating temperature (1500°F) to room
temperature under stresses caused by high-velocity
air.

The cycling tests of Hastelloy B were evaluated
by daily visual examination of the test piece which
had been through a heating cycle of 17 hr at 1500°F
and then coolingto room temperature in the furnace.
Weight changes were also measured daily. In
conjunction with these tests, identically prepared
specimens were tested at a constant temperature
of 1500°F, The results of these initial tests are
shown in Table 7.2 and in Fig. 7.1.

The results presented in Table 7.2 show that
the loss of metal from the Hastelloy B specimen
through oxide spalling was prevented by the use
of platings. in general, aluminum and aluminum
plus nickel are considered to be unsatisfactory
platings, however, because of spalling of the
protective coating.

Several other alloys have been tested for oxi-
dation at 1500°F for comparison with Hastelloy B
and to determine whether they need oxidation
protection during elevated-temperature mechanical

and corrosion tests. The results of the tests are

showp_ in Fig. 7.2.

 

PERIOD ENDING DECEMBER 10, 1954

At a constant elevated temperature in air, nickel-
molybdenum alloys of intermediate compositions
form a protective coating of NiMoO,. Upon cooling,
regardless of the rate, the coating will spall vigor-
ously at temperatures between 440 and 230°F.°
Experimental evidence shows that this spalling
may be due to a change in the crystalline structure
of the NiMoO, coating.

Radiator Fabrication

P. Patriarca K. W. Reber
R. E. Clausing G. M. Slaughter
Metallurgy Division

J. M. Cisar

Aircraft Reactor Engineering Divisicn

R. L. Heestand
Pratt & Whitney Aircraft

A series of test specimens of Hastelloy B radi-
ators has been fabricated for use in evaluating the
ability of Hastelloy B to withstand thermal shock
and oxidation., A specimen that includes approxi-
mately 3 in. of Inconel-clad copper fins spaced
15 fins per inch is illustrated in Fig. 7.3.

The 20 tube-to-header joints were inert-arc
welded by using the semiautomatic welding equip-
ment described previously.? The manifolds were
manually inert-arc welded, and the assembly was
brazed to effect the tube-to-fin joints and the

 

J W. Spretnak and R. Speiser, Protection of Mo
lybdenum Against Corrosion at High Temperatures,
QOhio State University Research Foundahon Status
Report No. 8,

2P Patriarca et al., ANP Quar, Prog. Rep. June 10,

' '1954 0RNL-]729 Flg. 65, p Y6,

 
  

TABLE\‘7.2 ) RESULTS OF CYCLING TESTS IN AIR oN HASTELLOY B
g ,AND PLATED HASTELLOY B SPECIMENS

 

 
 

 

We'igh'r'Chunge iri

 

 

 

 

 

L ,inf_quingJon Hustéilqy B T gy CYdies Surface Condition
- -0.0775 Spalled
™ of chromium ST seomsT 0 Nospalling
_0 00] |n of nlckel p|us 0 0001 in. of chromium +0.0035 No spealling
0. 001 ine of mckérl‘ plus a!ummum spray TUR0.0512 " Spalled
Aluminum spray +0.0370 Spalled
103

 

 
ANP QUARTERLY PROGRESS REPORT

WEIGHT INCREASE (10~ 3g/cm?)

 

 

0 20 40 60 80

backing up of the tube-to-header joints with Coast
Metals_No. 52 brazing alloy.

 

The fabrication of the three test assemblies
made to date has been routine, and no unusual
behavior or difficulties have been encountered,
Although it is expected that the characteristic
aging of Hastelloy B and the subsequent loss in
room-temperature ductility will affect the longevity
of these test specimens, the first assembly has
withstood, without failure, as determined by a
hehum Ieok test, 100 accrued hours of life with
~one air quench and one water quench directed at
the fin surfaces. Othe_r test assemblies are being
'fobncofed with type 310 stainless-steel-clad

 

~ the evaluation of Hastelloy B, the dimensional
~ stability and oxidation resistance of this material
- under thermal shock.

104

 

 

copper in order to determine, in conjunction wr’rh‘

UNCLASSIFIED
ORNL~LR—~DWG 4426

Ni+Al PLATED HASTELLOY B

Al PLATED HASTELLOY

HASTELLOY B

Ni+ Cr PLATED HASTELLOY B

Cr PLATED HASTELLOY B

100 120 140 160 180 200
ELAPSED TIME (hr)

Fig. 7.1. Oxidotion of Plated Hastelloy B in Air ot 1500°F,

Welding

P. Patriarca K. W. Reber
R. E. Clausing. G. M. Slaughter
Metallurgy Division

J. M. Cisar

Aircraft Reactor Engineering Division

R. L. Heestand
Pratt & Whitney  Aircraft

Preparations are under way for experiments to
evaluate the properties of Hastelloy B weld metal
as affected by thermal history. A comparative
basis for the tests has been prepared through
experiments in which the effect of aging time at
1500°F on the room-temperature ductility of wrought
Hastelloy” B was determined For these tests,
guided bend specimens, /8 X ]/2 X 13/ in,, were
cut from /16-|n.~fh1ck wrought Has'relloy B plate

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"

s

" Hastelloy B whenwag’é:

PERIOD ENDING DECEMBER 10, 1954

UNCLASSIFIED
ORNL—LR—DWG 4427

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

14 |
12
//‘
‘° <
~ 7 20 % Mo —80% Ni
Sa /
o g s
Q A"
% ' ///
© /ln ‘ HASTELLOY B '
&6 — ' 20 % Mo —5 % Cb — 75 % Ni —
£ y 20% Mo — 2% Al — 78 % Ni
‘ A |
| - =
4 // ] V’
& T | {fl_—&-—-
A o : _..,-KJ——V-A‘ -
//.’/ : : e O ’——I——— et ——— =
2 ——l = ___-___.—- = .
./7_ 1T " \ 24% Mo ~ 76 T Ni
P
7(;‘ = 32% Mo — 68 % Ni
O .
0 20 40 60 80 100 120 140 160 180 200

ELAPSED TIME (hr}

Fig. 7.2. Oxidation of Nickel-Molybdenum Base Alloys in Air at 1500°F

which had been solution annealed at 2150°F and  a period of exposure of less than 50 hr. If the
quenched. One set of specimens was then aged material is given o stabilizing heat treatment,
at 1500°F in vacuum for various perlods of up to "'however, this reduction in ductility is prolonged
300 hr before feshng A" second group oF the "T“for a period of between 100 and 200 hr. The
spec:mens ‘wds’ heat treated in hydrogen for /2 hr results of this preliminary investigation indicate

. at 2]50°E and then_ for. 30 hr “at 1950°F and wa’rer ___that the application of a stabilizing heat treatment
’quenched bef ]

     
    

ficml cnd that longer treatments at 1950°F
might effect more complete stabilization,

‘A similar group of specimens is being cut from
inert-arc butt-welded %wn.-‘rhuck plate in such a
"'M‘:'mcnner that the weld metal will constitute the
'ter T/4 in. of fhe bend specimen. Thfea effect of
g9

   

re‘wfhey were aged cn‘ ]500°F

 

   
   
   

 
  
 

”‘Aerc’rure ducflllty of toom-temperature ‘and 1500°F bend tests, and the
‘fa'r a simulated service  results will be compured wuth those for wrought
temperature of 1500°F is markedly reduced after  Hastelloy B.

be seen that the

105

 

ng-| g time ‘@nd prior thermal hlsfory on these
”"""geomefry and are presenfed in Table 7.3. If'ymaf”'"i'fi'él:'f-arc ‘welds will be determined by using both

 

TomurErT

 

 
 

 

 

 

 

ANP QUARTERLY PROGRESS REPORT

;UNCLASSIFIED

Y3617

 

Fig. 7.3, NaK-to-Air Radiator with Hastelloy B
Tubing ond Manifolding and a 3-in. Section of
High-Conductivity Copper Fins Spaced 15 Fins
per Inch.

Mechanical Properties Studies

R. B. Oliver D. A. Douglas
J. H. DeVan J. W, Woods
‘ Metallurgy Division

- M. D'Amore
Pratt & Whitney Aircraft

The stress-rupture properties of the nickel-
molybdenum base alloys are being studied in
fused salts, sodium, and various gases. Six new
lever arm machines are being used to testHastelloy
B sheet specimens, and eight additional tube
burst units have Hastelloy B pipe in test. |t is
anticipated that design curves for properties in
the fused salts at 1500 and 1650°F will be available
by February 1955. Data are also being obtained
for a design curve for properties in argon. The
results obtained in argon at 1500°F are presented

in Table 7.4.

As Table 7.4 indicates, Hastelloy B has sub-
stantially greater strength at 1500°F than Inconel,
which will rupture in 1200 hr when stressed to
3500 psi. A Hastelloy B specimen tested in hydro-
gen at 12,000 psi ruptured in 1300 hr; o similar
time to rupture was found in argon. Thus there is
an indication that Hastelloy B may not be sensitive
to a hydrogen atmosphere when tested in the so-
lution-annealed condition.

TABLE 7.3. EFFECT OF AGING TIME ON ROOM-TEMPERATURE BEND
DUCTILITY OF WROUGHT HASTELLOY B

 

 

 

Tilme at
' 1500°F Bend Angle Before Fracture {deg)
(hr) | As-Received Specimen? Heat«Treated Spec_imenb
0 ~ Greater than 135,€ no fracture Greater than 135,l no friacfure_

25 - Greater than 135, fissuring evident Greater than 135, no fracture
50 92, complete fracture Greater than 135, no fracture
775 90, complete fracture Greater than 135, no fraci‘ufe
TOO | 84,> complete fracture Greater than .]35,_.no fracture

- 200 80, complete fracture 90, complete fracture |
223 76, complete fracture 82, complete i';rc‘;lcfure
300 | 80, éompiete fracture 75, complete fracture

 

Solution annealed at 2150°F and quenched.

bHeat treated in hydrogen for ]/2 hr at 2150°F and then for 30 hr at 1950°F followed by water quench.

_ “Maximum bend angle available with apparatus.

106

o4

-

 
 

 

 

"

n

FTI

o

TABLE 7,4, STRESS-RUPTURE PROPERTIES OF
HASTELLOY B IN ARGON AT 1500°F

 

 

 

 Stress Time to Rupture Elongation
{psi) (hr) (%)
15,000 157 45.0
13,500 180 15.0
12,000 1150 20.0
8,500 2200* 4.0
*Specimen still in test.

The new alloys being developed are also tested
for stress rupture at 1500°F in argon in order to
check the effect on strength which results from
the various modifications. The results of tests
made to date are presented in Table 7.5, These
modified alloys show a decrease in high-tempera-
ture strength and elongation in comparison with
Hastelloy B in the solution-annealed condition.

SPECIAL MATERIALS FABRICATION
H. Inouye J. H. Coobs
Mefallurgy Division

Duplex Tublng

The efforts to produce duplex seamless ’rubmg
that ‘will have good oxidation resistance on the
outer surface and good corrosion resistance on
the inner surface have continued. In the previous
report,3 tests were described in which hot-rolled

and hot-pressed composnes were deep drawn.
An alternaflve and more dlrecf method of producing

 

 

~ Sept. 10, 1954, ORNL-1771

3,

-

J. P Coobs and H. lnouyle,QaANP Quar. Prog. Rep,
p

 
 

  
  

 

PERIOD ENDING DECEMBER 10, 1954

tubing is now being used in which the composite
billets are extruded.

A study of the flow of metals during an impact
extrusion is being made, For the study, the effect
of the shape of the nose of the extrusion billet
was determined on extrusions of Zircaloy canned in
copper and extrusions of vanadium canned in steel.
In addition, a two-ply and a three-ply billet of
stainless steel—carbon steel were extruded, but
the extrusions have not yet been evaluated.

The results of the tests show that large-grained
billets, such as cast billets, produce a rough
interface between the outer cladding and the core.
A tapered nose on the can and @ square edge on
the core result in a short section of can material
restricted to the front of the billet; thereafter, the
core has a thin layer of the canning material on
the surface. A square nose on the can and a
tapered core result in a thin section of core begin-
ning very soon after the front of the extrusion and
increasing in thicknesses toward the end of the
extrusion,

Boroh Carbide Shielding

Suitable compositions for the boron carbide
shield for the ART heat exchanger are being de-
veloped by The Carborundum Company and by the
Norton Company by using various nonmetallic
bonding materials, including BN, SiC, and carbon,
as well as those mentioned previously.® All these
bonding materials have been used to some extent
in fabricating useful forms of B,C. Another pos-
sible source for fabrication of these shield pieces
is Sylvania Electric Products, Inc., which only

Vrecenfly became mteresfed in the prob!em.

In" a recent conference ‘with representatives of

The Carborundum Company it was stated that they

 

 

‘Elongation

 

7 Time foRupture R

 

 

 

(hr) (%)
gy TR AT g
8,000 510 "
60T 111
L A N 4 vt b wie

 

 

 
 

ks 1l

i

 

 

 

ANP QUARTERLY PROGRESS REPORT

have developed a suitable composition by using
SiC as the bonding material. However, specific
details and samples of the material have yet to be
received,

Tubular Fuel Elements

Twelve fuel plates have been prepared for forming
into tubes, and eight more are being prepared.
These tubes are to be drawn at Superior Tube Co.
by the plug-drawing technique with low (10%)
reductions. Other techniques such as hot drawing

~or hot quging may be used.

Control Rods

Th:rty-five control rods are belng prepared for
the GE-ANP project. They were received as

-specially straightened tubes, 0.504 in. ID, 0.625
‘in, OD, 35 in. long, with one end plugged. The
tubes are b'e'ihgwfi“e'd with a mixture of 50% alumi-
~num powder and 50% B 4C, both supplied by the
“General Electric Company. The B,C is a special

grade containing 81% boron. The mixture is pre-
pared with 2% paraffin as a binder and is being
packed tightly into the tubes with a pneumatic
hammer. Very small increments are used to obtain
a high density. Twenty tubes have been filled
with average boron concentrations of about 0.8
g/cm3, that is, somewhat in excess of the 0.7
g/cm® required. After filling is completed, the

. tubes will be heated and evacuated to remove

the paraffin and then cold swaged to final size.

Al-UO, Elements for Shielding Expetiment

Fuel plates 28 in. long and 21/4 or 21/2 in, wide
are being prepared for an experiment to determine
the shielding necessary for delayed neutrons.
These plates are to be mounted on a belt and
rotated at speeds up to 22 fps. The fuel concen-
tration of 0.25 g/cm® required for these plates is

quife high compared with that usually specified

for fuel plates, and therefore new fabrication
problems must be resclved.
" Several plates have been fabricated by using

.UOZ and aluminum powder cores roll clad with

aluminum.  Three plates having 0.040-in. core

~ thickness wnh 62 wt % UOQ, had the required fuel

'concenfrahon Both Hhigh- flred and steam-oxidized

UO2 were used with good results. The edges of

 the core were quite clean and straight, and the

U0, was distributed fairly uniformly, as shown by

radiograph examination.  Because of the high

108

percentage of UG, it was necessary to hot roll
the plates with the covers on at all stages to
prevent oxidation. Plates with lower percentages
of UO, may be protected with thin aluminum foil
pressed on the core compact during cold pressing.

The finished plates, as<clad with 25 aluminum,
were judged to be too weak for the high stresses
involved in the rapid cycling. Several more ele-
ments are being assembled by using 52§ aluminum
as frame and cladding in the hope that work harden-
ing obtained in finish cold rolling will yield plates
of sufficient strength.

BRAZING ALLOY DEVELOPMENT

" P. Patriarca K. W. Reber
R. E. Clausing | G. M Slaughter
Metqllurgy Division
J. M. Cisar
Aircraft Reactor Engineering Division

R. L. Heestand
Pratt & Whitney Aircraft

The resnsfance to oxidation of a number of high-
temperature brazing alloys when used in con-
junction with Inconel as a base material was
described in a previous report. This study has
been extended to include tests of several additional
brazing alloys. These tests have been conducted
in accordance with the experimental procedures
used in the previous studies. The results of
metallographic examination of inverted T-joint
specimens after exposure to static air at 1500°F
for periods of 200 and 500 hr are summarized in
Table 7.6, It appears that the majority of these
alloys are suitable for service within the limits of
this investigation. The periods of exposure cur-

rently being studied extend to 1000 hr. Tests are,

also being conducted at 1700°F in static air,

It is recognized that the test results obtained to
date may not be applicable except under isothermal
conditions. Therefore a duplicate set of experi-
ments will be conducted to determine the effect
of thermal cycling on the adherence of protective
oxide films. Included in these tests will be a
determination of the effect of the presence of water
vapor on the nature of the scale formation and the
resistance of this scale to spalling under con-
ditions characterized by temperature fluctuations.

 

4P. Patriarca et al., ANP Quar, Prog. Rep. June 10,
1954, ORNL-1729, p 94, '

P
PR

L peeile
™o

 
PERIOD ENDING DECEMBER 10, 1954

 

Y

 

 

 

 

 

 

"TABLE 7.6. OXIDATION RESISTANCE OF DRY-HYDROGEN-BRAZED T-JOINTS
T ' R -  Parent metal: Inconel o
< o " Average fillet throat: 0.015 in.
- Air temperature: 1500°F
: Brozing’  Oxidation in Statie Air
o o S at 1500°F*
T Brazing Alloy Composition (wt %) Temperature
' Lo e ' (°F) After 200 hr  After 500 hr
: - Commércicfl AI!§y§ '
 Coast Metals NP 50 Ni_12 §i~28 Fe4 Mo—4.5 P—1 Mn—0.5 Cr 2050 Slight Slight
=, Coast Metals 50 93 Ni—3.5 $i-2.5 B-1 Fe 2050 Slight Slight
Coast Metals 51~ 92 Ni-4.5 S|7—3 B-0.5 Fe 2050 Slight Slight
Coast Metals 53 81 Ni-4 Si—4 B-8 Cr-3 Fe 1950 Slight Slight
E*perimenml Nickel-Base Alloys
GEalloy62  69Ni_20Cro11Si 2150 Slight Slight
- - G-E olloy 81 ' 66 Ni=19 Cr—=10 Si—4 Fe-1 Mn 2150 Slight Slight
S Ni-Cr-Si _‘"73.5 Ni-10 Si-16.5 Cr 2150 Slight Slight
T NisSi 88 Ni-12'Si 2200 Slight Slight
5 . Ni-Mo-Ge ,,50 Nl-25 Mo-25 Ge 2150 . Slight Slight
: Ni-Mn-Cr. ‘ ' _ 35 N|—5§ Mnf-]O Cr - 2050 Severe Severe
- E#perifnefiful Pl-'ec‘iofis-M-eml-Buse Alloys
Pd- Nir . o 60 Pd—-40 NI - 2300 Vefy slight Very slight
* Pd-Ni-Si . 60 Pd=37 Ni-3 Si - 2150 Very slight  Very slight
~ Pd-Ge ) 7'90 Pd.-10 Ge 2050 Very slight  Slight
o 90 Au...m Co 1830 Slight

  

 

 

Au-Co

Very slight

 

 

*Very slight -'-”'I'esé than 1 mil of utfdck-} slight — 1 to 2 mils of attack; severe - greater than 5 mils of benefra'rion-

”"V“;omf was exposed to stohc air at 1500"F for a

T L

wFlg. 7.4,

S T DR IR

 

""perlod of 1300 hr prtor to mefa“ogrcphu exarm-

'-’"""“fcorrugated fins were formed from 0,010-in, -thlck"

3

nickel sheet and were brazed to a 7%-in.-OD Inconel

 tube with an 82% AU:-]S% Ni brazing alloy to form

 

The resistance of this brazing alloy to corrosion
by fluoride salts has been found to be satisfactory

109

n
s

IR

A

 

 
ANP QUARTERLY PROGRESS REPORT

i i

 

on the basis of short-time static tests. However,

an experiment was conducted to determine the :
influence of a high gold concentration gradient on *
the corrosion resistance of the internal surface ‘
of an Inconel tube when exposed to fused fluoride v

salt at 1500°F,

The test specimen consisted of a ]/2—in.-0D,
0.035-in.-wall Inconel tube formed into a seesaw
test capsule to which were brazed several nickel
fins of the dish type. An excessive amount of
brazing alloy was used, and the brazing time was <
extended to 1 hr at 1900°F to promote dilution
and diffusion. The capsule was then filled with
fluorides and operated as a seesaw test for a -

P _ period of 500 hr at 1500°F prior to examination.

Fig. 7.4. Corrugated Fin Segment of a Heat Since the observed 1- to 2-mil attack of the internal

Exchanger Brazed with an 82% Au~18% Ni Alloy. surface was considered to be normal, a microspark

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED ]

Az )
0.04 s -
Q.02
0.03 '
0.04 )
INCH _

. | Fi‘g‘)r."7.5.' Inconel T-Joint Brazed with 82% Au—18% Ni After Exposure to Static Air for 1300 hr at : ®

3 o 1500°F, Only minor attack can be seen. Etched with G. regia. 100X,

 

g

v
Srawnds

' ;}.m_{,
Fve
“qere

 

 
 

 

»

"

”»

R

spectrographic examination was conducted from
the braze alloy—Inconel tube interface inward to
the internal surface of the tube, Solid state dif-
fusion of gold was found to be hegllglble in that
it was not possible to detect any trace of gold at
the internal surface, and an abrupt change in the
gold concentration still existed at the braze
alloy—Inconel interface. '

It would appear that this braze alloy is satis-
factory for fused fluoride salt-to-cir heat ex-
changer fabrication within the limits of this study.
The use of this alloy in direct contact with fused
salts has also been considered, since the ease of
fabrication of components with moving parts, such
as pumps, would be improved.
expected that dissimilar metal mass transfer will
prevent this application, a thermal-convection
loop with several 82% Au-18% Ni braze joint

- inserts will be operated to experimentally determine
the effec'r.

BERYLLIUM OXIDE FUEL ELEMENTS

C.E. Cur’r:s L. M. Doney
- J. R. Johnson
Metallurgy Division

Bery“aum oxide ceramics are being produced
probably for the first time, by casting from a slip
which is basic. For successful slip casting, a
well-deflocculated slip is necessary, and hitherto
this has been accomplished through the use of an
acid, usually HCl. "In comparison, a basically

deflocculated slip is more easily and rapidly

Although it is -

PERIOD ENDING DECEMBER 10, 1954

prepared, does not attack the plaster of paris
molds, is not unpleasant to the operator, and
requires very small amounts of deflocculants.
As an exdmple, tubes 4 in. long, 0.25 in. in outside
diameter, with a 0.04-in, wall, have been slip cast
recently from beryllium oxide and also from be-
ryllium oxide containing 16.2 wt % of natural
uranium oxide. These tubes are to be used for
tests by the Advanced Design Section of GE-ANP.
To 83.8 g of BeO (previously milled to about 24
average particle size) plus 16.2 wt % of UD, were
added about 26 ml of distilled water and 4 ml of a
2% solution of sodium pyrophosphate, After 5 min
of stirring, a creamy slip with a pH of 8 fo 9 was
obtained which spread evenly over a glass stirring
rod and flowed off the rod in a smooth stream.
Casting of the tubes required a two-piece plaster
mold with an opening 5 in. long to allow for
shrinkage and trimming to 4 in. and a diameter of
0.29 in. to allow for shrinkage on the diameter.
The deflocculated slip required about 10 to 15 sec
to form a tube of the proper wall thickness in this
mold. The excess slip was drained off, and, after
an additional 3 to 5 min in the mold, the tubes
released readily from the mold and had sufficient
green and dry strength to permit handling.

A firing operation of T hr at 1700°C in a hydrogen
or argon atmosphere is necessary to produce the
density of 3.0to 3.4 tequired. In order to determine
how warpage in firing may be avoided, some tubes
are being fired suspended in a furnace, while
others will be fired lying on their sides.

 

.

 

11

]
i,
iy
Fromamih
3
-’

 

 

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

8. HEAT TRANSFER AND PHYS!CAL PROPERTIES

H F. F’oppendlek

Reactor Experimental Engineering Division

The heat transfer experimentation during the

- quarter included studies of ARE fuel in an Inconel
tube, the determination of performance character-

istics of an in-pile loop heat exchanger, studies of
velocny and temperature distributions for several

, "free—convechon volume-heat-source systems, and

prellmlnary determinations of fluid flow character-

“istics of a proposed ART core geomeiry. Several
" approximate mathematical temperature solutions for

forced-flow volume-heat-source entrance systems
were developed, and the question as to whether

‘electric currents which generate heat in the circu-

lating fluids of experimental volume-heat-source

“systems affect the fluid flow characteristics was
mveshga’red
- The enthalpies and heat capacities of NaF-ZrF ,-

UF4 (53-43-4 mole %) and LiF-KF- -UF, (48-48-4
mole %) were determined. The viscometry equip-
ment was modified for greater accuracy, and new
measurements were made for NaF-ZrF ,-UF ,. The
thermal conductivities of NaF-ZrF -UF4 and of
NaF-KF-LiF with and without UF, were measured.

FUSED SALT HEAT TRANSFER
 H. W. Hoffman J. Lones

"Reactor Experimental Engineering Division

" Preliminary heat transfer experiments have been
performed with the ARE fuel NaF-ZrF -UF , (53.5-
40-6.5 mole %) flowing in an Inconel tube. In
brief, the heat transfer system used consisted of
two small tanks (approximately /-ff capacity)
connected by an electrically heoted Y -in.-OD tube
with a length-to-diameter ratio of 43. Flow through
the heated section was produced by helium gas
pressure and controlled by an aqutomatic cycling

‘mechanism actuafed by the fluid level in the tanks.

The helium was purified by passage through a
NoK bubbler.

Operahon of the system was terminated after
]]5 hr in order to inspect the Inconel tube. The

‘average flmd temperatures during the period of

operation were 24 hr at 1200°F, 85 hr at 1300°F,

“and 6 hr ot 1400°F. Heat transfer measurements,

made in the heated section over the Reynolds

‘modulus range of 5,600 to 10,000, yielded results
‘which were 24% below the generally accepted

nmo

correlation for ordinary fluids. The p.recision of

the measurements was approximately 10%. Visual
examination of the test section after removal from
the system showed no apparent film. Samples of
the tube have been submitted for metallographic
examination. It is recalled that in the case of the
previous experiments for molten NaF-KF-LiF
flowing in Inconel tubes, 55% reductions in heat
transfer were measured and corrosion deposits
were found on the tube walls.

About 15 individual heat transfer measurements
were made between 24 and 115 hr of operation of
the NaF-ZrF,-UF -Inconel system. During this
period no reductions of the heat transfer coef-
ficients with time were observed. A new lnconel
test section is currently being installed for ob-
taining measurements in the period between 0 and

24 hr.,

IN-PILE LOOP HEAT EXCHANGER ANALYSIS
M. W. Rosenthal

Reactor Experimental Engineering Dw:swn

P. I. Perry A. D. Rossin
F. D. Miraldi
M.I.T. Practice School

The heat exchanger for the LITR fluoride fuel
loop was constructed in an unusual geometry to
fit in the limited space available in a beam hole.
In order to determine the operating characteristics
of this exchanger, an experimental investigation
of its fluid flow and heat transfer performance was
conducted. The heat exchanger for the loop is
formed with helical fins brazed around a section
of straight pipe. The fins are enclosed in a
closely fitting tube and form an annulus in which
air flows in a helical path. Molten salt passes
through the center tube.

A duplicate of the actual exchanger to be used
in the loop was connected to an air supply so that
air passed through the annulus just as it would
in in-pile operation. However, instead of a fused

salt, dry steam was admitted to the center tube
as the hot fluid. The air flow rate and inlet and

outlet temperatures were measured, as were the

steam pressure and the condensate rate. From this

W

e

 
 

W a

.

 

€1

"

 

e
=

{HEAT} REMOVAL :RATE, {tkw) /%

information fhe air How rate and the thermal re-
sistance of the air snde were obtained as functions
of the air pressure drop across the exchanger (the
steam-side resistance was negligible compared
with the air- -side res:stqnce) Combination of
these data with estimated coefficients for the
fused salt, after allowance for difference in oper-
ating temperatures, yielded performance curves for
the exchanger under anticipated in-pile operating
conditions. The data and a description of the
experiment are reported elsewhere.! The method
of conversion of the results of the experiment to
in- p!le conditions and the performonce curves for
the exchanger have been reported.? A curve of
the heat removc] rate as a function of pressure
drop of the in- plle loop heat exchanger is glven

in Flg 8 1
FREE CONVECTION IN FI-UIDS WITH VOLUME
- HEAT souncss -
D C Humllton ' F. E. Lynch

Reactor Expertmental Engmeermg Dlv15|on

The free- convecflon research, whlch had been

inachve for some tlme, hos been reochvated wuth

 

1P . Perry, A. D Rossin, and F. D. Miraldi, Mo-
mentum and Heat Transfer Studies o{ an In-Pile Loop
Heat Exchanger, KT-174 (to be issved

2M W. Rosenthal, Performance of ANP In-Pile Loop
Heat Exchanger, ORNL CF -54-9-83 (Sepf. 10, 1954).

ORNL—LR—DWG 4452

30.0

 
 
 
 
     

20.0

 

5.0

       
  
 
 
 
 
  

FUEL INLET TEMPERATURE, 1500°F
AR INLET. TEMPERATURE, 100°F

"Fig. 8.1

formance.

' l‘n-Pile Loop Heat Exchanger Per

 

1952,

Also, it hcs been proposed to measure' the over-

PERIOD ENDING DECEMBER 10, 1954

however, more limited objectives than before. In

the beginning of the ORNL-ANP effort, there was

‘no information on free convection in fluids with

volume heat sources. Since the temperature struc-
ture in the fuel pins of the first fused salt aircraft
reactor experiment could not be readily predicted
because of a lack of knowledge about the free-
convection heat transfer in such a system, a
free-convection research program was initiated.
The principal objective was to determine both
theoretically and experimentally the temperature
and velocity structure in a free-convection flat-
plate system and, from these data, to discover
basic postulates that would permit analytical pre-
diction of different systems which might be of
interest in future fuel element designs. This
project was well under way when the change to
the present circulating-fuel reactor design was
made. At that time the effort was greatly reduced,
and later this research was temporarily abandoned

" in favor of more pressing problems. A series of

memorandums summarizing the theoretical work
which had been accomplished were prepared.
Also, some preliminary experimental velocity
measurements in a flat-plate apparatus were ob-
tained. _

In the intervening years there has been a con-
tinued interest in the temperature structure in
fluids with volume heat sources that are losing
heat by free convection in applications such as

- dump tanks for liquid fuel reactors, radiation

damage capsules, and pins containing liquid fuel.
The repeated queries on this subject prompted
the decision to reactivate the research, but on a
less ambitious basis than before. Hence, a report’

was prepared recenf[y which provides « W|der'

distribution to the theoretical analyses which were
dlstrtbuted locolly as memorondums in ]95] ond

 

all radial temperuture dlfference (or heat transfer
coefficient) for a modified versmn of the flat-plate
system in which it will not be necessary fo

_measure‘flmd ’rempero{tures. Th o_rlgmal flat-plate
' odified and _is_almost reody'

  
 

for opercmon " The new opporctueconsmts es-

 

 

and L. D, Palmer, Free Convection in Fluids Having a
Volume Heat Source, ORNL-1769 (Nov. 15, 1954).

117 113

 

. senfially of three parallel flat copper plates 8 in.
“wide, 36 in. high, and separated by 1/ -in. gops.

3D C. Homllfon, H F. Poppend:ek R F. Redmond

s o8

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

~Plastic bottom and end sections make leak-proof
containers of the two channels which are filled
with a dilute electrolyte (HCl in H ,0).  VWith the
two outer plates grounded and the center plate
maintained at a given a-c voltage, heat is gener-
ated within the fluid volume in each channel.
‘Coolant passages ‘are attached to the outside of
‘the outer plates which uniformly remove the heat
generated within the free-convecting fluid. Small

 holes drilled vertically down each of the three

* ‘plates for the entire 36 in. permit insertion of wall

: :-'_;;.t_hgrchoup]e probes for measuring the wall temper-

atures at any level. The variables to be measured
‘gre coolant flow rate and temperature rise, power
input to the appdrotus, and the three wall temper-
atures at vcrlous Ievels.

HEAT TRANSFER EFFECTIVEN ESS OF
' “REACTOR COOLANTS
’ M W Rosenfhcl R. M. Burnett
' H. F. Poppendiek
Reacfor Experlmen’ml Engineering Division

 

One of the considerations in the selection of a
qoo]anf for a nuclear reactor or any heat exchange
system is the amount of flow work required for
removal of a unit of heat. This quantity, which
is referred to as the ‘‘cooling work modulus,”
varies with the geometry of the system, the fluid
velocity (and flow regime), and the properties of
the fluid.

A menid_rqrjdurh is being prepared that gives the
development of equations for calculation of the
cooling work modulus for turbulent flow. Curves
which simplify computation are provided, and the
results of a study of a specific reactor are pre-
sented for illustration. This study has been
performed for an idealized reactor coolant system
which is considered to be somewhat representative
of actual solid-fuel-element reactors. In this
unalys:s the cooling work modulus is related to
the reactor geometry, the heat removal rate, the
thermal properhes of the coolant, and certain
coolant temperature differences which are de-
-'rermmed by such factors as mean coolant temper-
'afure, coo[qn’r melting and boiling temperatures,
'V_fube wall corrosmn and mass transfer, and thermal
siresses.

e All the i‘hermal properties of a coolant are found
to mfluence its effechveness in transferring heat
'~ from fhe reactor. An ideal coolant is indicated

“to be one that has high thermal conductivity, heat

14

capacity, and density and a low viscosity, but the
relative merits of actual fluids depend strongly
on the temperature differences which are taken
as limiting in the system.

ART CORE HYDRODYNAMICS

J. O. Bradfute L. D. Palmer
F. E. Lynch G. M. Winn
Reactor Experimental Engineering Division
G. L. Muller |

Pratt & Whitney Alrcraf'r

The experimental flow sysfem for studymg the
hydrodynamic characteristics of proposed ART
cores was completed, and several photographs of
particles and grids were made which yielded much
information about the apparatus but no quantitative
data. Some qualitative evidence for the existence
of two, and possibly three, regions of flow sepa-
ration in the ART core currently being tested
was obtained by visual observation. Additional
qualitative observations are continuing which are
soon to be supplemented with quantitative dataq,
but these preliminary observations indicate that
the core with ideal entrance conditions must be
considered to be unacceptable and that it will
be necessary to use vanes at the entrance.

It was found that sufficient light is available
to produce photographs of particles flowing with
the water at a Reynolds number of 100,000, al-
though at such high velocities the flash duration
is a little long. It is believed that this can be
remedied with a different flash tube, which is
being procured. In addition, these first photo-
graphs revealed the need for an additional light
source arranged to illuminate the particles con-
tinuously so that their tracks would represent a

line on the photograph. This is necessary to

reduce the uncertainty associated with identifying
the two images of each particle. A 1000-w incan-
descent projection bulb proved to be inadequate,
and arrangements are being made to test two
different mercury-vapor light sources; they will
increase the light intensity by a factor of 10 to
20, which should resolve the illumination problem.

Several details such as the quality of the grid
photographs, the settling velocity of the particles,
and the reliability of the Hewlett & Packard Co.
electric counter have received at’ren‘rlon. A new
grid was made with much finer, shallower lines
than the first. The photographs of the new grid
were far superior to those of the old one, since

y

fl e TR o
5.‘ o i s
> : .

«0 K Bt

T,
L s

s

o
e N ot .

 

 

W

B o0

"

¥

i

a much smaller area was obscured by the grid
lines themselves. Inherent difficulties have beset
attempts to measure the settling velocity of the
particles; however, it can be said that the terminal
velocity of free fall of Lycopodium spores in water
under gravity is certainly less than 0.2 cm/sec
and probably much less. Even this is only about
0.16% of the mean velocity which will be measured,
The Hewlett-Packard counter which measures the
time interval between flashes was adjusted; its
absolute accuracy was checked; and it was found
to be satisfactory. '

The design of another entrance section which
will |mpcrt a rotational component of velocity to
the fluid as it flows through the test section has
been started. This will produce a velocity struc-
ture which is more realistic in terms of what is
envisaged for the ART, but it is much more
difficult to measure. It is conceivable that a
rotational compbhen‘f will help to réduce the
'possfl)le adverse thermal effect of separation
regions, perhaps by eliminating these regions.

. The velocity distribution in a diverging channel
under turbulent flow conditions was studied with
the phosphore$centéparfic|e technique.  Flow
asymmetry at a half-angle of 4 deg was observed.
This angle is somewhat less than the 5-deg half-
angle given in the classical report on velocity
profiles in channels by Nikuradse.? A large flow-
visualization system that utilizes the phospho-
rescent particle and dye techniques for studying
fluid flow characteristics has nearly been com-
pleted. -

 
 
  
   
   
 
    

sxs descnlfi)e_d "rhe“

BAET IR AT

‘strdcture ind samp]e forced flow volume heat-
“golirce “system,’

5 Vconducted “on -‘the electrlc':ro! mefhod, oi?&.generchng

 

 

1-49

. Nikuradse, Forschungsarbeiten VDI 289,

 

. sandw:ches

 

fiacnd solutlon wn‘h qnd wn‘hout elecfr

PERIOD ENDING DECEMBER 10, 1954

volume heat sources within the fluids of experi-
mental forced-flow volume-heat-source systems.

An experimental study of the factors affecting
gas formation at an electrode-electrolyte interface
was carried out pursuant to the problem cf gener-
ating volume heat sources electrically. The
generation of gas at an electrode was found to
be dependent upon the material of the electrode
and the current density and independent of the
applied potential, as well as the type of electro-
lyte employed. Platinum was found to have the
most desirable characteristic of conducting the
maximum amount of current per unit area of elec-
trode to an electrolyte for a given degree of gas
generation.  This effect is most probably the
result of an increased surface area because of
the known porosity of this metal. For brightly
polished platinum, a current density of approxi-
mately 8 amp/in.? was achieved before gas liber-
ation was noticeable. This value exceeded that
of Carpenter 20 stainless steel by two orders of
magnitude. Also, tests were carried out to study
the compatibility of the more common stainless
steels and dilute solutions of strong electrolytes
in those portions of an experimental volume-heat-
source system where no electric currents are
flowing. Carpenter 20, type 347, ond type 316
stainless steels were found to be accordant.

Concomitant with the problem of generotmg
uniform volume heat sources within annuli of
nonuniform cross section, a method for embedding
platinum electrodes to form a hydrodynamically
smooth surface has been developed. It has been
established that several of the new epoxy bonding

..~ plastics adhere to suitably roughened metals in

such a manner that the casting of ‘composite
of metol _cmd plastic is feasible.
Three such expenmenfcl cushngs were fobrlcated

  
 

!ong p[ashc tubes which were ductmg a' su!furlc

AL

  

7 ‘The results, expressed in terms
of the friction factor and Reynolds modulus in
Fig. 8.2, indicated that there was no detectable

m TN r
oy ),,‘ _, v n; ‘rs,
el 4 115

 

(rents, o

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

0.2 —

. UNCLASSIFIED
ORNL-LR—DWG 4453

OPEN POINTS INDICATE NC CURRENT FLOW
SOLID POINTS INDICATE CURRENT FLOW IN FLUID

0.05

{, FRICTION FACTOR .

LAMINAR FLOW
g 64
Re

" Goe

Q.01

 

LENGTH {in.)

20
83
155
{99

BULENT FLOW
_0.184
g - Re0-2

10? 2 5 10°

Re, REYNOLDS NUMBER

Fié. 8.2, Friction Factors With and Without Electric Cutrent Flow in Fluid.

difference in pressure drop through the tubes under
the two different conditions and that, hence, the
hydrodynamic structure was not influenced by the
electric currents for the laminar and turbulent
ranges investigated.

PHYSICAL PROPERTIES MEASUREMENTS

T HeatCupacfl'y
W, D. Powers G. C. Blalock

Reoctor Exper:mental Engineering Division

,7 - The em‘halpies and heat capacities of NaF- ZtF -
V_UF4 (53-43-4 mole %) were determined with ’rhe

S ‘_Bunsen ice calorimeter in the liquid and solid
. states ond wnh the copper-block calorimeter in
- ,the hqund state The results are as follows:

Sohd (70 to 525°C) Bunsen calorlm_efer

__ H = 0.18(2)t

0.18 = 0.01

fi
i

16

Liquid (550 to 900°C): Bunsen calorimeter
Ht - HOOC = 32 + 0.25¢ '
Cp = 0.25 + 0.02

Liquid (570 to 885°C): copper-block calorimeter
H, ~ Hpo = 19.5(0) + 0.265(6)
<, 0.265(6) + 0.002

The enthalpies of this salt have also been de-

termined by the National Bureau of Standards, and

a comparison is made in Table 8.1 of the results,

The enthalpies and heat capacities of L:F KF-
UF, (48-48-4 mole %) were determined by the
c0pper -block calorimeter in the solid and liquid
states.

Solid (125 to 465°C)
H, - Hyo C_0234t+048>< 10-4 2
= 0.234 + 0.95 x 10~4

lLiquid (565 to 880°C)
H, - Hgpo s = ~82.54 + 0.657: - 1. 97 x 10=4 42
= 0.657 - 3.93 x 10™*

.J
(AR
m}g
Sorendh
LD
Iy

4
 

 

%y

X oy

iy

   
  
 
 
  
   
  
 
 
  
   

i

Ay

9

L

PERIOD ENDING DECEMBER 10, 1954

TABLE 8.1. COMPARISON OF ENTHALPY DETERMINATIONS AT ORNL AND NBS FOR
NuF-ZrF -UF (53-43-4 mole %)

 

Enrfh‘alpy (cal/g)

 

Temperature

 

 

. ORNL

("C) NBS
? Bunsen Copper Block

600 182 178.9 178.9

700 207 205.4 205.3

800 ‘ 232 232.0 231.1

900 257 258.5 256.0

 

In these expressions H is the enthalpy in cal/g,
c, is the heat capacity in cal/g-°C, and ¢ is the
temperature in °C.

| Viscosity
S.‘-I. Cohen T. N. Jones

Reac'ror‘Experimen'ral Engineering Division

‘The program of viscometry technique refinement
previously discussed® is under way, and measure-
ments have been made on one mixture. Studies
will be made on others as samples become avail-
able.

A number of changes have been made in the
viscosity apparatus. The length of the Brookfield
spindle has been doubled, which allows the speed
of rotation of the instrument to be halved without
reducing the shear force. This reduction in speed
assures the existence of laminar flow. Salts are
contained in new tubes with a smaller diameter

and consequenfiy a smaller clearonce between the

 

55. I. Cohen, ANP Quar. Prog. Rep. Sept. 10, 1954,
CORNL-1771, p 127.

 

handling of the instruments and improve atmos-
pheric control. New calibrating liquids which have
densities comparable to the densities of the
fluorides are being used for both viscometers.
Among these liquids are low-melting-point, easily
handled, fused salt mixtures and heavy organic-
halogen compounds.

Recent measurements made on NaF-ZrF,-UF,
(53.5-40-6.5 mole %) are presented in Fig. 8.4.
It may be noted that the average values obtained
from the recent measurements are about 30% lower
than the preliminary values given previously.’
However, when the preliminary measurements were
reported, only the more pessimistic, Brookfield
values were given. It can also be seen from
Fig. 8.4 that the agreement between the two instru-
ments for the recent measurements is much better
than for the earlier set of measurements. This
result is due to improved fluoride preparation and
handling techniques, as well as to the refined

viscometry methods.

Thermol Conduchwty

W D Powers R. M. Burnett
S. J. Claiborne

Reactor Experimen’ral Engmeenng Duvnsuon

The\ ’rhermai conduc'rsvn‘y “of NaF- ZrF -UF
(53.5-40- 6 5 mole %) in the hqu:d ‘state was found

f° be 1.2 BfU/hr-ff~°F at an average sample temper-
o "afure of_666°C The varmble -gap devnce was Used

T_he conduchvr’ry of

) V‘T-i‘hls salt ‘mixture had’ prewously been estimated
box to be used solely for wscosn‘y work is bemg"“ “fo be 1.3 Btu/hr-#+-°F on the basis of an empirical

Cbuilt, This dry box should further facilitafe

 

65.. {. Cohen and T. N, lJones, Preliminary Measure-
ments of the Viscosity of Fluoride Mixture No. 44,
ORNL CF-53-8-217 {Aug, 31, 1953).

117

L}

_
oy
o

o

Ty}

Fenh

 

 
i

 

 

 

ANP QUARTERLY PROGRESS REPORT

STAINLESS STEEL MELT TUBE
FABRICATED FROM SCH 40 1-in.-IPS PIPE

    

 

 

  

 

CHROMEL-ALUMEL
THERMOCOUPLES (4)

UNCLASSIFIED
ORNL-LR-DWG 4454

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

BROWN
MULTIPOINT
RECORDER
7
. =3 SIMPLITROL
COPPER LINER BRONZED / :
WITH ALUMINUM ——— | . : ,
. 71 RELAY
o N
FURNACE ELEMENT —0 | / 4P VARIAC [ 110y
. | % 10
/] 1o
R . ':g CHROMEL-ALUMEL
]NSULAT|ON // g O CONTROLLING COUPLES
it = e
// 3‘ ; O
% A Ok
/ T | SIMPLITROL
A e
Z o J
- o POSITION OF THERMOCOUPLE
Z / RELAY WELLS IN MELT TUBE WALL
? A% (CHROMEL -ALUMEL COUPLES
v /,O IN 0.08-in.- OD STAINLESS
. v S STEEL WELLS)
/ A5 VARIAGC HO v
/ g
®
o

 

 

 

 

Fig. 8.3. Furnace for Yiscosity Measurements.

conductivity relation developed for the fluoride
mixtures previously studied.

The solid and liquid thermal conductivity meas-
urements of NaF-KF-LiF (11.5-42-46.5 mole %) and
I\lcl"'-KF«LiF--UF4 (10.9-43.5-44.5-1.1 mole %) were

reported in part in previous quarterlies. All con-

_ducrtivirty ‘data obtained with several types of
““conductivity devices for these two fluoride mix-
‘tures are shown in Fig. 8.5, The liquid and solid
~ conductivity values of the ternary fluoride mixture

are nearly the same, and the same is true of the

‘quaternary.

118

A longitudinal thermal conductivity apparatus
has been designed and constructed for the meas-
urement of the conductivity of metallic beryllium,
It is essentially the same apparatus as that re-
ported previously’ and is contained in an inert-
gas-filled box to protect personnel from beryllium
poisoning.

 

7W. D. Powers, S. J. Claiborne, and R. M, Bumett,
ANP Quar. Prog. Rep. June 10, 1953, ORNL-1556, p 86.

Pl T37F 9 ™
.-if’,g::ié x_ff:)

s P S,

#w

i
 

Ny,

-

Sy

“

[y

=

Ay

%,

3

 

SEGRE
ORNL-LR-DWG 4455

TEMPERATURE {°K)

600 700 800 1000 1200
100 ‘

| CUP | BROOKFIELD | AVERAGE

50 PRELIMINARY MEASUREMENT

MADE AUG. 1853
MEASUREMENT MADE
AUG. 1954

o a

—_—————

* A

& (CENTIPOISES)
- ™)
o C

o

 

300 400 500 700 1000
TEMPERATURE (°C)

Frig. 8.4. Vi'scos'ify of NaF-ZrF“-UflF.; (53.5-

40"6.5 m°|e %)o

 

PERIOD ENDING DECEMBER 10, 1954

 

 

 

 

 

 

 

 

 

O
ORNL—-LR—DWG 4456
4.00
so1_1o——4—'———- LIQUID —— ]
o
o 300 ™
£ & .
. A
= o
< A u e
2 .o
m
- ™~ = —— O v
S
5 N \
> "0 . _ 5
£ 2.00 A ~o <3
o ~ ~
2 ~
2 ¥
Z ~
o
-
a
Z
L:EJ 1.00— METHOD NoF -KF-LiF NcF—KF—LiF—UF,;
= ' {11.5-42-46.5 mole %) (10.9-43.5-44.5~11 mole %)
- COOLING SPHERE A A
(S0LID)
FLAT PLATE n o
(SOLID)
VARIABLE GAP » o
{LIQUID} - I |
o r
0 500 1000 1500

TEMPERATURE (°F}

_' Flg. 8.5. Th-e‘rlfiél;Cofi;h;c-ti‘v'i.fies of NaF-KF-LiF
and NaF-KF-LiF-UF, in the Liquid and Solid

States.

 

Fo

2 119

:'?lvncf‘s

 

 

 

 

 

sy

 
 

 

 

 

ANP QUARTERLY PROGRESS REPORT

9. RADIATION DAMAGE

J. B. Trice
Solid State Division

Ac Jo Mi”ef
ANP Project

- The program of MTR irradiations of Inconel

capsules containing fluoride fuels has continued,
and the capsules examined thus far have revealed
no evidence of radiation damage. The horizontal
type of fuel-circulating loop designed for irradiation
in the LITR has been operated out-of-pile with a
non-uranium-bearing salt and is now being inserted
in the HB-2 facility of the LITR, A small loop suit-
able for vertical operation in the LITR lattice has
also been successfully bench tested, and a second
mode] for in-pile operation is being constructed.

The creep test apparatus for testing Inconel at
high temperatures in the MTR is being bench
tested. The stress-corrosion apparatus for LITR
“operation has been successfully bench tested, and
an in-pile apparatus is being constructed,

Remote metallographic studies of solid fuel
elements were continued, and additional information
on the relationship between UO, particle size and
radiation damage was obtained.

MTR STATIC CORROSION TESTS

W. E. Browning G. W. Keilholtz
Solid State Division

Ha Lr Hemphi”
Analytical Chemistry Division

The program of MTR irradiations of Inconel
capsules containing fused fluoride fuels has been
continued. Additional irradiations were carried out
on capsules containing NaF-ZrF ,-UF, (48.9-49.3-
1.79 mole %) and on capsules containing NaF-
ZrF,-UF, (50.1-48.2-1.74 mole %) These fuels
generate 1100 w/cm® in the A-38 position in the
MTR. To date, five capsules in this series, three
with UF, and two with UF, fuels, have been
" successfully irradiated for a two-week period with
metal-liquid interface temperatures of 1500 £ 50°F,
The capsules have been examined metallographi-
cally, and the fuels have been chemically analyzed.
One out-of-pile electrically heated control capsule
containing UF, and one containing UF, have also
been examined, and the fuel batches have been
chemically analyzed.

120

The irradiated UF3-bearing CO.pSUIé‘.S and one of

the irradiated UF ; capsules showed practically no
corrosion — that is, they were similar to the control
capsules — while one irradiated UF ,-bearing cap-
sule, previously reported,! had subsurface voids
to a depth of 2 mils. There were no significant dif-
ferences in the iron, chromium, or nickel contents
of the irradiated fuels, as compared with the starting
fuel batches, and there was no evidence of segrega-
tion of either uranium or impurities. The chemical
analyses of the two out-of-pile controls that have
been examined to date were accidentally spoiled,
The uranium in the UF -bearing fuel analyzed not
less than 96% ,UF3, and the UF4-beuring fuel

showed no trivalent vranium. In an attempt to de-

tect any minute radiation effects which might be
occurring, the irradiation period is being extended
to six weeks for the other capsules in this series,
Additional examinations were made of capsules
containing NaF-ZrF4-UF4 (50-46-4 mole %) that
had been irradiated at a nominal temperature of
1620°F and had generated 2700 w/cm®. In this
series of irradiations, the in-pile temperature
history was quite complex. No evidence of chemi-
cal damage could be found in the fuel, and there
were no high concentrations of Inconel components.

There was some corrosion evident, and therefore

out-of-pile control tests will be made in which the
in-pile temperature patterns will be duplicated
insofar as possible. New experiments with this
fuel are to be made in which more closely controlied
temperatures and planned temperature excursions
will be used,

The chemical analyses of samples of fuel taken
from irradiated Inconel capsules have at times
shown increases in iron such that the iron-to-
chromium ratio was greater than that found in the
Inconel, and therefore a series of radioactivation
analyses of the iron content in the irradiated fuel
was made; these analyses demonstrated that the
high 'values were due to contamination of the fuel

 

w. E. Browning and G. W. Keilholtz, ANP Quar.
Prog. Rep. Sept. 10, 1954, ORNL-1771, p 134,

i
 

ot

ot

0y

with unirradiated iron that had entered the fuel
during or after opening of the Inconel capsule. The
techniques developed by G. Smith (Analytical
Chemistry Division) are particularly adaptable to
the separation of minute amounts of iron (10 to 100
pg) from very large amounts of fission products.
Four samples of fuel from irradiated capsules
showed that between 70 and 99% of the iron came
from an unirradiated source, whereas samples from
an irradiated Inconel copsule showed that less than
6% of its iron content was not radioactive, Since
the amount of iron contamination involved is very
small (of the order of 50 ug), some possible sources
of contamination would include the steel drill used
to remove the fuel from the capsules and also the
lmpurlhes in the reagenfs used in preparing sample
solutions. In the future, samples will be obtained
by using a nonferrous drill. Unfortunately, radio-
activation techniques cannot be applied to analyses
for chromium and ‘nickel because of unfavorable
activation yields, insensitivity of detection, and
chemical separation difficulties associated with

trace amounts of 'rhese elements.

An improved and modified version of the capsule
irradiation facility is being put into service (Fig.
9.1). Before being shipped from ORNL, each
capsule is mounted in a sleeve which provides a
sufficiently large annulus for the passage of cooling
air between it and the capsule. The bottom of the

AIR PATH

 

[NSTALLATION OF IRRADIATION TUBES
ToTAN MTR REFI_ECTOR "A" PIECE

PERIOD ENDING DECEMBER 10, 1954

sleeve is tapered to fit an inside taper at the
bottom of the air tube in the MTR. The advantage
of this arrangement is that the capsule can be
fitted to the sleeve prior to insertion into the
reactor, Several tubes and capsules prepared with
this new arrangement have been used satisfactorily
in the MTR. In order to speed up the capsule
program, arrangements have been made for irradiating
two capsules simultaneously in separate, but
adjacent, facilities in the MTR.

An improved set of temperature controls fabricated
by the Instrumentation and Controls Division has
been subjected to dynamic performance fests by
using an electrically heated capsule. These tests
have shown that the new controls will maintain a
more constant temperature in the capsule during
irradiation,

EFFECT OF IRRADIATION ON UF ,.C,F,,

W. E. Browning - G. W. Keilholiz
Solid State Division

Exommahons were made of material from welded
nickel capsules that were filled under vacuum with
a UF -C,F, ( solution containing 20 wt % UF , and
then irradiated in the ORNL Graphite Reactor and
in the Tower Shielding Facility. These irradiations
were made to determine the suitability of UF -
C,F,¢ for use as fuel in a Lid Tank Shielding

Facility mockup of the circulating-fuel reactor,

- epeme
ORNL-LR-DWG 4603

 

 
 

' %\\\\\\\

X'iwrmcm_ THERMOCOUPL7 /
. iCAPSULE ,

 

 

  

 

ELLLLLI SIS IL SIS ///11////1////

IRRADIATION TUBE WITH CAPSULE INSTALLED

ANNULUS SLEEVE

7 T T r

 

 

\\\\\\\\\\\\\\\
YL IIII/)II}II/IIIIII L7 PP LA T >I\>>\

 

fRRADlATION TUBE

IRRADIATION TUBE WITH CAPSULE REMOVED

Fig. 9.1. New Apparatus for Insertion of Capsu|és in the MTR.

121

 

 

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

Three of the capsules were exposed to 1.1 x 10717
-"nvt'f'h.‘ Gray-green solid residues which filled as
much as 90% of the volume were found in all the
“capsules. The residues in the capsules exposed

. to a flux of 1.1 x 10'7 analyzed 92% uranium, and

~the residues in those exposed to a flux of 2.0 x
, ]0]5 analyzed 24% uranium. In the capsules given
' fhe higher exposure, gas pressures of between 55
‘and 100 psi developed. Analyses of these gases
~showed about 30% CF, and about 25% C,F; the

* remaining gas was unidentified,

2" ¢
' - MIN}ATUR‘E IN-PILE LOOP
"~ Bench Test

W, R.'Willi_s M. F. Osborne
H. E. Robertson G. W, Keitholtz
v Sd‘lird___State Division

The first successful bench test of the miniature
in-pile loop designed for insertion in a vertical
__hole in the LITR included four freezing and melting
cycles in 260 hr of operation. The loop was oper-
ated at 1466°F, The linear velocity of the fused
salt circulated in the loop was between 3.3 fps
(as calculated from the flowmeter) and 3.8 fps (as
calculated from the pump speed); thus a Reynolds
number of about 3000 was obtained, Since it was
possible to freeze and melt the fused salt without
causing failure of the loop, it may be advisable to
fill the in-pile loop before it is inserted in the
reactor and then melt the fuel mixture after the
loop is in position. Since, during the bench test,
the flowmeter was found to be temperature sensitive,
the dependence of the measurement on the temper-
ature must be established.’

The bench test was terminated because of a leak
in a collar that was welded over a joint in the fuel
tube. The two ends of the loop did not butt together

“at this location, and the annular space thus formed
between the collar and the tube sections trapped
the fused salt in such a manner that expansion of
the salt caused the collar (not the weld) to rupture.

“The metal collar stretched from 0.4 in, in diameter

to 0.5 in. before it ruptured. The salt from the
" leak, which existed during operation for about 4
“hr, showed a tendency to oxidize and to stick to

- the Inconel surface rather than to flow rapidly down
~‘the outside tube wall. Thus, a leak in the colder
_portion of a loop that was operating in-pile would

~“ " probably not flow downward into the high-flux

region before a safety alarm from the released
“radioactivity could scram the reactor,

A Delco motor from the group being used in bench
tests has been rebuilt to withstand radiation and
a higher temperature, The wire on both the rotor
and the stator was replaced with glass-insulated
wire. All paper was removed, and mica was used
in the commutator and in the rotor segments; glass
cloth was used, where possible, for other insula-
tion. These changes required that the shaft as-
sembly be slightly modified to allow more space
for windings. The rebuilt motor was tested during

the bench test of the miniature loop, and it was found

to be satisfactory. In a comparison of the operation
of the rebuilt motor with that of the original motor
under no load and under load cond i-fio'ns,' the rebuilt
motor was found to be slightly more efficient than
the original motor. Furthermore, since the rebuilt
motor can withstand a higher ambient temperature
and since for a given voltage it produces a higher
speed, higher fuel velocities can be obtained than
were possible previously,

Heat Transfer Calculations

M. T. Robinson
Solid State Division

E. R. Mann F. P. Green
R. S. Stone

Instrumentation and Controls Division
D. F. Weekes, Consultant

An extensive series of heat transfer calculations
has been carried out on the ORNL. Reactor Controls
Computer in order to predict the thermal behavior
of a miniature in-pile circulating-fuel loop. The
derivations of appropriate differential equations
and the details of their solution are given in a
forthcoming report,? The model finally adopted for
the calculations and for the in-pile loop is shown in
Fig. 9.2. It was assumed to be mounted in position
C-48 of the LITR, with the lower end at the location
of maximum thermal-neutron flux.® The maximum
power density in the fuel in position C-48 was
assumed to be 540 w/cm?3.

The computer results were obtained for a variety

of different flow rates of fuel (NaF-ZrF -UF,,

53.5-40-6.5 mole %) and of cooling air expressed
as Reynolds number of the cooling air stream, Re ,
or fuel stream, Ref. The heat transfer, y, that is,

 

2\. T. Robinson and D. F. Weekes, Design Calcula-
tion for Miniature Hifib Temperature In-Pile Circulating
Fuel Loop, ORNL-1808 {in press).

M. T. Robinson, Solid State Semiann. Prog. Rep.
Feb. 28, 1954, ORNL-1677, p 27.

(e

n
 

L1

W

gy

r 'i{'4

 

the amount of heat flowing radially through a unit
length of the loop, vs the distance, s, from the
inlet is shown in Fig. 9. 3 and the devmtlon, .

" UNCLASSIFIED
- .S5D-B-988
ORNL-LR-DWG 22524

FUEL INLET

FUEL OUTLET '

  
 

AIR INLET

   
  
  

 

 

 

 

 

 

 

 

i [
ol
/ 1
i
FUEL LOOP ’ i 1
¢ \
]
AIR COOLING ’ * ’l ’*
ANNULUS g 1N ”‘

 

 

 

 

' AIR OUTLET

F lg. -- 9.2.

‘Miniature In-Pile 'Clrculuhng-Fuel“
_Loop Model Used for Desngn Culculahon .

PERIOD ENDING DECEMBER 10, 1954

of the mixed-mean fuel temperature from its average
value vs s is shown in Fig. 9.4. These results
meet only. one of the three required boundary con-
ditions, namely, that all fission heat be removed
into the cooling air stream. The other conditions,
that the maximum fuel temperature be 815°C and

~ that the initial air temperature be 30°C, are met

by suitable interpolation of the calculated data,
Some of the results of the calculations are summa-

rized in Table 9.1.

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

SESFEF
ORNL—LR—-DWG 37824
60
55 ‘ : A
Re, /
k 10,000 _ /]
0\ 20,000 — ' 7
N 30,000 2
€ 45 AN ) 4/
S \ 60,000 ————— 7,/
z 3 ' |
L 40 N\ Vs
5 4
w) .
=
g J
= ' : .
: Y
T 30 Af—
" \\ / /
25 N / /'
| \\\\\J/ /|
20 . A
15 :
O 20 40 60 80 1025 125 145 165 185 205
s, DISTANCE FROM INLET (cm)
Fig. 9.3. Heut Transfer in Miniature In-Plle

-7;1:00p for [ Fuel ReY"°|d5 Number of 3000 as @

_F‘uncflon of Dlstunce from fhe lnlet und the '

 

 

’;leferenhol
~Cay

  
   
 

 
   

Temperature T
R Air Flow
© (scfm)

’ Te mperat ure

{OC)_ -

 

 

 

) “ 6 000 D 15,5b0 R A S e T e 25 ~5;IOV e T R T 20 ST T
10,000 15,500 25 550 20
123

 

 

 

 
ANP QUARTERLY PROGRESS REPORT

SEereT
ORNL-LR—DWG 3783A
58

50
45 \

40

 

 

 

 

35

 

 

 

 

 

 

 

 

 

 

 

 

 

 

20 /

o 25 | / . \

3 ' : I // \\\

4 2 6000 j X

Re, = ] M

. 7 {40,000 J’l \ \

; _ o 15 II \
w - \
5 10 'I / \\h \‘ \
<t I \
7 /7 \
z 5 1/ \\\ \
- i/
i _ ] 6000
W F Rey {40,000 Ll/ \3
L0 \
N \\</ \

\ /f

 

 

 

—{0
| ~_7 1’/
15 \.\ il
RuN
/

 

 

Ref-——-‘|500
—-25

LN
N

o 20 40 60 80 1025 425 145 4865 {85 205
5, DISTANCE FROM INLET {cm)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

7 AFig‘.‘9.4. Fuel Temperature in Miniature In-Pile

. Loop for a Cooling Air Reynolds Number of 20,000

as a Function of Distance from the Inlet and the
Reynolds Number for the Fuel.

 

 

124

 

REMOVAL OF Xe 133 FROM
MOLTEN FLUORIDE FUELS

M. T. Robinson
Solid State Division

W. A. Brooksbank
Analytical Chemistry Division

D. E. Guss
USAF

The study, described previously,? for determining
whether Xe‘]35 would escape from ART-type purging
equipment rapidly enough to prevent serious poi-
soning was continued., It was realized at the time
this study was initiated that the ARE might answer
the question satisfactorily, but it was decided to
go ahead with experiments to measure the xenon
solubility under the more pertinent ART conditions.
It was considered unlikely that 97% of the equi-
librium xenon would be removed from the ARE with-
out special purging equipment, and, yet, this is
what happened according to two independent sets
of calculations. During the initial stages of
reactor operation, data were obtained from a re-
activity calibration of the ARE regulating rod by
noting changes in rod position required to maintain
criticality as increments of U235 were added to the
fuel. The value of the rod was calculated to be
0.030 +0.006% Ak/in., which agrees well with the
preliminary value reported by J. L, Meem (cf. Sec.
1, “‘Circulating-Fuel Aircraft Reactor Experiment’’).

When the ARE was operated for 25 hr at an esti-
mated 2 Mw, constant power was maintained by
adjustment of the regulating rod. At the end of
this period of operation, the rod had been withdrawn
an amount equivalent to a reactivity decrease of
0.015 £ 0.003%. The expected decrease in reachvnfy
was calculated to be:

From burnup of U%33 0.001%
From Sm'4? poisoning 0.008%
From Xe'35 poisoning 0.41%

Since over one-half the observed change in reactiv-
ity may be accounted for by burnup and Sm'4?
poisoning, the amount of Xe!3® retained in the fuel
was less than 1 to 2% of the amount expected.

 

4M. T. Robinson et al., ANP Quar. Prog. Rep. Sept.
10, 1954, ORNL-1771, p 140.

3

™~
, gy
a0
./""X:}.

.

“a

I

o
 

. w

w o

o5

 

 

--.\__Vsoluble,' as’ 7|s the
The lsotope Ze?®

Thérefo're, the ARE supplied evidence that the
rare gases escape readily from the fluoride fuels
under dynamic conditions. Since the amount of
xenon present in a circulating-fuel reactor is prob-
ably more than the amount produced in one complete
fuel cycle (50 sec in the ARE), the amount remain-
ing in the fuel will be measured by using an ARE
fuel sample for radiochemical comparison of the
concentration of the fission products (Sr8?, Cs'37,
and Ce'4') which come through rare-gas precursors
with those (Cs'®% and Zr?) which do not.

The gamma-ray scintillation spectrometer to be
used for the experiments has been calibrated with
Cs'37, Na?2, and Hg?%3, and the equipment has
been tested with radicactive noble gases taken
from samples of off-gas from the iodine dissolver
used by the Radicisotopes Department of the
Operations Division. The gases used were pri-
marily Xe'33 and Xe'33, In order to prevent the
gamma rays from other short-lived nuclides from
competing unfavorably with those from Xe'3% and
thus making resolution of the Xe!33 isotope diffi-
cult, a charcoal adsorption column (0.190 in, in
diameter and 0.5 in. long) for delaying isotopes
such as fhose of krypton was tested at room temper-
ature by using Kr®3 and the gamma-ray spectrometer.
[t was found that the charcoal trap would delay
krypton 'isofope's'fo'r about 10 min {(at a flow rate
of about 15 cm3/min) and xenon isotopes for about
200 min, Thus the specfrometer would measure
only long-lived 1so‘ropes such as Xe!33,

Fission products may be lost from ARE- fype fuel
by volahll’fy of the flssron product or of one of its
precursors; by interaction of the fission product or

of one of VlfS precursors ‘with the contdlner wall
through c emlccl reachon, for example, B

“should not be lost from the fuel-

by any of these means, since it will be ‘‘carried”’

by the large quantity (50 mole %) of ordinary ZrF,

 

  
  
  
 
 
  
 
  
 
 

 

b eSS A

by an experiment described in Sec. 8, ‘‘Heaf Transfer

PERIOD ENDING DECEMBER 10, 1954

in the fuel and since it has no known rare-gas
precursor (if Kr?® exists, it probably has a very
short half life, say 1 or 2 sec). This isotope will
be used therefore as the basis for normalizing the
yields of the other radicisotopes to be studied,

Volatility of the rare gases will be examined by
studying the yields of S5r%% and of Cs'37 which
come through 2.6-min Kr®% and 3.0-min Xe'37,
respectively, Low yields for these two isotopes
would indicate high volatilities for their parents,
Kr8? and Xe 37, A further check of the experiments
will be made by using Ce'#!, a descendant of
3-sec Xe'4?}, which should be retained in the fuel,
and Cs'3¢, a shielded nuclide formed directly in
fission. Comparison of Zr?> with Cs'3% and Ce'4]
will give a measure of the loss of the important
alkali and rare-earth elements through adsorption
on metal surfaces, since Cs'3¢ and rare-earth
elements will be present only in trace quantities
but chemical zirconium will be present in very
large quantities, as mentioned above,

The data obtained from the ARE fuel sample will
be compared with similar data to be obtained from an
irradiated sample of NaF-ZrF4-UF4(50-46-4 mole %)
that will be maintained solid to retain all the fission
products. The cooling period for the irradiated
sample will be the same as that for the ARE fuel
sample.

LITR HORIZONTAL-BEAM-HOLE
FLUORIDE-.FUEL LOOP

J. G, Morgan
M. T. Morgan

O. Sisman

W, E. Brundage
C. D. Baumann A. S. Olson

R M Carroll W. W, Purkmson
T Solid Stcte DlVISIon

  
 

";‘The vs-econ 'floop fcbrlccted f;)r c1rcu|at|ng flyo-

ride fuel in the LITR has, been opercn‘ed with a

_non-uranlum-bearmg fused sah‘ msxfure for 6 hr in
“the Eaborotory and is bemg mserted in the HB-2

After the |oop ‘has been

facility of fhe LITR

.“ci:'omplefely installed, cnrcu!aflon “of the barren
fused salf\wnll ‘be _resumed. W : '
‘operahon has been obtdlfié‘d, ‘the bcrr-e(n mixture
‘will be drained and the loop will be filled with B

 

atis facfory

 

vTi’ne”desxgn and calculations for the fue

T, SRR}

R AT e

and Physical Properties’ (this report). The curve
obtained for pressure of air vs heat removal gives

125

 

  
 
 
     

   

C ] ¢=| fo-alr.;
!\hea'r exchdnger used in this loop were substantiated

 

T T

 
 

Lt e

 

 

o b ke

 

iE ook gk

“canned for insertion in the LITR.

AVN'P’ 'OUAR‘TERLY PROGRESS REPORT

_.3__‘20 kw at a pressure drop of 50 Ib, 10 kw at 5 Ib,

- and 5 kw at approximately 1 Ib. Since the maximum

_ _lrhedt removal requ:red is expected to be no more
" than 10 kw, the heat exchanger appears to be more
: thcm odequate.

 

CREEP AND STRESS-CORROSION TESTS

W W Dav:s J. C. Wilson
N. E. Hinkle J. C. Zukas
~ Solid State Division

" The ‘stress-corrosion apparatus described pre-

i V-IZVIidt'Js;lys has been bench tested. The design ap-

‘pears to be sound and therefore a new rig is being
Temperature
control to within +2°F at 1500°F has been achieved
by using a Speedomax air-controller to drive the
Variac that supplies the furnace power. Operation

_ _'.ové'r"f)'_-e_ridds_ of 500 hr has shown that, if sodium_
" distills out of the hot zone of the furnace, so little
‘-_sodiu_m is iost fhat ifs effectiveness as a heat

 

SW. W. Davis et al.,, ANP Quar. Prog. Rep. Sept. 10,
1954, ORNL ]77], p 142,

F;g. 9.5. Sandwmh Type UO ,—Stainless Steel Fuel Element After 29% Burnup in MTR (UO Parhcle
Size, 3u). 250X.

126

transfer agent is not diminishe_d".r Welding of some
joints has been poor, and therefore Dy-Chek is
being used in the inspection of all new welds,
The LITR Experiment Review Committee has ap-
proved the apparatus for irradiation, and the
necessary safety circuit for the LITR has been
completed,

In the bench tests the oufsrde of the spec;men
was in air, rather than in sodium as it would be in
the in-pile apparatus. The tests were for periods
of 432 and 866 hr at 1500°F and a stress of 1500
psi. The specimen exposed for the longer time
showed roughly twice as many voids per unit area
on the sides of the tube stressed in tension and
compression as there were on the sides at zero
stress. The maximum depth of voids was about
0.001 in. it was also noticed that the etching (or
perhaps staining) characteristics of the salt-metadl
interface were not the same for the stressed and
unstressed regions when etched in aqua regia,
The reproducibility of the creep data obtained was
not good, and therefore the specimen shape has
been modified in order to concentrate more of the

 

,.
£

»

ERY

It
iR IS a . o i G, - S

 

 

>

 

creep deformation in the gage length. A specimen
tube with thicker walls will be used, and the wall
thickness will be ‘r'\eduééd only at that part of the
tube which is in contact with the fused salt, This
modification should greatly increase the accuracy
of creep measurements,

The MTR creep test equipment is being bench
tested prior to shipment to NRTS,

REMOTE METALLOGRAPHY

M. J. Feldman R. N. Ramsey
W. B. Parsley A. E. Richt
Solid State Division

Two additional Pratt & Whitney Aircraft capsules,
each containing sandwich-type UO,—stainless steel
fuel elements, were opened, and the elements were
examined metallographically, The stainless steel—
uo, particle size was less than 3y for one capsule,
and it was between 15 and 44y for the other. The
capsules were irradiated at a temperature of about
500°F, and the total burnup for each capsule was
29%. Comparison of the results from the element

PERIOD ENDING DECEMBER 10, 1954

with 3-u particle size and the element with 15- to
44-u particle size showed that irradiation induced
greater final hardness in the elements with the
smaller-pdrticle-size material (Figs. 9.5 and 9.6).
Four samples from the smaller-particle-size elements
showed cracking of the core without bending.
Since cracking did not occur in elements with
lower total burnups, it appears that the increased
burnup (29%), together with the small particle size,
can cause cracking of the core, as illustrated in

Fig. 9.5.
MASS SPECTROGRAPHIC ANALYSES
R. Baldock
Stable Isotope Research and Production Division

The isotope dilution method has been used during
the last one and one-half years for analyzing ARE-
type (NaF-ZrF ,-UF,) fuels for their uranium con-
tent.% As a result of this experience, a need was

 

5.. o. Gilpatrick and J. R. Sites, Stable Isctope Re-
search and Production Semiann. Prog. Rep. May 20,
1954, ORNL-1732, p 24,

UNCLASSIFIED
_RMG-

Fig. 9.6. Sandwich-Type UO,-Stainless Steel Fuel Element After 29% Burnup in MTR and o Bend

Test (UO2 Particle Size, 15 to 44y). 250X,

   

127

T
Mo
™M
Mgl
ey
L
bk

 

 
PR \

i R

 

 

 

 

 

 

' ANP QUARTERLY PROGRESS REPORT

felt for further s’rudy of the accuracy of the method.
A testing program was inaugurated in which an
" enriched sample of U,0p containing 99.86 at. %
U235 gnd 0.07 at, % U238, obtained" from B.

Harmatz,” was used as the starting material. This
material was chemically purified for use as a
standard, and the isotope ratio was carefully de-
termined.  The material was then dissolved in

) HNO3 and diluted to make a standard stock solution,

Small portions of this solution were put through the
normal ether extraction process used for extracting

‘ uraniuvm from ARE -type fuel samples. Any change

) ?“of |sotop|c composmon was attributed to contami-

nation by natural uranium present in reagents and
glassware. The observed uranium contamination
in these blanks wos less than 0.5 pg per sample,
cnd ‘thus it appears “that the normal uranium con-

‘ fcmlncmon found durmg chemlccl separchons of

 

_ B Horma'rz, H. C McCurdy, and F. N, Case, Catalog
of Uranium, Thorium, and Plutonium Isotopes, ORNL-
1724, p 2 (May 19, ]954)

fuel uranium is neg||g|ble for the usuol fue[ samples
which exceed 500 ug of uranium.

The second test of the isotopic dilution method
verified the concentration of the standard solutions,
the accuracy of the spiking, and the reliability of
the mass spectrometer. Known mixturés of normal
and enriched uranium solutions were examined in
the mass spectrometer, and the results were com-
pared with the expected U233 concentrations. The
greatest difference between calculated percentages
and measured percentages was only 0.65%, and the
average difference was 0. 37%. |

The results of these studies show that the
isotopic dilution method is reliable and that good
results can be obtained under the conditions which

have been used for ARE-type fuel analyses. The

method of determining burnup based on measure-
ments of the U236 grown in is less subject to error
from sample contamination than is the method
based on measurements of U%38 and is therefore
considered to give the more reliable data.

¥
. LA AR e s i R v ol . S . R

it S B

e

ok i Bk, slscRin, bk

 

il o
Ol gy e e

#

cuhey

fond the volume of evolved hydrogen |s measured

 

@

L : ; o
10. ANALYTICAL STUDIES OF REACTOR MATERIALS - j"’ ;o

PERIOD ENDING DECEMBER 10, 1954

"oy
S vy e 15

-.Wr—’ Y
o
&

 

C. D. Susano
Analytical Chemistry Division

J. M. Warde
,Metollurgy Division

" The research efforf in analyhcal chemlstry was
concentrated primarily on NaF-KF-LiF-base reactor
fuel, and, in particular, on analysis after it had
been utilized as- a solvent for the reduction of
UF, by U° to UF,. Tentative methods for the
determination of U° and UF, were developed.
Studies were continued on the determination of
oxygen as oxides in fluoride fuels. Development
work was completed on the determination of sulfur
in various reactor fuels and coolants and in off-
gases from the production of fluoride fuels. Be-
cause of the urgency of other activities, no further
work was done onthe determination of alkali metals
in NaF-KF-LiF-base fuels.!

ANALYTICAL CHEMISTRY OF
REACTOR MATERIALS

, J. C. White

Analytical Chemistry Division

Deferm:nutlon of Uranium Metal in
Fluorlde Sal't Mleures

A S Meyer, Jr. B. L. McDowell
~ Analytical Chemlsfry Division
A method based on the measurement of the
hydrogen derived from the decomposition of UH3

was developed for the determination of U° in
proposed reactor fuels.

is converted fo UH by heo’rmg in an atmo _phere’:'
is de-

of hydrogen at 250°C for 1 hr. "The UH
composed by hea'rmg ina stream ‘of CO at 400°C,

bY the SOIUfronéof KOH{gkzs produced by ‘rhe reduc-:‘ B
8 lhon °f CO to CO." The’ CO is reduced at femper*’”.%

:tweenz400 and 600 OC by both U® and UF,.

 

). C. White, G. Goldberg, and B. L. McDowell, ANP
Quar. Prog. Rep. Sept. 10, 1954, ORNL-1771, p 152.

In this defermmahon,r UO‘
‘osphere  niodified procedure that is being considered, UH,

“'will be decomposed in an atmosphere of c:mmoma

ofH SO

  
  
   
  

" "Further’
developmg a titrimetric mefhod for the defermlno-'
_, tion of trivalent uranium in the presence of tetrava-
“lent uranium in Hluoride salt” mixfures. Under appro-

priate conditions, trivalent uranium is selectively

   

 

The volumes of gas derived from reactions 1 and
2 are reproducible; 1.5 moles of CO is formed for
each mole of UH, decomposed when the ignition
is carried out at temperatures between 500 and
600°C, and 0.5 mole of CO is formed per mole of
UF, under similar conditions. The CO must be
removed from the effluent gases before the meas-
urement of hydrogen is made. This is accomplished
by passing the gases through a tube packed with
] O and powdered pumice, which is maintained at
a temperoture of 150°C. The CO is oxidized to
CO, in the following reaction:

5C0 + 1,05 —> 5C0, + |,

When the method was testedby analyzing samples
of pure uranium metal, the hydrogen evolved
corresponded to 97% of the theoretical value with
a coefficient of variation of 2%. Satisfactory
precision has also been obtained for the deter-
mination of U° in NaF-KF-LiF-base fuels and in
UF,. The concentration limits of the methcds have
not been measured experimentally, but, on the
basis of the quantities measured, concentrations of
U® as low as 0.05% may be determined when 1-g

samples are utilized.

Since the [,0. tubes may become inactivated
after only limited service, a more dependable
method of eliminating the CO is needed. In a

and the hydrogen will be measured over a solution

Determmahon of Trlvaleni Urumum |
- in Fluoride Fuels

“A. S. Meyer, Jr. W, J. Ross

- D. L. Manning B. L. McDowell
: Analyhcol Chemistry Division

 

 

    

mvestlgahons were carried ouf for

oxidized by cupric chloride, CuClz, titanium tetra-

129
A

WA é»m? 1.3

$ A

 

 

 

 
 

 

ANP QUARTERLY PROGRESS REPORT

chloride, TiCl,, and methylene blue, C, ¢HygN5SCHL
Of these reagents, methylene blue appears to be
the most promising reagent for a possible routine
method for the determination.

Cupric Chloride (CuCl,). Previous attempts to
oxidize UF; by Cu(ll) in H,SO, solution were un-
successful. Since the formal oxidation potential of

the [Cu(H),Cu(l})] couple is increased in solutions

of high chloride concentration, experiments were

performed to ascertain whether quantitative oxida-
fion of trivalent uranium could be achieved by

~ adjusting the acidity and chloride concentration of

the CuCl, solvent solution.

j Sainples of UF, were dissolved under an atmos-
here of CO in acidic solutions of NaCl which
confcmed a mecsured excess of a standard solu-
tion of CUC| Affer dissolution of the samples,
the unreduced CuCl, was determined iodometrically
and 1rhe trlvalent urcmlum was calculated on the
bc15|s of the stmchlometry of the equation

Cu2* 1 U3t —> Cut + U4t

‘When the dissolution was carried out in a solu-

tion 1 M in HCl and 2 M in NaCl, the results were
in agreement with those obtained by the hydrogen
evolution method.?2 Lower values were obtained
when either the chloride concentration or the
acidity was altered.

Samples of UF3 which had been fused with NaF-
KF-LiF evolved hydrogen and subsequently
yielded low results when analyzed by the above-
described procedure. The method therefore appears
to be of limited applicability.

Titanium Tetrachloride (TiCl,). It was found that
UF, dissolved without evolution of hydrogen in
solutions of TiCl, in concentrated HCl to yield
dark-brown solutions of TiCl.. The color of these

‘solutions, when diluted, reverts to the rose tint

usually associated with Ti(lll) solution. When
these diluted solutions are titrated with a standard

~solution of K,Cr,0,, the equivalents which are
_consumed at the first sharp change in potential

correspond to about 95% of the trivalent uranium
as determined by the hydrogen evolution method.
If the solutions are then heated to a temperature

" of 90°C, the total uranium can be determined by
'ox1d|zmg U(lY) to U(VI) with furfher addition of

'rhe dlchromcn‘e soluhon.

 

 

Bl e el R, S

2D. L. Mahmng, W. K. Miller, and R. Rowan, Jr.,

Methods of Determination of Uranium Trifluoride, ORNL-

1279 (Apr. 25, 1952)

. 1éo"r

While this is the only method which has been
found to give a titration for trivalent uranium
without subsequent back-titration of excess
oxidant, it is not readily adaptable to routine
analysis because of the slow equilibrium of the
electrode potentials and the rapid oxidation of
titanous solution by air after the dilution of the
concentrated acid solutions. Experiments are now
being conducted to determine whether the TiCl,
can be titrated in the concentrated acid with
bromine which is generated coulometrically.

End points obtained potentiometrically in the
concentrated solutions were found to be poorly
defined, but sharp breaks with potential changes
of about 400 mv were obtained: with polarized
platinum electrodes. Titrations between 5 and
10% in excess of the theoretical value were
obtained when UF, samples were titrated cou-
lometrically. This excess titration is probably a
result of diffusion of TiCl, from the cathode com-
partment. ‘

Methylene Blue. A proposed method based on
the oxidation of trivalent uranium to the tetravalent
state by methylene blue was developed for the
determination of UF, in a fluoride fuel. In the
proposed procedure, the sample is dissolved in @
measured volume of 0.02 N methylene blue solution
in 3 to 6 N HCl| under an atmosphere of CO, by
stirring for 2 hr at room temperature., The excess
methylene blue is then titrated to its reduced
state, methylene white, with 0.05 N chromous
sulfate solution. The equations involved in the
determination are as follows:

2U3* + methylene blue + 2HY — 204"

+ methylene white

2Cr?* + methylene blue + 2H"—> 2Cr3%
+ methylene white

Because of the intense color of the dye, the
visual end point, blue to green, is sharp and re-
producible even when the titrations are carried out
with 0.025 N chromous sulfate.

For the determination of trivalent uranium in
UF,, the coefficient of variation is approximately
1%, and the results are in excellent agreement with
those obtained by the hydrogen evolution method.?
While somewhat poorer precision was obtained in
the analyses of NaF-LiF-KF-UF,.-UF, samples,
the variations were probably a result of hetero-
geneous sampling, as is also indicated by the

0627 134

N
 

L

e

vader

*

  

lrreprodumb:llty of determinations by the hydrogen

evolution method. In the ona[yses of some of the

NaZrF .-UF samples, low results were obtained
becouse fhe samples were not completefy dissolved
in 6 N HCI.

~ The methylene blue procedure requires less
operational time than the hydrogen evolution
method, and, for the samples which have been

‘tested, it appears to give results of comparable

precision and accuracy. Tests are being conducted
to determine the extent of the reactions between
methylene blue and uranium hydride, uranium
metal, and other metalllc contaminants of the
fluoride samp]es.

‘In order to carry out subsequent determinations
of the total uranium concentration on the same
samples, the final soluhons from the methylene

_‘blue determlnotlon were titrated with oxidizing

agents. Titrations with Ce(lV) and Fe(!H) solu-
tions did not yield stoichiometric end points.
Al_'t»ht_:ug'h'kfhe reaction with K_Cr 07'v'vds'exfreme|y
slow, two well-defined potentiometric end points
were obtained when the solutions were titrated
slowly. The second of these end points corre-
sponded to the reoxidation of methylene white to
methylene blue plus the oxidation of the U(IV) to
U(V!). The stoichiometry of the first break in the

“potential curve has not yet been accurately de-

fined, but it appears to be consistent with the
oxidation of U(IV) to U(V). Further investigation
is under way to elucidate the nature of these
changes in potential.

Determlnohon of Oxygen in Fluonde Fuels
,.A S, Meyer Jr.»{ 4 M Peele

  

 

 

3A. S, Meyer, Jr., and J. M. Peele, ANP Quar. Prog.
Rep, Sept. 10, 1954, ORNL-1771, p 148.

 

  

~metallic oxides.

wafer alcohol mixfure.

PERIOD ENDING DECEMBER 10, 1954

is used, a concentration of 1% water corresponds
to 235 mg of oxygen. From the above linear
relation the calibration curve can be extended to
concentration ranges too low for direct measure-
ment by weighed primary standards.

At the temperatures at which liquid HF could be
safely maintained in the reaction vessel, the rate
of the reaction between metallic oxides such as
uo, and HF was found to be too slow for applica-
’rlon to analytical measurements, The oxides were
found to react rapidly with fused KHF, according
to the postulated reaction

4KHF , + U0, —> UF,, + 4KF + 2H,0

The procedure has been modified accordingly to
carry out the dissolution of the sample in molten
KHF .

In the revised procedure, the sample is mixed
with about five times its weight of anhydrous
potassium fluoride, KF, in a platinum crucible,
The crucible is then placed in the reaction vessel
and the KF is converted to the acid salt by trans-
ferring a portion of the HF acid from the conduc-
tivity cell to the reaction vessel. After the excess
HF has been returned to the cell, the KHF
fused by heating the reaction vessel to about
300°C., The water is then transferred to the con-
ductivity cell for measurement by repeated distilla-
tions with portions of HF.

Because part of the KF was carried over to the
cell during the distillation, the original apparatus
was modified by placing a silver-lined vessel,
which is similar in design to a Kjeldahl trap,
dlrec’riy above the reaction vessel. The revised
~ procedure is now being tested on pure samples of

  

 

     
   
  
 
 
  

 

| ermlnahon of Sulfur

J C Whlte G Goidberg
Analyhcal Chemustry va15|on

Methods were modified ‘and adapted for the deter-

'm‘lnahqn Of 'rraces of sulfur_m ’rhree types of

 

   

“Qualitative tests with
starch-iodide paper were performed to confirm the

0627 138 131

 

NS PR~ W raere—— pyp—-r g

 

 
e i

 

 

 

1929).

 

ANP QUARTERLY PROGRESS REPORT

presence of sulfur. A colorimetric method? was
adapted to determine the concentration of sulfur
quantitatively. In this method, sulfur is released
from solution as H ,5 by acidification with HCL.
The gas is cdsorbed in a solution of zinc acetate
solution, and, upon the further addition of ferric

'dm'moriium'Sulfate, a reaction occurs between

the organic reagent and sulfur to form methylene

blue, an intensely colored dye. The absorbancy

of the soluflon of methylene blue is measured with

a phofomefer' at 670 my. The method is extremely
'4sensmve _wnh a workable range of from 5 to 50 pug
per 100 ml of solution. This range corresponds fo

a practical lower limit of determination of about

1 part of sulfur per million parts of sodium.
Analyses of the sodium revealed that the sulfur

was not unn‘orm]y distributed. In particular spots,

the concenfrcmon was of the order of 0.1 to 1% or

higher. The sulfur content of the sodium sampled
from the bulk of the trap was, however, much lower,
50 to 100 ppm, and that of the sodium in the system
was less than 1 ppm. These results indicate that
sulfur in sodium precipitates rapidly as sodium
sulfide and can be effectively trapped and removed
from the system. The solubility of sodium sulfide
in sodium is evidently extremely small.

Sulfur in Flubride Salt Mixtures, The presence
of sulfur in mixtures of fluoride salts which are
being considered as proposed reactor fuels is con-
sidered deleterious, principally from the standpoint

of corrosion. The major problem in determining

sulfur in fluorides is that sulfur exists in at least

two oxidation states, sulfaie and sulfide, and, in
order to adapt the methylene blue colorimetric
method, the sulfate must be reduced to the sulfide.
The reducing mixture recommended by Johnson
and Nishita® for sulfate in soils is used for this
purpose. About 1 g of the fluoride mix is heated
at the boiling point with 4 ml of a reducing mixture
composed of 15 g of red phosphorus, 100 ml of
hydriodic acid, and 75 ml of formic acid. Reduction
of the sulfate is complete within 30 min, and
then the colorimetric method can be utilized. When

“sulfur is found in fluoride mixtures, the principal
| portion of the sulfur is in the form of sulfate rather
| than su|f|de as mlgh'r be expected

 

4. F. Fogo and M. Popowsky, Anal. Chem. 21, 732-4

5C° M. Johnson and H. Nishita, Anal Chem. 24, 736

(]952)

132

 

Sulfur in HZ-H‘F Gas Streams. The odor of HéS '

has often been noted in gas streams in the fuel
production work. Two sources of this sulfur are
known: sulfate contaminant in fluoride salts and
fluorosulfonic acid, HSO,F, in hydrogen fluoride.
Sulfur from the first source is known to be of the
order of a few parts per million. A test of the
sulfur content of hydrogen fluoride was made by
dissolving the gas in a solution of NaOH, deter-
mining sulfur in the solution by boiling the basic
solution in the presence of H,0,, and precipitating
the sulfate as BaSO, with BaCl,. The sulfur

concentration was 2.7 mg per liter of HF,

A semiquantitative method for determining sulfur
as sulfide in off-gas from fuel production was set
up in which the gas was passed through a 6%
solution of NaOH. An equal volume of bismuth
nitrate in glacial acetic acid was added to the
scrub solution, and the turbidity, as a result of
formation of Bi,S,, was compared with previously
prepared standards. The procedure was made more
precise by measuring the absorbancy of the turbid
solution at 350 my in a 7.5-cm cell with a total

volume equal to the volume of the test solution. -

The concentration of sulfur found by this pro-
cedure ranged from 2 to 35 pug per liter of off-gas.

Determination of Fluoride in

NaF-KF-LiF-Base Fuels

J. C. White B. L. McDowell
Analytical Chemistry Division

Investigation was continued on the feasibility of
a spectrophotometric titration of fluoride based on
the decolorization of a zirconium complex or lake,
Zirconium alizarin sulfonate and zirconium Erio
Chrome cyanine were tested as possible titrants.
A titration cell based on the design of Sweetser
and Bricker® was fabricated so that the Beckman
Model DU spectrophotometer couldbe usedto meas-
ure absorbancy. Although the data have not yet
been thoroughly evaluated, the technique does not
appear to be feasible for application to NaF-KF-
LiF-base fuels because of the slowness with

which equilibrium is reached. The procedure will

be evaluated before further work is done.

 

5P, B. Sweetser and C. E. Bricker, Anal, Chem. 25,
253 (1953).

0627 135

oW

¥

o
Sl .

L e b B iR

 

vy

[R13

L

PETROGRAPHIC INVESTIGATIONS OF
FLUORIDE FUELS

G. D. White, Metallurgy Division
T. N, Mch:y, Consultant

Petrographic examinations were made of several
hundred samples of fluoride melts. The majority of
the samples were from alkali fluoride systems con-
taining UF ., |

When small amounts of UF, (<20 mole %) are
melted with NaF- KF- LiF, KF-NaF, or KF, a red

compound is formed which is thought to be K UF,.

In melts which contain only KF and UF,, this com-
pound is isotropic. However, in the ’rwo with NaF
present, the compound exhibits slight anisotropism.
This is thought to be due to a slight solubility of
NaF in the K sUF (.

“Some work was done on the systems KF-LaF,,

NaF-LaF,, and RbF- LaF,. The first two systems

7'<:0n’rcm colorless, 1:1 compounds which are uni-
axial posmve with refractive mdlces in the

wcmn‘y of 1.50. The system RbF- LaF contains
a 1:1 compound which is biaxial posmve with a
moderate optic angle and refractive lndlces also

near 1.50.

ANP SERVICE LABORATORY
J. C. -Wh"ife' C.R. W||||c1ms

~ W. F. Yaughan
Anclynccl Chemistry Division

The nonumforml'ry of the NaF-KF-LiF-base fuel
samples that have been received in the past
necessitated 'a'chcnge in the handling methods.
The entire bqfch of fuel resulting from a particular

experlmem‘ is now submn‘ted for analys:s. The_w_
‘-'"“sample |s gro"'

 

' "'d to pass a No. 50 s:eve quarfered

    

 

PERIOD ENDING DECEMBER 10, 1954

sampling, and weighing of these hydroscopic
samples are carried out in a dry box in order to
ensure that the samples are always dry and that
the U3* has not changed valence. Since the date
of inception of this procedure, more uniform re-
sults have been obtained. :

fn addition, a change has been made in the
method of determining uranium in samples received
for analysis. Heretofore, uranium was determined
by the zinc reduction, ceric sulfate titration
method. The present method? is that based on the
reduction with chromous sulfate and final titration
of the U(IV) to U(VD) with standard potassium
dichromate., This determination is carried out at
elevated temperatures with the Beckman Model K
automatic titrator.

A total of 1,698 samples was received, and 1,587
samples, involving 11,541 determinations, were
analyzed and reported. A breakdown of the work
load is given in Table 10.,1.

 

7). s, Decker, Application of Beckman Model K
Automatic Titrator to the Determination of Uranium (to

be published).

TABLE 10.1. SUMMARY OF SERVICE
ANALYSES REPORTED

 

_The balance of ~ Tot

 

Number Number
- of of
Samples Determinations
Reactor Chemistry 1,118 8,038
Experimental Engineering 448 3,421

 

Miscellaneous 21 82
Total 587

 

 

0627 137 133

 

 

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

11. RECOVERY AND REPROCESSING OF REACTOR FUEL

| D. E. Ferguson

G. I. Cathers J. T. Long

M. R. Bennett S. H. Stainker

W. K. Eister H. E. Goeller
R. P. Milford

Chemical Technology Division

A plant for recovering (in seven batches) the
_uranium from the ARE fuel and rinse by the fluoride-
volatility process is being designed, and con-

~ struction is scheduled for completion by December

31, 1955, It is estimated that the amount of ma-
terial tobe processed will be 12.4 i3 of NaF-ZrF ,-
UF , containing 65 kg of uranium, This plant will
‘demonstrate, on a pilot-plant scale, the feasibility
-of the fluoride-volatility process as applied to the

_ processing of the fuel from a circulating-fuel
- aircraft reactor. The feasibility of this process

(Fig. 11.1) has been established on a laboratory
scale.’*? The basic equipment as now envisioned
will consist of a fluorinator, an absorption column
packed with NaF, a cold-trap system, and a fluorine
disposal unit,

The fluoride-volatility process can be adapted
for recycling uranium to a fresh fuel concentrate
by adding a UF, reduction step and dissolving

 

—-—t

D. E. Ferguson et al., ANP Quar. Prog. Rep. Sept.
10, 1954, ORNL-1771, p 12.

' 2G. . Cathers, Recovery and Decontamination of
Uranium from Fused Fluoride Fuels by Fluorination,
ORNL-1709 (May 26, 1954).

-

“Eoner
ORNL-LR-DWG 4386
NaF ABSORBENT BED
650°¢C

<0.02% U

 

 

 

 

 

 

]

 

 

 

 

/ FLUORINATOR

650°C r——

\\\\Q,

 

 

 

 

 

 

%
ARE FUEL F, WASTE
(~ 6 mole % UF,
Z IN NaF~ZrF,)
' + COLD TRAP
-80°C

WASTE SALT 99.95 %, OF U

>99% F.P.
<0.02% U

Fig. 11.1. Fluoride-Yolatility Process for ARE
Fuel.

134

the resulting partially decontaminated UFA in
NaF-ZrF,. However, if the uranium is to be
returned to a diffusion plant or made into heter-
ogeneous fuel elements, additional decontamination
must be obtained. This may be accomplished by
adding a complete UF , fractional distillation step
to the procedure shown in Fig. T1.1,

This method of uranium recovery and decon-
tamination can also be used for processing heter-
ogeneous reactor fuel elements of the type that
can be dissolved in fused fluoride salt by means
of hydrogen fluoride.  Compactness of plant,
operation at atmospheric pressure, and economical
waste disposal are some of the advantages of this
type of process,

FISSION-PRODUCT REMOVAL

Three methods have been tested for removing
fission products from the UF ,-F, mixture obtained
by fluorination of a wuraniumecontaining fused
fluoride salt: scrubbing the gas with a molten salt
(NaF-ZrF ), scrubbing the gas with C,F,,, and
passing the gas through a solid NaF absorbent
bed at650°C. Of the three methods, NaF absorption
is the best; it gives an over-all gross beta decon-
tamination factor of more than 104,

Use of a molten salt scrub, NaF-ZrF , (56-44
mole %) at 650°C, followed by sublimation, for
removing fission-product activity from the UF,
product of the fluorination step gave, as reported
previously,] over-all decontamination factors of
2 % 10% and 430 to 750 for gross beta and ruthenium
beta activities, respectively. However, in the
scrub step alone, the decontamination factor for
the ruthenium, which is by far the most important
volatile activity in the long-cooled material, was
only 6, as shown by determination of the ruthenium
in the molten salt. In a repetition of this experi-
ment, the ruthenium decontamination factor in the
scrub step was only 2 {run 1, Table 11.1), and the
over-all ruthenium decontamination factor was 350.

T 6T 128

a
 

i, G o,

 

 

n

T

A%

)

19% 5

¥

PERIOD ENDING DECEMBER 10, 1954

DECONTAMINATION OBTAINED IN THE FLUORIDE-VOLATILITY
PROCESS BY SCRUBBING UF,

TABLE 111,

Synthetic ARE fuel, prepared by hydro-fluorinution of 6 g of irradiated

uranium metal in 67 g of NaF- ZrF4 (56+44 mole %), fluorinated at
650°C wn‘h 40- to 80 fold excess fl uorine.

_ -R_un L UF6 producf p]us excess fluorme passed ihrough 67 g

' of molten NuF-ZrF at 650°C into a dry-ice trap and

then resublimed mto second trap.

Runs 2 and 3: UF6 product separated from excess fluorine in dry-ice
trap and then volatilized through « C8F16 still into a

second trap.

 

Beta Decgntumi nation Factors

 

 

 

Activity Run 1 Run 2 Run 3
Over-all Serub* Over-all Scrub* Over-all Scrub*
Gross 4800 2.7 290 .2 250 2,2
Ru 350 2.3 16 14
Zr 3.2 x 104 6.8 x 10° 4.7 x 104
Nb 1800 47 1200 86
2 x 108 2 x 108 2 x 10°

TRE

 

*Decontamination factor calculated for scrub step alone on basis of activity extracted from scrub material,

Scrubbing the UF product from the fluorlnahon
and sublimation sfeps with fluorocarbon (C.F )
did not improve the decantamination {runs 2 and 3,
Table 11.1). Approximately 9 g of UF, sublimed
from a fluorination run, was passed into the bottom
of a CSF distillation column (0.5 in. in diameter
by 14 in. high, packed with /32-|n. nickel Fenske
hehces) that was operctmg at full reflux at about
102°C. The UF6 was taken out of the still as gas’
through an 80°C head. The over-all ruthenium

. beta decontammohon factors of 14 and 16 |
T in these runs_ s:gh‘[flcanfly grea er‘fhanm
the deconfommot‘lon fcctor usuolly obtamed m ’rhe -

 

 
 
  

 

   

“In “an explorotory'tr'lul -

*solid NaF (Tablg 11,2,

 

 

1 (run 1, Table ”, 2), ‘the UF -F stream from the__
'fluormcl’ror was sed ’rhrough a 25-9 bed of .
Y-in. NaF pellets ‘at 300°C.” About 32% of the

8

had to be refluorinated from the NaF bed at 650°C

before being trapped. However, the gross decon-

 

“obtained”
“Tactivities is probably a filtration effect rather than

 

“runs.

  

“salt waste.

uranium was ot absorbed by the NaF, while 38%

tamination factors in the two fractions were of the
same order, that is, 2 x 10°. The uranium loss on
the NaF was only 0.01%. Two experiments were
then performed with 30-g beds of 20- to 40-mesh
NaF held at 650°C to eliminate the refluorination
step (runs 2 and 3, Table 11,2}, The results in
both tests showed high absorption of ruthenium
and niobium beta activities, and the over-all gross
beta decontamination factor was greater than 104,

The shght increase ([ess than 10) in decon-
“tamination from zrrcomum and total rare-earth
on the

an absorption effect. The uranium loss

_20- to 40-mesh NaF was 0.03%.

A poor ‘material balance for ‘ruthenium and niobium
“beta actlvrty ‘has been observed in all fluorination
In general 1 to 10% of the rufheruum and
“about 50% of the niobium remained in the fused
After run 3 (Table n 2) the entire
nlckel fluormatlon 'vessel was cut into three parts
‘and d:ssolved in dllufe nitric ocui fo obtain the
material balance summary “shown in Table 11.3,
Nearly all the niobium activity was accounted for
in the reactor and fused salt, while 75% of the

135

& S Ty

ff} b s

bl WP g f 3 w-s ‘uj'
oy .,

 

 

 
* 'ANP QUARTERLY PROGRESS REPORT

TABLE 11,2, DECONTAMINATION OBTAINED IN THE FLUORIDE-VOLATILITY
PROCESS BY PASSING UF, THROUGH SOLID NaF '

| Synthetic ARE fuel, prepared by hydrofluorination éf 6g of irradiated uranium metal in 67 g
of NaF-ZrF, (56-44 mole %), fluorinated at 650°C.,
Run 1: UF& product plus 28-fold excess fluorine passed through 30-9 bed of NuF pellets
- (/ in.) held at 300°C; 38% (fraction A) was absorbed and then refluorinated off at
650 C into a dry=ice trap, while 32% (fraction B} was not absorbed and was dlrecfly
trapped,
Run 2: UFéproduct plus 11-fold excess fluorine passed through 30 g of 20+ to 40-mesh NaF
held at 650°C and then trapped,

<~ Run 3 UF6 product plus 9-fold excess fluorine passed through 30 g of 20- to 40-mesh NaF
held at 650°C, trapped, and then resublimed.

 

Beta Decontamination Factors

 

 

 

 

 

 

 

 

 

L Run 1 Run 2 o Run 3
- Activity ;
: Sl Fraction Fraction i e
A B Scrub®* Over-all Scrub* Oversall Scrub*
Gross 2200 1600 3.5 2.3 x 104 50 1.6 x 104 24
" Ru 140 120 3.0 5200 90 3100 50
Zr 4300 870 1.6 4.1 x 104 7.3 8.3 x 104 9.4
Nb 430 47 ' 17 3400 280 2500 100
TRE 108 5 x 104 3.5 3 x 10° 7.3 4 x 108 2
*Caleulated for scrub step alone on basis of activity extracted from scrub material,
- TABLE 11,3, MATERIAL BALANCE FOR RUTHENIUM AND NIOBIUM ACTIVITY IN
A FLUORIDE-YOLATILITY RUN
_ Ru 8 Nb 8
Location (% of original) (% of original)
Bottom of reactor {in contact with 4 . 0.4 -
fused salt) S
;g ...~ Fused salt waste 0.8 69
' "Tép §f re’achr (not in contact with 8 _ 3___
_fused salt)
" Qutlet tube 10 27
: | NqF qbsol;benf 2 . - N .1 _'
Total 25 100 , |
3 ]3 . ol . ) ) . o~ oy - ,
| © 627 140

 

 

X,

»

 
 

s mE

"

i gF

»

gy

 

  

o olumma

ruthenium was apparently volatilized out of the
reactor entirely. Ruthenium and niobium plated
out heavily on the metal walls of the reactor. The
distribution ratio between metal wall and salt was
about 5 for ruthenium activity and less than 0.01
for niobium activity,

The behavior of plufomum in the fused salt
fluoride-volatility process is important inprocessing
fuel that has a high proportion of U238, The
results of many fluorination runs at 650°C show
that only about 0.01% of the plutonium is carried
over with the UF6 product,

Further study was carried out on the absorption
of UF, in NaF-ZrF, (56-44 mole %) at 650°C.
In one trlal 9 g of UF, was absorbed in 30 g of
NaF-ZrF , in 1 hr with no noticeable loss. How-
ever, passage of helium through the fused salt,

~ either concurrently with or after the UF , addition,
produced considerable fuming, A positive test for

fluorine with potaséiurh iodide paper was obtained,
which indicated reduction of the adsorbed UF to
the tetra- or pentavalent form of uranium. A
gravimetric test, based on the reaction of fluorine
with sodium chloride, showed that the breakdown
of UF, to Fé and UF4 amounts to 1 to 2% per hour
at 650°C.

Engineering information is needed on two steps
in the fluorlde-volthlfy process: fluorination and
UF, cold-trapping, Effective contacting of fluorine
with molten ARE fuel is desirable to minimize
fluorine consumption and gaseous waste and to
assure complete recovery of the uranium. The cold
trap should be operated in such a manner that all

the UF, will be condensed and collected on fhe' )
walls of the trap rafher than lost as

¢ smoke.

 

Cons;deruble experience “has’ been

- "‘,K’25 in fhe operatlon"‘bf‘ cq[d traps equlpped wu’rhm'
catch any UF6 ‘escaping
'fiiust be mmlmlzedfi“_

Also, the experzence'f'riflrUzss

”;'H:'Therma[ Reactor (STR) and the U2 5_.stainless
‘steel fue| elemenfs proposad for ‘the Submarine
B !ntermedmte Reaci‘or w(SIR) and_the Army Package

 
 
  
 

 

bromlde, whlch has physu:al propert!es very neur!y

the same as those of molten NuF-ZrF4. No effect”

of equipment dimensions on vertical mixing or
turbulence was found, and the effects of physical

 

 

 
 
 
  

 
 
 
 

  

_Lwhlchfl th-e\‘ fl uoride-
fostered and expanded “Fot example, fuel element

PERIOD ENDING DECEMBER 10, 1954

properties of the liquid could not be demonstrated
by a simple correlation of friction factor with
Reynolds number, such as is found in ordinary
fluid flow through pipes. The acetylene tetra-

‘bromide studies showed that vertical mixing of the

fiquid phase was induced by a gas rate of 9.5 cfm
in a 12-in,-dia column, Equipment is now being
assembled to verify these conclusions with molten
NaF-ZrF .

Cold traps to be used for a quantitative study
of the effect of temperature and gas flow rate on
the completeness of UF, removal are being con-
structed. The design of these traps is based on a
K-25 cold-trap design.

APPLICATIONS OF FUSED SALT~VOLATILITY
PROCESSES

A long-range study has been made to survey the
over-all feasibility of fused salt—volatility tech-
niques in the chemical processing of ARE-type
reactor fuels and certain types of heterogeneous
reactor fuel elements. The volumes of radioactive
waste from such processes should be much lower
than those from the aqueous processes, and
processing costs should be low, even though a
means of disposing of excess fluorine will have
to be provided. Total chemical costs, which in
present aqueous processes represent approximately
10% of operating costs, have been estimated to be
20¢ per gram of U235, The operational procedure
is much simpler, and the equipment should be

inexpensive, even though nickel will be required

as the material of construction.
Results ofrecem‘ work on fuel element dl':SOIUflon

  

dér V

n of fused fluoride salt tech-

' mques' in a vcrlety“of radloochve chemical pro-

‘Cesses. Probable appllcahons in addition fo those
for ARE-type fuels include the processmg of the
-ercoloy fuel elemen’rs “used in the Submarine

     

     

vo!ofll‘n‘y techmques may “be

dissolution in the conventional nitric acid or
sulfuric acid systems is very complex from the

g ' 137

 

 

 
fakh b

 

 

 

ANP QUARTERLY PROGRESS REPORT

standpoint of both operational and corrosion prob-

" lems but is efficient and rapid in the high-tempera-
”fure hqmd medlum of fused fluoride salts.

Anrcrofi Reucior Fuels

The fecovery and reprocessing schedule for

' fluid-fuel curcroft reactors will probably be dictated

by aircraft reactor and turbine maintenance sched-
ules rather than by the rate of formation of neutron
poisons. [t is anticipated that operation will
follo’w a schedule such as: (1) one day of oper-

‘ation’ aand one day of downtime for an accumulation

“of s seven operahng days, (2) seven days of downtime

for minor maintenance, (3) repetition of this sched-
ule until 1000 operating hours have been accumu-
lated. After 1000 hr of operation, the entire reactor

will be dumped and the fuel will be reprocessed.

The anMCIpated cooling period before repro-

‘:"'cessmg will be 10 days to allow decay of short-

lived activities. The minimum decontamination
factor requ;red for the process would be approxi-
mately 100 for poison removal only. The other
steps in fuel makeup can easily be handled remotely.

The following essential steps are used in the
chemical processing (Fig. 11.2). First, the UF
in the molten fuel mixture is fluorinated to volatile
UF, by introducing a 10-fold excess of elemental
fluorine to achieve separation and partial decon-
tamination from the other fuel components and
fission products; second, the UF, is reduced to
UF, with hydrogen in the gas phase in the Y
reactor, as designed by K-25; and third, the re-
sulting UF, is added to 2 moles of NaF to prepare
a fuel concentrate for subsequent return to the
reactor. Presentknowledge of this systemindicates
that all steps are adaptable for radicactive remote
operation. Considerable engineering development
and operational experience have been obtained
with the second and third steps, while extended
laboratory development has indicated the feasibility
of the dlrec’r fluormatlon of the molten fuel mixture.

Heferogeneous Reactor Fuels

The separation of U233 from zirconium alloy
fuel elements is currently the most promising

- 'applicafi'onrdf"the_ fluoride process to fixed fuel
’élemenf reactors. The dissolution rate of zirconium

Cwith HF in the NaF- ZrF fused salt is very high,

that is, 22 to 35 mlls/hr (Table 11.4). The range

is  probably due to metallurgical differences.

138

Although the ratio of zirconium to uranium is very
high in STR elements, the cost of HF to dissolve
zirconium will be only 5¢ per gram of U235 A
probable method for zirconium separation as a
guide for process development is indicated by the
flow diagram shown in Fig. 11.3. This flow sheet
is similar to that for fluid-fuel processing, but it

“includes hydrofluorination of the fuel element in a

TABLE 11.4, RATES OF HF PENETRATION OF
VARIOUS METALS AND ALLOYSIN A
TYPICAL FUSED FLUORIDE
SALT BATH

Bath composition: NeF-KF-ZrF4 (7-48. 5-44 5 mole %)
HF flow rate: 250 em>/min

Temperature: 675°C

Nitrogen or argonblanket incases ofopen test vessels

 

Penetration Rate
Material

 

{mils/hr)
Vanadium Not detected
Silicon Not detected
Nickel 0.0001
Monel .02
Molybdenum 0.03
Tungsten 0.06
Silicon carbide 2*
Type 304 stainless steel 4
Type 347 Nb stainless steel 7
Niobium 7
Tantalum 8
Manganese 10
Mild steel (Unistrut) 13
Thorium, 1/8-in. plate 14
Uronium 17
Zirconium 22 to 35%*
Chromium 31
Titanium 31
Zircdloy-2 22 to 46**
95 wi %.uranium-—s wt % 50 '
zirconium
Tin Sompledissolvéd
instantly
Zinc Sample dissolved
instantly

 

*Material disintegrated ond left suspendé&ufi‘firticles. o

**Range is due to metallurgical differences in indi-

vidual specimens,

2
o
w3
ot

o
7
-

-
.
e
NS

1

 

3

-
 

 

 

}

© .

N .
2
2
& v
&
&
S
=
-
@
®

 

»
>

0

(2)

 

 

 

 

 

 

 

PERIOD ENDING DECEMBER 10, 1954

 

 

 

: Fa.F.RFy 2HF(H,)
e — = WASTE WASTE
| H
L Y o, UFg UFg 2
2
B d.f.>102 B d.f.>10°
Yy , Y Y
7 e U, —
// Y REACTOR UFq
SPENT FUEL _

 

-80°C(TO 50°C)
COLD TRAP

650°¢C
FLUORINATION

 

WASTE STORAGE [

 

NaF - ZrF, WASTE
<99% F.P.
<0.02% U

 

 

 

 

 

~800°C
REDUCTION

SEeREt

ORNL-LR-DWG 4387

2NaF

(3)
Na,UFg

FUEL

CONCENTRATE

 

REACTOR FUEL Bl

 

 

3.2% U?°°F,
46.4% ZrF,
50.4 % NaF

 

 

(1) UF,+10 F,—> UF+ 9 F, (EXCESS F, REQUIRED FOR COMPLETE UFg DISTILLATION).

(2) F, DISPOSAL REQUIRED; RECOVERY OR RECYCLE NOT ECONOMICAL

(3) REPRESENTS PREFERRED FORM OF CONCENTRATE FOR ADDITION TQO REACTOR.

Fig. 11.2. Tentative Flow Diagram for Aircraft Reactor Fuel Reprocessing.

600°C
NaF ABSORBER

 

 

 

  
   
     
 
 

 

 

 

 

 

   
 

d.f. 50 F2
Ty FRR 2HF (H,)
2y X F 2
—_———— — —— COLD
e WASTE | Thip WASTE
HE Fs, UFg, F.EF, i i,
WASTE y v
§
el Fz(s) Mg
800°C Ufs, F.PF, / UFg /
NoF'2 } NoF ABSORBER 41 =104 Y b s 508 REFABRICATION
1 31 :
650°C %
FUEL DISSOLVER .
ELEMENT (HYDRO- 650°C -80°C(T0 50°C) 0=-50°C (5 ~800°C o
. FLUORINATION) FLUORINATION COLD TRAP DISTILLATION REDUCTION UoR o
' £.PF. UR,

 

NaF —Z¢F, RECYCLE

 

 

 

 

WASTE STORAGE  |me

T
ORNL-LR-DWG 4338

 

L/

 

 

 

 

 

=D

 

 

STR FUEL ELEMENT |-t

 

 

   
   
 

NoF — ZrF, WASTE
>99% F.P
<002% U

 

 

 

 

{3 ASSEMBLIES)
560¢g U ’
71,0009 Zr
2,000 g Sn
~309F.R

 

 

   
    
  
   
 

S
OVE ASSUMES NO PRIOR’ MECHANICAL SEPARATION OF U AND Zr}.

gy F2 DISPOSAL REQUIRED; RECOVERY OR RECYCLE OF F, NOT ECONOMICAL.
(5} PROCEDURE UNKNOWN ; DEVELOPMENT REQUIRED )

 

(3 ): U, +10'F, == UR; + 9F, (EXCESS 'R} REGUIRED “FOR COMPLETE UR; DISTILLATION).

 

ot—— 7 | RCALOY

139

T T

 
coa bl

 

 

 

fused salt bath to permit dissolution at a pene-
tration rate of 22 to 35 mils/hr, passage of the
L'Jl'-'fi‘«l"':2 mixture through a bed of NaF at 650°C to
obtain additional decontamination, and final puri-
fication of the UF ; by distillation,

140

Recovery of the uranium by fluoride volatilization
will be easy, although methods for obtaining a
decontamination factor of about 106 will have to be
developed in order to allow metallurgical pro-
cessing of the uranium, | |

-

&

 
Part

K2

SHIELDING RESEARCH

»

"

"
d

i

 

 

 

 

 

T

 

 

 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   

 

s
. -
<
. ¢ ;
-
-
=
»
-
=
2
<
-
=
-
. -
-
-
=
=
-
- w
-’ - *
-~
=
. N
~

 

   

 

 

 

 
 

R B SR aR . LR L e & S

 

e o

 

Ry .‘

* FRACTION' OF ENERGY TRANSMITTED
o

oot L

 

.

>

12, SHIELDING ANALYSIS

E. P. Blizard

F. H. Murray

C. D. Zerby

Physics Division
H. E. Stern
Consolldcted Vulfee Alrcroff Corporation

S. Auslender
Pratt & Whitney Aircraft

The Monte Carlo method was used for two calcu-
lations: the penetration of gamma rays through
composite slab shields and the heating in beryllium
slabs resulting from gomma rays in an adjacent
source. In addition, a three-parameter mathematical
formulation of the variation of radiation-induced
injury to pilots of nuclear-powered aircraft with
time after exposure was derived, and analyses

were made of the preliminary differential experi-

ments and of the first experiments with the GE-ANP
R-1 reactor shield mockup at the Tower Shielding
Facility (TSF).

SLANT PENETRATION OF COMPOSITE
SLAB SHIEL DS BY GAMMA RAYS

C. D. Zerby

A Monte Carlo calculation of the penetrcmon of
gamma rays through the side of a crew-compartment
shield has been completed, The study was ex-

S e UNCLASSIFIED
S - 2-01-059-10

ANGLE. OF
ENCE, 60deg

in. POLY-
ETHYLENE

3 in. POLY-
ETHYLENE

3in. POLY-
ETHYLENE

o
o

o
o

0.05

0.02
. ANGLE OF
INCIDENCE, 30deg

ANGLE OF
| INCIDENCE, Odeg

 

- \.\»Ov:-“:fhfl 0’2 we 0.4.- e emi S NS e e o ]

  

 

 

 

" Fig. 12.1. “Energy Resulting from Slant-Inei

2#m062 Photons Transmitted Through Composite

Slabs of Polyethylene Backed by Lead.

 

 

“LEAD THiCKNESS () T T

  

tended to include an investigation of the effect of
the variation of the parameters involved so that
the penetration of any complex spectrum of incident
gamma-ray flux could be determined by integration
of the resulting data, Results of a calculation with
one set of boundary conditions were given pre-
viously,!
are avo:lable.

The results are presented graphically in Figs.
12.1 and 12.2 as the fraction of incident energy
penetrating composite slabs. The exponential

 

IC. D, Zerby, ANP Quar. Prog. Rep. Sept. 10, 1954,
ORNL-1771, p 157.

2¢. b Zerby, Preliminary Report on the Penetration
of Composite Slabs by Slant Incident Gamma Radiation,

ORNL CF-54-9-120 (Sept. 21, 1954).

UNCLASSIFIED
2-01-059-9

      

1.0 POLYETHYLENE
3

POLYETHYLENE
3 in,

POLYETHYLENE
05 9

0.2
ot

205 ANGLE OF

30 deg

FRACTION OF ENERGY TRANSMITTED

   

ANGL.E OF
INCIDENCE, 60 deg

 

    
   
  
 

0o 02 04 0 02 04
-~ LEAD THICKNESS (in.

 

6-m0c2 Photons Transmitted Through Composite
Slabs of Polyethylene Backed by Lead.

143

cnd summary tabulations of the results.

12.2. Energy Resulting from Slant-Incident

T

 
i ot L R

 

 

 

 

ANP QUARTERLY PROGRESS REPORT

character of the data is evident from the straight-
line fit to the points on semilogarithm paper.

ENERGY ABSORPTION RESULTING FROM
INCIDENT GAMMA RADIATION AS A FUNCTION
' OF THICKNESS OF MATERIALS
WITH SLAB GEOMETRY

C. D. Zerby

The interaction of radiation with the atomic
particles of matter results in an energy transfer to

‘the particles or in the creation of secondary radia-

tion. The secondary radiation, in turn, is usually
absorbed close to the primary interaction. With

" high radiation intensity, the absorption of radiation

energy in biological shields or in structural members
can be important, Since essentially all the energy
absorbed becomes heat, problems of induced thermal
stresses in already stressed structural members
may well determine design limitations, The heating
effect may also be a limiting design factor in some
cases 'in which material strength is seriously de-
pendent on temperature. In all cases some knowl-
edge of the energy absorption distribution within a
material is necessary, Only a preliminary study of
heating in beryllium has been completed thus far,
and results are reported here.

Most radiation absorption problems can be at
least approximated by known analytical methods.
For the gomma radiation absorption distribution,
however, solution of the Boltzmann transport equa-
tion is necessary for accuracy. At NDA, extensive
calculations have been carried out® to solve the
transport equation for the penetration of gamma
rays, The NDA calculations included the energy
absorption as a function of position for a point
isotropic source,* but it was not possible to in-
clude the absorption distribution for plane slabs,
At ORNL, the total energy absorption of incident
gamma radiation in the different materials of finitely
thick composite slabs was also calculated,? but,
again, the calculation did not include the absorption
distribution,

In this study the Monte Carlo method is being
used for the calculation of the energy absorption

~ distribution in finitely thick slabs, and the calcu-

lctions are progrommed for solution on the ORACLE

 

3H Goldstein and J. E. Wilkins, Jr., Calculations o
the Penetration of Gamma Rays, NYO0-3075 (also ND
]5C-4'|) (June 30, 1954).

Ibzci, Tables 7.113, 7.116, 7.119, 7.122, and 7.127,

144

at ORNL, For a monodirectional beam of mono-
energetic photons on a slab of material of finite
thickness, the energy absorption was calculated
in every interval of 0.10 in, through the slab. The
absorption included the energy transmitted to the
material in a scattering collision as well as that
transmitted in an absorption collision; in every
case the photon was followed until it was reflected
from the slab, was absorbed, or had penetrated the
slab. The scattering cross sections were calcy-
lated with the use of the Klein-Nishina scattering-
cross-section formula,® while the absorption cross
sections were obtained by empirical fits to published
data,®

A 9-in.-thick slab of beryllium was used, and the
incident photons were considered to have energies
of 0.5, 1.5, and 5.0 m0c2 and to be normally incident.
For the 1.5-m0c2 case, angles of incidence of 30,
60, and 85 deg were also investigated (1 m062
= 0.51 Mev).

The results obtained for the normally incident
0.5-m0c2 photons on the slab of beryllium are shown
in Fig. 12.3, The histogram indicates the statistical
nature of the Monte Carlo solution, while the smooth
curve is a fit, by eye, to the data, The values
given on the histogram are the averages of the
values for the 0.10-in, intervals. In the future it is
intended that a rigorous method of fitting the data
will be employed. The area under the curve is the
total energy absorption.

 

Sw. Heitler, The Quantum Theory of Radiation, 3d ed.,
Eq.39, p 219, Oxford University Press, New York, 1954,

6. R. White, X-Ray Attenuation Coefficients from 10
Kev to 100 Mev, Table |, NBS-]OOS (May 13, 1952).

UNCLASSIFIED

 

2-01—059—13
. 0.2(; &=
¢ O REFLECTED 0187
=% 06 ABSORBED  0.726
L= o PENETRATING ~ 0.087
x

B o2 STATISTICAL ERROR % yli=d)
=g 00 J/=FRACTION
Lch 008
zQ 006
£ < 004
g 002

o

o 1 2 3 4 5 & T 8 9 10

THICKNESS (in.)

Fig. 12.3. Radiation Heating of a 9-in.-Thick
Beryllium Slab by Normally Incident 0.5--mnc2
Photons. Smooth fit to statistical data shown.
 

 

L

oy

s

u) .

i\

'_ Vthrs is a smooth fit, by eye, to the statistical data.

 crews of nuclear«powered mrcroff some
: should be taken of the recovery from ‘radiation-

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
- . ) . 2—01-059-15
S 030 — : ——— : - ,
= 028 'Npi’é’%“J FRACTION OF ENERGY -
B 0.26 | —{ENERGY (mgc?)| REFLECTED | ABSORBED PENETRATING ||
Ro2a— 05 0.187 0726 0.087 ||
g 022 — 15 0.065 0.763 o172z |
%.;3 02004 50’ 0012 0595 0393 ||
018 Lt A JTO=F —
% g STATIST!CAL ERROR VL) s
® 016 J/=FRACTION
Z 014 >§\ |
— \< E,,/O.Smoc
5 012 e
2 010 \‘\mjc
Z 008 -
% 006 [— z,>“~—-\:\\
O ' 5.0 moc . “T—.é
% 004 [ - ]
= \
5 002
<[
e 0 :
0 1 2 3 4 5 6 7 8 9
THICKNESS (in.)
Fig. 12.4. Radiation Heating of a 9-in,-Thick

Beryllium Slab by Normally Incident Photons.

In- Fig. 12.4 the smooth fits to the statistical
dota are presented for normally incident photons of
energy 0.5, 1.5, and 5.0 m, c2, The effect of the
reflected photons on the energy absorption curve
is readily seen at the entrance face of the slab.
The value of the fraction of energy absorbed per
mch at the entrance face for the incident photons
can be calculated by multiplying e the scattering
cross section, by f, the average fraction of energy
absorbed by the material.” These values for 0.5,
1.5, and 5.0 myc? are, respectively, 0.118, 0.123,
and 0.089. The large contribution of the reflected
component is especially evident for the 0.5-mqc?
photons at the entrance face as the difference be-

tween 0.118 and the frccnon 0. 175 ’rcken from Flg.

 
 

12.4. ‘T‘he effec

 

 

In “the Specn‘ncqnon of the tolerance dose for

 

 

 

" The fcctor fwas obfamed as 1 minus the value given
in a table by U. Fano, Nucleonics 11, Table 3, p 11
" (Aug. 1953).

 

PERIOD ENDING DECEMBER 10, 1954

 

" UNCLASSIFIED
2—=01—059-14

% 060 i
£ 056 FRACTION OF ENERGY
i 052 ’ PENETRATING
a 0.48 0.065 Q763 oi7e
D 044 0.079 0.785 0136
@ 040 0.135 0.813 0.052
< 436 0.277 0.704 0.019
;‘5 0.32 STATISTICAL ERROR + ./(1 J)
W J=
Z 028
= -'deg ANGLE OF
£ 024 -
& 020
QO
Z ole
& 012
g 008
=
G 004
E 0

0 { 2 3 4 5 6 7 8 9

THICKNESS (in.)

Fig. 12.5. Radiation Heating of a 9-in.-Thick
Beryllium Slab by Slant-Incident '|.5-moc2 Fhotons.

induced injury with time. Just recently the ANP
Medical Advisory Group (ANP-MAG) issued recom-
mendations® which made some concessions in this
direction in that increased total doses were allowed
for schedules which distributed the radiation over
longer times. It has been demonstrated® that the
ANP-MAG recommendations can be expressed to
within 4% by a mathematical formulation with three
parameters based on a model of partly irreparable
injury and partly injury which recovers exponen-
tially with time, The model was previously sug-
gested by Blair.'=12 The mathematical formula-
tion is much more flexible than the Biological

Planning Chart which expresses the ANP-MAG

"'recommendahons since it glves a umque value of

 

    
 
 
  
 
  
 
  

“Radiation, ORNL CF-54-9-119 (S

  
 
 

to the Guznea Pzg, Rat and Dog, _U R—207 (July

phofons is rshO.Wn in Fig.” 12,5 Aga;n,“""""’ R

8“Mlnu'res of the 3rd ANP-MAG Meehng — 11-12 May

) ]954 Scbool of szatzon Medzcme, 5-18

 

4.9E P thord Tbe sze Varzatzon {or In,'ury /rom
epf.2] 954)

   

]OH A Blctr A Formulatzon of the In]ury, sze Span,

i Dose Relations Féo:v' lonizing Radiations — I Applzcatzon
oLte tbe Mouse, U

206 (May 13, 1952).

H, A Blcur, A Formulatzon of tbe In]ury, sze Spcm,

Dose Relations for lonizing Radiations — 11, g %zggzt;on

A 2H A. Blmr, Recovery from Rczdzatzon In]ury in Mice
and Its Effect on LDSO for Durations of Exposure up to
Several Weeks, UR-312 (Feb. 10, 1954).

145
:f’."'n "“.QM 2
Yo i 148

 

 
where

 

 

 

ANP QUARTERLY PROGRESS REPORT

injury for any radiation schedule no matter how
irregular, The expression is

t

- dD -
I(t) = [fi 1.08 x 1073
o dt”’

Jo & .

+1.05 x 1072 e°°'713(f"‘")}dt” ,

I(t) = injury a;r' time ¢ in units of maximum safe
. injury (MSI) which is just that allowed for

©" . training by the ANP-MAG (z 2 t’),

a7 | '

2

[

. — = dose rate, rem per unit time, at time :”
dt (0<t"<t’)._l

It may be noted that the half life of the reparable

. injury is about one year and that this component
~ appears to constitute 90% of the total.

Other formulations are suggested to take account

- of other phenomena, such as enhancement of sensi-

tivity by previous doses. Nothing basic to the
biology of the problem is intended; rather, it is
suggested that if the restrictive planning chart is
satisfactory for airplane design and strategic plan-
ning, then the formulation should also be satis-

~ factory from a biological point of view, Since it

is much more flexible than the chart, it would be
much more useful for interpretation of actual radia-
tion schedules in terms of injury.

This same model predicts that the AEC laboratory
tolerance doses administered over a 30-year period
give the same injury as does an instantaneous
dose of about 60 rem. These numbers are not in-
consistent with presently accepted concepts.

ANALYSIS OF‘SOME PRELIMINARY
DIFFERENTIAL EXPERIMENTS

M. F. Valerino!3

In order to investigate the adequacy of simple

analytical models proposed for describing some of

the neutron transport processes involved in the

aircraft divided-shield concept, some of the data

obtained in the preliminary differential experi-
ments'? at the TSF have been examined. Only

‘thermal-neutron measurements have been made.

 

]3_On loan to Tower Shielding Facility from National
Advisory Committee for Aeronautics, Cleveland, Ohie.

- 14¢. E. Clifford et al,, Preliminary Study of Fast Neu-
tron Ground and Air Scattering at the Tower Shielding
Facility, ORNL CF-54-8-95 (Aug. 23, 1954).

146

The thermal-neutron flux measured at the rear of
the detector tank as o function of the angle @ (angle
between the reactor axis of symmetry and the
source-detector axis) for a 85-ft reactor-detector
separation distance and a 195-ft altitude is shown
in Fig. 12.6. The flux is plotted relative to that
for 6 = 0 deg and is compared with a curve of the
direct beam flux to be expected at the detector, as
calculated on the basis of the total water thickness
between the reactor surface and the detector. In
the calculation of the direct beam flux the leakage
neutrons were assumed to have a cosine distribu-
tion at the reactor surface, and the attenuation
lengths were based on Lid Tank Shielding Facility
(LTSF) data corrected to give the point-to-point
material attenuation kernel. The calculated curve
agrees well with the experimental curve for angles
up to 45 deg. Beyond the 45-deg angle, the calcu-
lated flux drops off rapidly, and at 90 deg the direct
beam is gone. The measured flux for 6 = 90 deg
must hence be presumed to be due to air-scattered
neutrons only. Single-scatter theory calculations
have not yet been performed for the case of a beam
whose axis of symmetry is at an angle with the
source-detector axis, and hence it was not possible
to calculate the magnitude of the air scattering to
be expected.

In order to estimate the air scattering into the
rear of the detector tank, the air-scattered flux

 

 

 

< REAR\\L_?_‘
0.5 N %D:RECT SIDE %

X \__ - SCATTERED DETECTOR
Ao\ REACTOR , TANK

G2 NS, TANK
\\k
. N\
0.4 -

= MEASURED DIRECT PLUS =]
X SCATTERED INTO REAR ]

 

 

AR
2-04~-056-2-T7
P
=
¢

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

= |GALCULATED DIRECT PLUS \ IS\
L|® SCATTERED INTO REAR o/ \
L 0.02 HSCATTERING BASED ON /‘i

 

 

MEASUREMENTS AT SIDE)

|

 

0.04

 

 

 

I
T
[
!

 

 

 

 

0.005 CALClJLATE) DIRECT BEAM —

 

\ MEASURED SCATTERED |
\ FLUX INTO SIDE A

 

 

0.002

0.004 \ '
0

15 30 45 80 75 L 105 120 135 150
g {deq)

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 12.6. Comparison of Calculated and Meas-
ured Fluxes in Rear of Detector Tank.

¥u

 
 

£

  
  
  
 
 
   
  
    

 

 

2

 

measurements taken at the side of the detector tank
were considered. For 6 = 90 deg, the experiments
indicate that the scafiermg into the side and that
into the rear of the detector tank are approximately

‘equal, If it is assumed that this is nearly true for

angles between 45 and 90 deg, the air scattering
into the rear can be estimated from the measure-
ments at the side. When the olr-scaflered neutron
flux estimated in this manner is added to the
calculated direct beam flux, the dashed curve in
Fig. 12.6 is obtained. The fair agreement between
the dashed curve and the measured curve indicates

that the simple model used to calculqte the angle’

of distribution of the leakage neutrons from the
shield sutface may be adequate for divided-shield
design. |

“Figure 12.7 represents an attempt to compare the
measurements of the air scattering into the side of
the detector tunk wnh ‘the air scattering indicated
by single-scatter calculations. The ordinate in
Fig. 12.7 is the ratio of the flux at the side surface
of the detector tank for a given beam direction @ to
the flux at the rear of the detector tank obtained

» .
2-04-056-2-T10

 

 

 

 

 

 

 

 

 

 

   

 
 

o

ING EXPER!MENT AT

 

040

 

e 0,08

 

 

3806

 

 

 

 

 

 

“Fig. 12.7. Ratio of Extrapolated Surface Flu

 

S_cqflelf Caleulation.

 

  
 
  

 

  

S
as lnd:cuted by Measurements and by a Single-

PERIOD ENDING DECEMBER 10, 1954

when the beam is pointed directly towards the
detector tank, that is, for 8 = 0 deg. Hence, the
ordinate is a measure of the fraction of the beam
scattered into the sides of the detector tank. The
abscissa is the beam direction 8§, The points indi-
cated by the experiments for 6 = 30 and 90 deg were
obtained by extrapolation of the fluxes measured
in the water to give the extrapolated surface.values.
The measured relaxation lengths were used in the
extrapolation; no relaxation length measurements
were made for 6 = 0 deg, and therefore this pro-
cedure could not be used for this angle. The only
single-scatter calculation presently available is for
6 =0 deg, and it is based on the assumption of a
2r-isotropic detector, that is, an isotropic detector
with one hemisphere shielded. This calculation
gives a value of 0.048 for 6 = 0 deg, which is to be
compared with the values indicated by the experi-

“ments of 0.063 for 6 = 30 deg and 0.048 for 0 = 90

deg. As the beam is rotated from 30 to 0 deg, it is
nearly cut in half insofar as scattering into the one
side of the detector tank is concerned, and this
should result in a decrease of the scattering into

- the one side of the detector tank. The scattered

flux measurements taken within the water show
this to be the case, The point calculated by
single-scatter theory for § = 0 deg appears to agree
with the points obtained from the experiment for

= 30 and € = 90 deg. However, it should be
cautioned that the fair agreement indicated in the

figure could be fortuitous. In the comparison, the

assumption is involved that the thermal flux re-
sulting from a given fast-neutron current into the
side of the detector tank is the same for air-scattered

_neutrons as for direct-beam neutrons. Also, some
‘untertalnhes are mfroduced in the ex?rapofa‘hon of
;fhe thermal flux to obtaln the surfoce value..

INTERPRETATION OF A!R AND GROUND
' SCATTERING AT THE
TOWER SHIELDING FACILITY

J. Van Hoorriis'sen] 5

Measurements of neutrons scaflered in the side
of the detector fank asa funchon of reacfor-detector

- altitude for two dn‘ferem‘ source geome'rr:es have

) esfcbl:sh the fact

been performed at the TSF
per;ments mcde at

that neutron air-scattering ¢

     

  
   

 

 

 

150n loan to Tower Shielding Focility from Boeing
Airplane Company.

147

 

 

 
ANP QUARTERLY PROGRESS REPORT

ot 1-01-056-2-512R

 

10° L

Q
™

 

fos ft

64 ft
45.2 cm
VARIABLE
0 deg

RELATIVE THERMAL NEUTRON FLUXV(ARBITRARY SCALE)
. o

CENTER OF TRIPLET BFy CHAMBER AT
¥ 0.0cm
y 14.0cm
zZ 0.0cm

1072
360 330 300 270 240 210 180

8, HORIZONTAL ANGLE BETWEEN p AND & AXES (deg)

 

‘Fig. 12.8. Ditect-Beam Radiation as a Function of Angle; Preliminary Differential Experiments.

148
{“ {""a i I
| o2 15

N

ks
"

 

. o Sk S )

 

[

<

"

¥

u™

iy

@. :

w

e

W

 

200 ft are relatively free from a background of
ground-scattered neutrons.

In each experiment the neutron flux reaching a
BF, detector placed inside the detector tank and
near a side wall was measured as the altitudes of
the source and the detector were varied simul-
taneously from the ground to 200 ft. The only
major difference between the two experiments was
the shielding placed around the TSF reactor. In
the first case the reactor was placed in the reactor
tank at a distance (p) of 45 cm from the tank wall
and a reactor angle (6) of 330 deg from the source-
detector axis;'? in the second case the reactor
was placed in a mockup of the GE-ANP R-1 shield
design (see ‘‘TSF Experiment with the Mockup of
the GE-ANP R-1 Shield Design,”’ in Sec. 15).

In order to estimate the actual ground-scattered
contribution at the 200-ft altitude, attempts were
made to analytically separate the experimental
measurements into their air- and ground-scattered
components. A calculation'® of the number of
neutrons singly scattered into a point isotropic
detector as a function of the solid angle in which
they had left the source had been made previously
for the case of infinite height. Also, a calculation'’
of the number of air-scattered neutrons lost because
of the presence of the ground and the number of
ground-scattered neutrons reaching the detector as
a function of source solid angle and altitude had
been made. All three calculations assumed single
isotropic scattering, air attenuation, and a separation
distance of 64 ft.!8

The integration over the source solid angle to
obtain curves of air and ground scattering as a

function of altitude was welghted in each case by
the angu!ar ‘distribution of neutrons emitted from
In ’rhe dlfferenfml type of experument_

the sh:eld

(reactor in reac#or fonk) fhe angulor dlsfrnbuhon

the d:rec’r\ beam” reqchmg ‘the detec':for"r-ank as a

  

 

tering in the Presence of the Ground, D-14624 (July 8,
]954§-

‘SThe measurements with the GE-ANP R-1 mockup

were actually for a 70.8-ft separation distance.

 

   

- water-filled reactor tank.
for this confribution in the GE- ANP R-1 sh:e]d
j?""/'mockup expenmem‘ (Flg. 12 ]]) was 5%,

WM D Peurson, Calculatzon of Smgle Neutron Scat;'

PERIOD ENDING DECEMBER 10, 1954

GEefeT
2-01—056—3-T 20

 

 

 

 

RELATIVE FAST-NEUTRON FLUX (arbitrary units)

Vg \\/ | \

0 20 40 60 80 100 {20 140 180 180
8, ANGLE WITH RESPECT TO REACTOR—CREW SHIELD AXIS {deg)

 

 

 

 

 

 

 

 

 

 

 

o

Fig. 12.9. Angular Distribution of Fast Neutrons
Emitted from GE-ANP R-1 Shield Mockup.

scattered contribution to the data at 200 ft was
negligible, and the single air-scattering curve as a
function of altitude was normalized at this point.
This normalized curve was then used to subtract
the air-scattered contribution from the experimental
measurements at lower altitudes, and thus a series
of experimental ground-scattering points was ob-
tained. A fit of the calculated ground-scattering

- curve through these points gave an estimate of the

ground-scattered contribution to the experimental
data at the high altitude. The air-scattering was

-.-- then renormalized, with this d-scattered con-
(F:g. 12.8) was obtained by the measuremenf of' on renormalizec, with this ground-scattered con
: : -~ tribution taken into account.

'f 3 f . k] .
Gng]e 9 The cor espondmg iteration, of course, quickly converges,

This method of

The ancslysns (Fig. 12. 10) for the differential
type of experiment resulted in an eshmate of 1% for
the ground-scattered contribution at 200 ft with the
The estimate cbtained

 

 

4. E Hungerford, Bulk Slazeldmg Facility Tests on
the GE-ANP R-1 Divided Shield Mockup, ORNL CF-54-
8-94 (to be issued); see also ANP Quar. Prog. Rep.
Mar. 10, 1954, ORNL-1692, p 124.

149

 

 

P

 
RELATIVE THERMAL-NEUTRON FLUX (arbitrary units}

 

ANP QUARTERLY PROGRESS REPORT

In conclusion it can be said that at the 200-ft
altitude the ground-scattered contributions to air-

‘'scattering experiments at the TSF are dependent

ASEORE
2-01-056-2-T 24

EXPERIMENTAL

CALCULATED SINGLE AIR
SCATTERING .

CALCULATED SINGLE

GROUND SCATTERING

 

a0 80 120 160 200
A, REACTOR-DETECTOR ALTITUDE (ft)

Fig. 12.10. Cohpurison of Scattered-Neutron
Calculations with Preliminary Differential Experi-
ment,

150

on the source geometry; however, the contributions
will be small, that is, not more than 5%, in all
cases. - '

- Crone
2-01—-056—3—-T22

S

o

™

—_

AIR-SCATTERING

o
o

o
o

CALCULATED SINGLE GROUND—

o

RELATIVE THERMAL NEUTRON FLUX (arbitrary units)
o
Q
a

Q
O
n

 

o
Q

20 40 &0 80 {00 {20 {140 {60 180 200
4, REACTOR-DETECTOR ALTITUDE (ft)

Fig. 12.11. Comparison of Scattered-Neuvtron
Calculations with GE-ANP R-1 Mockup Experiment,

 

 

<

 

 
 

 

« . \ | G. T.

PERIOD ENDING DECEMBER 10, 1954

13. LID TANK SHIELDING FACILITY

¥ Chapman
z J. M. Miller
i _ _ : ‘ Physics Division
o 7_ ' J. B. Dee H. C. Woodsum
: o -~ W. H. McCool

Pratt & Whitney Aircraft

: . A measurement of the removal cross section of
lithium was made at the Lid Tank Shielding Fa-
cility (LTSF), and the experimentation on the

. GE-ANP helical air ducts was continued. Prepa-
rations for the cnrcu!ahng fuel reflector-moderated
reactor and shleld mockup tests are nearing com-
pletion.

EFFECTIVE NEUTRON REMOVAL CROSS
' SECTION OF LITHIUM

, G T. Chapman
J. M. Miller C. L. Storrs!

An effective neufron ‘removal cross section (0p)

= of 1.01 + 0.04 barns was determined for lithium at

the LTSF. This vclue was obtained from measure-

. ments made in c ‘medium of oil behind a slab of

= ' metalllc hthlum whlch was ploced against the
- source plate.

o The slab of h’rh:um (p = 0.53 g/cm ) was con-

" structed by filling a 122-cm-dia by 29.1-cm stain-

less steel tank with melted lithium and allowing

the material to sohdn‘y The walls of the tank

were remforced with o framework constructed of

- 6-in, lbeams to assure fhof they would remain
_parallel durmg the poqrmg “and sohd:fymg proc-
Figure 13, 'l isa schemahc d:agram of the: .

: __esses. _
hfhlum fIHed tank.

 

 

ieh g

 

= . use with future experiments,

 

= _]Generul Electric Company.

 

a»|y5|s of the I:thmm '"dlccted thdf.};v::;k

  
      

ank des:gned. for
-WhICh has a I/ -in.-

Inconel window in the wall adjacent to the source.
The presence of the Inconel resulted in ¢ 9-Mev
capture gamma ray,? and the photoneutron contri-
bution from this high-energy gamma and the
naturally occurring deuterium in the oil markedly
affected the thermal-neutron measurements in oil
(no lithium present) for distances greater than
120 cm from the source (Fig., 13.2). A 3.81-cm-
thick bismuth slab placed in the oil approximately
45 ¢m from the source suppressed this gamma ray
and the resulting photoneutrons without any ob-
servable effect on any other component of the
“thermal-neutron flux.

The lithium tank was then placed in the oil tank
adjacent to the source, and thermal-neutron flux

 

2), B. Dee, D. K. Trubey, and W. Steyert, ANP Quar.
Prog. Rep. Sept. 10, 1954, ORNL-1771, p 164.

SESRE™
2-01-057-65-27
APPROXIMATE LEVEL ‘
OF LITHIUM

 

 

 

 

 

Sl

=
]
H

0.95-cm STAINLESS STEEL WAL
I
C
|

 

 

 

 

|
9
%

 

 

 

 

 

 

 

 

t
|
|
|
|
L —

 

 

 

 

 

“ ‘-:.':7/3_,";; \LID TANK SOURCE ) r=—29.43 cm—=

FRONT wiEw

 
 
  

 

o 4S'I-D<bE”\.{l_EW )

Fig. "13:;11“.’ Experimental Arrangem;;;ii;i;1fh{e
LTSF Lithium Removal Cross-Section Measure-
ment,

 

   

: F i e

Lh4 151

 

 

 
 

 

 

ANP QUARTERLY PROGRESS REPORT

SESRET
2-01-057-65-28

NORMAL LTSF WATER CURVE

OIL, NO BISMUTH GAMMA SUPPRESSOR

- THERMAL-NEUTRON, FLUX (nv,)

NORMAL LTSF WATER CURVE

QIL, WITH BISMUTH GAMMA
\

 

50 60 70 80 90 0O HO 120 130 140 150 160 170

2z, DISTANCE FROM SOURCE (cm)

Fig. 13.2.
and Without Bismuth Sleb to Suppress Photoneutron
Production.

measurements were made in the oil behind the
29.1-cm thickness of the lithium (Fig. 13.3). From
these measurements the value of 1.01 * 0.04 barns
was calculated for the removal cross section of
lithium by the procedure presented below.

All previous removal-cross-section measurements
were made in water. In order to verify that a
correction to a water medium would not be neces-

- sary for these measurements,3 thermal-neutron
- measurements were also made in oil behind a
“material which had a op predetermined in water.

 

..~ The definition of a removal cross section has been
. given as a parameter characterizing the attenuation

of fission neutrons by materials either mixed with

hydrogen or followed by a hydrogenous layer; see H,

Goldstein and R. Aronson, ‘‘Effective Removal Cross
Sections~Theory,’’ abstracted in Symposium on Fission
Physics and Classified Nuclear Physics for Reactors
-and Shielding, Oct, 13-—15, 1954, ORNL CF-54-10-11,

. p 100 (Oct, 8, 1954).

 

152

Thermal-Neutron Flux fn Oil With |

- WEORDH
2~01-057-65-29

EXPERIMENTAL CONFIGURATION

cm
10 WATER
W INCONEL
25 Ol
“CALCULATED VALUE ~

 

THERMAL-NEUTRON FLUX (v, )

-4
0 LITHIUM

2 |.. oL OnLY

OIL ONLY
\ {WITH BISMUTH GAMMA SUPPRESSOR}

 

102
50 60 70 80 90 100 0 20 30 {40 150 {60 {70

z, DISTANCE FROM SOURCE {cm)

Fig. 13.3. Thermal-Nébtron Flux Beyond Lithi-

vm in Oil.

The material chosen was bismuth, and the measure-
ments in oil behind a 22.9-cm-thick slab ‘are
plotted in Fig. 13.4. From these measurements
a value of o, = 3.48 barns was calculated, which
can be compared with a value of 3.49 + 0.35 barns
previously determined for bismuth from measure-
ments in water.4 Gamma-ray dose measurements
made throughout the experiment are plotted in
Fig. 13.5. _ _

The removal cross section of lithium was calcu-
lated by a method suggested by Blizard:®

 

4C. L. Storrs, G. T. Chapman, and E. P. Blizard,
‘*Effective  Removal Cross Sections—Experiments,’’

ORNL CF-54-10-11, op. cit., p 101 (Oct. 8, 1954).

SE. P, Blizard, Procedure for Obtaining Effective
Removal Cross Sections from Lid Tank Data, ORNL
CF-54-6-164 (June 22, 1954).

5

3627 '5R

-l

‘o

*a

»

"
)

e

v
 

i

o

ity

i

o

 

 

vail(z) Aolz) z + 172
DM(z + t) - )\M(zfl ; t)[ z }

,
a’t
_ L Y

 Adale + 1) - a’z

7

where _
Doil(z) = dose in oil at z cm from the
: source,
DM(z + t) = dose in oil behind ¢ cm thickness
~ - .+ of shield and z cm of oil,

A (=) = relaxation length of oil at z cm,

Aylz + 1) = relaxation length behind the shield
and in oil at (z + ¢) cm,
T SpereT
3 2-04-057-65-30

 
  
   

1,90 24.8% em

BISMUTH
9.80 g/cm3

0
- EXPERIMENTAL, CONFIGURATION

BISMUTH

OIL ONLY

THERMAL-NEUTRON FLUX (nvyy)

 

PERIOD ENDING DECEMBER 10, 1954

A = average of A (z) and )\M(z +t)
a = source radius = 35.56 cm,

2t = (21.‘)Li + (Et)ss, the sum of macro-
scopic removal cross sections of
sample and container, where the
subscript ‘‘ss’’ refers to the stain-
less steel walls of the lithium
tank.

Stainless steel was assumed to be the same as
iron (o, = 1.98 barns/atom). The numerical values
for these terms are given in Table 13.1.
GE-ANP HELICAL AIR DUCT
EXPERIMENTATION
J. M. Miller

The GE-ANP helical air duct experimentation

“at the LTSF has continued with thermal-neutron
flux measurements beyond a 35-duct array. As

described previously,® the ducts are 3-in.-ID steel

 

Fig. 13.4. il"hermai-rfleu}ronu-Frlrux -Beyo:rici. Blsmuth |
in Oil, . -

6J. M. Miller, ANP Quar. Prog. Rep. Sept. 10, 1954,
ORNL-1771, p 166.

SEURET
2—-01-087-65-31

H,0 CURVE IN TANK WiTH
INCONEL WINDOW -

OIL CURVE IN TANK WITH
INCONEL. WINDOW

OIL CURVE BEHIND
LITHIUM

  

LiD TANK H,C GURVE

GAMMA-RAY DOSE RATE (mr/hr)

   

| 0l CURVE BEHIND BIS

     
  

C 80 'gq ) 100~ 120 140 160
- 2z, DISTANCE FROM SOURCE fem) =~

 

Ff'i;g. 13.5. Glamm-ak-kfi:y:Dc;'setRaté in G:if Beyond
Lithium and Bismuth,

i ot 153

 

 

T T Iy, R T 1o PRy T i m

 

 
 

 

 

 

.

ANP QUARTERLY PROGRESS REPORT

conduit shaped around a 9-in. core. After removal
of the core, the ducts were stiffened by Fiberglas
wrapping. The projected length of each duct along
the z axis was 46.5 in., and the duct arrays were
arranged so that there was 5 in. between the duct
center lines.

Thermal-neutron measurements beyond a single
duct and a 3-duct array in a water medium were
reported previously.® Measurements made beyond
the 35-duct array in both a water medium and a
- gamma-shield medium are shown in Fig. 13.6. The
- gamma shield consisted of steel Raschig rings
- (35 vol %) and water borated to 1% by volume.
For both mediums the array was contained in a
4-ft-long iron tank (S/S-in.-fhick walls) which had
z-in.-thick iron windows in each end. The
windows were 29 in. in diameter and were slightly

| SapREm
2—-01-057-63-34

35 DUCTS IN PLAIN H,0, z=132

35 DUCTS IN MEDIUM OF RASCHIG
RINGS AND BORATED , Z=135¢em

3 DUCTS IN PLAIN H,0, z =130

Ho0, 22130 cm

THERMAL —NEUTRON FLUX (nvy,}

PLAIN H,0 ONLY, 7=130 cm

5 RINGS IN BORATED Ho0, 2=132 cm

 

102l
© —80 —50 —40 —%0 —-20 -0 O {0 20 30 40 50 60
’ ¥, DISTANCE FROM CENTERLINE OF SOURCE {cm)

Fig. 13.6. Thermal-Neutron Flux Beyond Various
Arrays of GE-ANP Helical Air Ducts (3 in. in
- Diameter, 46.5 in. Long). '

154

larger than, and concentric with, the LTSF source
plate. The measurements were made behind the
tank in the LTSF water. '

As indicated in Fig. 13.6, the presence of the
35 ducts in the gamma shield increased the
thermal-neutron flux by a factor of approximately
3000. In the medium of plain water, the flux was
increased by a factor of 300, For the water
medium the increase in flux seems to be consistent
with a calculation which takes into account only
the reduced density of the attenuating medium.
For the case in which the ducts were in the gamma
shield, the neutron streaming in the ducts appears
to be relatively more important, The latter is not
surprising, since the addition of the steel in-
creases the attenuation of the medium to neutrons
as well as to gamma rays.

The fast-neutron dose rates beyond the 35-duct
array both in @ medium of water and in the Raschig
ring—borated water medium are shown in Fig. 13.7.
The gamma-ray dose rate beyond the gamma shield
without the ducts is compared in Fig. 13.8 with
that beyond the shield with the 35 ducts, The
presence of the ducts increased the gamma dose
rate by a factor of approximately 160.

SECREF
2-04-057-63-33

DUCTS IN PLAIN H50, z =133 cm

DUCTS IN MEDIUM OF RASCHIG
RINGS AND BORATED z=132 cm

FAST-NEUTRON DOSE RATE {mrep/hr}

 

1072
-80 -50 -40 -30 -20 —10 O 10 20 30 40 50
y, DISTANCE FROM CENTERLINE OF SOURCE (&m)

Fig. 13.7. Fast-Neutron Dose Rate Beyond 35

GE-ANP Helical Air Ducts (3 in. in Diameter,
46.5 in. Long). |

3
PRy
aat bon ¥

< A, F

[

ia"?? S :
w3
3

e

S
2

é}
 

W)

’“‘1‘1'._'" v_\‘ .

 

Hehc“l Kir DUC*S; (37in, in Dlameter, "46.5 in.

PERIOD ENDING DECEMBER 10, 1954

TABLE 13.1. NUMERICAL VALUES OF TERMS IN REMOVAL-CROSS-SECTION CALCULATIONS

 

 

Term Value in Value in Value in
Bismuth Lifhium Stainless Steel
t, cm 22.856 27.23 1.91 (total)
2, em 127.1 120.9
z + &, em ' 150 150
D_.\@), nv,, 7.20 x 107! 1.44 x 100
D,z + 1), nv,, 6.20 x 10~ 230 x 10~
A(z), om 9.17 9.00
Mgz + 1) em 9.90 10.1
A em | 9.54 9.55
a, em | 35,56 35.56
p, g/cm® | 9.80 ) 0.53 7.8
A, Avogadro’s number 0.603 x 1024 0.603 x 102 0.603 x 1024
M, atomic mass 209 6.94 55.85

 

REFLECTOR-MODERATED REACTOR AND
SHIELD MOCKUP TESTS

J. B. Dee

It was previously reported? that the radiation
decomposition in samples of a solution of UF
in C,F, . was found to be too great for use of the
l:qu:d as the simulated fuel in the LTSF mockups
of the circulating-fuel reflector-moderated reactor,
[t is now planned to use solid UO, in the form of
enriched uranium~aluminum MTR-type fuel plates.
A mechanical system has been developed so that
a continuous series of plates can be mounted on
a sprocket-driven chain-link belt. The uranium

 
 
 
  
 
 
   
  
        
 
  
 
  
      
     
  

  

DUCTS IN MEDIUM OF RASCHIG
RINGS AND BORATED H 0

  

will be moved on this belt from the neutron
window, where it will be irradiated, to the heat
exchanger region, and back to the window. The
speed of the belt can be varied so that a complete
transit will be made in o minimum of % sec and
a maximum of 4 sec. The system will be tested
under operating conditions in a special rig outside
the LL.TSF. Other preparations for the tests are
~ being completed.

I BORATED ;0

" GAMMA DOSE RATE {mr/hr)

ASCHIG RING:

  

  

       

 

"CENTERLINE OF SOURCE (e

    

+Fig. 13,8, Gamma Dose Rate _Beyond 35 GE-ANP

Long) in Raschig Ring—Borated Water Medium;
= 136 cm.

A 155

w X (” if\‘

Doy dh,

177
;:‘3:)

   

 
B

- ANP QUARTERLY PROGRESS REPORT

14. BULK SHIELDING-'FAC_ILITY _

F. C. Maienschein’

. D. Flynn
M. P. Haydon'!
K. M. Henry

E. B. Johnson
T. A. Love

R. W. Peelle

W. Zobel

Physics Division

For much of the time during the last quarter the

Bulk Shielding Facility (BSF) was shut down for

'r'n'_od\ifikcqflons to the reactor pool. The modifi-

cations were being made to permit the use of
demineralized water in the reactor pool in order
to eliminate corrosion of the reactor fuel elements.
The set of high-power fuel elements which had

" previously been operating in process water that

had been treated by chromate addition had to be
discarded, even though burnup of the uranium in

"’rhese elements was negligible, because of the
" extent of pitting of the aluminum cladding. In

order to utilize demineralized water, the pool was
cleaned and painted, and all steel was either
removed or sprayed with a stainless steel pro-
tective layer. A bypass demineralizer is being
installed to maintain the water purity. The
measurement reported below was made by members

. of the BSF stcff in other facilities.

A SEARCH FOR SHORT-HALF-LIFE NUCLEAR
ISOMERS IN X379, Rb®%, RLE7, AND Z:7°

" R. W. Peelle
E. C. Campbell F. C. Maienschein

Rubidium, potassium, and zirconium are among
materials which have been considered as fuel
components for reactor systems in which the fuel
is circulated. Any short-period isomeric gamma-
ray fransitions would complicate the shielding of
the external fuel loop, especially if the gamma
rays were of high energy. Rubidium was previ-

‘ously known to possess a low-energy isomeric
'rfrcmsmon with a half life of about 1 min, but
_'measurements had not been made which would
" reveal any shorter period that might be present.

Elecfromagnetically enriched K37, Rb®3, and
Rb37 in the form of nitrates were injected into

‘the ORNL Graphite Reactor by use of a fast

 

 

 

1 Part time.

156

pneumatic probe assembly designed by E. C.
Campbell. Since the samples were ejected from
the cenfer of the reactor within about 0.1 sec from
the end of the bombardment, it was possible to
moke a search for very short-period activities.
The ejected sample was viewed by a thallium-
activated sodium iodide scintillation-crystal
gamma-ray detector. The signal from the photo-
multiplier was transmitted by coaxial cables to
a multichannel differential pulse-height analyzer
located at the BSF. In addition, the counting rate
of any part of the spectrum could be measured as
a function of time with a linear count-rate meter
and a very fast Brown recorder, Thus both
spectral and time decay analyses of gamma radi-
ation emitted by the irradiated sample were ob-
tainable.

A study of K37 yielded no short-period activity.
In the rubidium isotopes the previously known
radiations were found, but no shorter period
isomers were discovered. Isomers with an ap-
preciable energy and cross section would have
been detected if they had had a half life greater
than about 0.03 sec. A sample of Rb®> was found
to contain the known 550 + é-kev transition, with
a 1.05 * 0.05-min period, and bombardment of
Rb87 produced the known Rb3® activity.

The apparatus described above was also used
in the discovery of a number of new isomeric
transitions with half periods? between 0.05 and
5 sec. One of these is Zr?%”, which yields a
2.30 + 0.03-Mev gamma ray with a half life of
about 0.83 * 0.03 sec. This isomer is produced
by inelastic neutron scattering.>

 

2E, C. Campbell, R, W. Peelle, and F. C. Maienschein,
Phbys. Semiann. Prog. Rep. Sept. 10, 1954 (unclassified),
ORNL-1798, p 36.

3, c. Campbell, R. W. Peelle, F. C. Maienschein,
and P, H, Stelson, Decay and Fast-Neutron Excitation

of ngo , papet to be presented at the meeting of the
American Physical Society, New York, January 1955,

m i ""‘ e 7,; |
i _@ ,;' j 59 N

 
 

 

-y

N
»

In summary, it appears that no previously un-
known shorteperiod activities which might be of
importance in the shielding of circulating-fuel
reactors may be induced in K39 or Rb by reactor

PERIOD ENDING DECEMBER 10, 1954

neutron bombardment. The newly discovered 2.3-
Mev isomeric transition excited by inelastic
scattering in Zr’® may be of importance in design
considerations for a circulating-fuel reactor.

 

"

 

2
4
7

D

o
TR

i
"

157

7

60
{ f
b

 

 
  

 

 

 

~ ANP QUARTERLY PROGRESS REPORT

15. TOWER SHIELDING FACILITY
C. E. Clifford

T. V. Blosser
- L. B. Holland

J. L. Hull
F. N. Watson

Physics Division

D. L. Gilliland, General Electric Company
M. F. Valerino, NACA, Cleveland

J. Van Hoomissen, Boeing Airplane Company

© Tests on the GE-ANP R-1 divided shield mockup
began with measurements of the radiation around
“* " the reactor shield in the Tower Shielding Facility
" (TSF) reactor handling pool. The experimentation
~thus far has also included thermal-neutron flux and

 

some gamma-ray dose rate measurements in the

. TSF detector tank located a horizontal distance of
2264 ft from the reactor shield, This work was in-

errupted for a period of two and one-half weeks so

S ,,,that_ the TSF could participate in an Air Force
. project in which a group of monkeys were exposed
“to massive neutron radiation doses. The work on

the G-E experiment has now been resumed.
Analyses of some aspects of the TSF data have
been completed and are presented in Sec. 12,

“Shielding Analysis.”’

TSF EXPERIMENT WITH THE MOCKUP OF
THE GE-ANP R-1 SHIELD DESIGN

T. V. Blosser J. YVan Hoomissen

,D' L. Gilliland F. N. Watson

The mockup of the GE-ANP R-1 reactor shield
design (Fig. 15.1) is being used at the TSF for a
series of measurements, Fast-neutron and gamma-
ray dose rates were measured around the reactor
shield section while it was submerged in water in
the reactor handling pool. In addition, thermal-
neutron flux and gamma-ray dose rate measurements
have been made in the detector tank located a
fixed horizontal distance from the reactor shield.

The reactor loading for this experiment (Fig.
15.2) was a 5 x 7 fuel element array which gave a
symmetrical power distribution throughout the re-

< actor. In ofder to avoid excess reactivity, a fuel
- element was removed from the center of each of
" the two seven-element faces, The measurements

mthe ‘pool "at the points indicated in Fig. 15.1

 

i H. E. fil"\‘l‘Ufl’gerf.orc;{, Bulk Shielding F'acz'lfty Tests on
“the GE-ANP R-1 Divided Shield Mockup, ORNL CF-54-8-94

(to be issued); see also ANP Quar. Prog. Rep. Mar. 10,

1954, ORNL-1692, p 124.

 

were compared with similar measureme'nfs‘nfl'n'ade
earlier at the BSF.! L

In general, the BSF and TSF data were in agree-
ment, as is indicated in Figs. 15.3 and 15.4. How-
ever, they were not in complete agreement, largely
because of differences in the experimental setups
at the two facilities. For example, while the TSF
had a symmetrical power density distribution, the
fuel element loading at the BSF gave an asym-
metrical distribution which had a maximum 8.1 ecm
from the center line of the reactor grid. Also,
only compartments A and D contained borated water
at the BSF, whereas at the TSF all compartments
contained borated water for the first in-pool meas-
urements, Later, plain water was used in Com-
partment D. The BSF boration was 1.1 wt %, while
the TSF boration was 0.85 wt %, For the measure-
ments off the rear section at the BSF, no lead or
iron side shielding was present, while at the TSF
the side shielding shown in Fig. 15.1 was in
position, For the measurements off the front and
side sections, the side shield at the BSF had «
total of 1in. less steel than the side shield at the
TSF.

Measurements of the neutrons scattered into the
side of the detector tank (located a horizontal
distance of 70.8 ft from the G-E reactor shield)
were made as a function of altitude.. The flux
(Fig. 15.5) is a slowly varying function of the
altitude, and it exhibits a peak at about 35 ft.
A decrease of about 6% in the readings between
150 and 200 ft indicates that the ground-scattered
neufrons are still observable at these altitudes for
this particular reactor-shield combination. The
shape of the curve is quite different from that ob-
tained in the differential shielding experiments

with the water tank,? because the G-E shield

 

2C. E. Clifford et al., Preliminary Study of Fast Neutron
Ground and Air Scattering at the Tower Shielding Facility,
ORNL CF-54-8-95 (Aug. 23, 1954),

 

 

 
 

 

651

  
    
   
 
  
 
  
   
 
 

 

—SEcRE
2-04-056-3-T6A

'35.2510n. 38.00 in,

.50-in. GASKET

38in. STEEL

     
   
 

0.50in. STEEL

   

0.62in. STEEL
.381in. STEEL

012in. STEEL

  

40 cm

. I 20 ¢cm l-‘
T
TSF ) REAR r—

REACTOR COMPARTMENT F

0.75in. STEEL

  
     
  

FRONT
COMPARTMENT A

 
  

   
      
  

      
 
     
 
    

   

2.05in. LEAD
(3 PIECES)

    

381in. STEEL!

 

     
   
   
  

1.5in. STEEL -1.5in. LEAD

3in. STEEL

COMPARTMENT D

0.62in.STEEL
0.121in. ALUMINUM

in. STEEL

  

 

  
 

3in X 3in.
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VOID

 
 

COMPARTMENT C

   
    

COMPARTMENT B

   

    

0.62in.STEEL COMPARTMENT E

§

22.7
_L em ! 10.cm
S . ¥
[
20¢
; m X ~ POSITIONS OF GEOMETRICAL
)'(4i_ CENTERS OF COUNTERS

Fig. 15.1. Mockup of the GE-ANP R-1 Reactor Shield.

pS61°0L 439W3ID3Q ONIANI QOIY3d

 

 

TR T T R
 

 

 

 

ANP QUARTERLY PROGRESS REPORT

UNCLASSIFIED
2-01-056 -0-Ti4R{

 

- STANDARD FUEL
ELEMENT
RALF FUEL
L ] ELEMENT
|
.
AFETY RODS
fffff J

W

%3 i “ - OPEN LATTICE
= POSITION
R E@\ REGULATING ROD
s
= NOTE

S AND R ARE HALF
ELEMENTS

 

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|

 

 

 

 

 

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T T

Fig. 15.2, Tower Shielding Reactor Loading
No. 20

“SEeRET.
2—O01-056-3-T4A

9

1072

104

3
ol

1S

REAR AND SIDE FAST NEUTRON DOSE RATE (mrep/hr/wm‘t)' '
9
B

1078

FRONT FAST NEUTRON DOSE RATE (mrep/hr/watt)

 

0 10 20 30 40 50 &0 70
DISTANCE FROM SHIELD SURFACE (cm)

| Flg. 15.3. .qus't»Neuh"on‘ DoseA Rate Measure-
‘ments in Water Near GE-ANP R-1 Mockup.

160

4 2-01-056-3-T3A

O BSF

O TSF

2 & TSF (NO BORATED WATER IN
COMPARTMENT D)

GAMMA DOSE RATE (r/hr/watt)

 

1678
0 10 20 30 40 50 60 70

DISTANCE FROM SHIELD SURFACE (cm)

Fig. 15.4. Gamma-Ray Dose Rate Measurements
in Water Near GE-ANP R-1 Mockup.

mockup emits neutrons in a direction more favorable
for ground scattering. The intensity from the G-E
shield peaks at about 55 and 120 deg from the
reactor-detector center line and is symmetrical
about the reactor-detector axis,

Measurements were also made of the thermal-
neutron flux and the gomma-ray dose rates within
the detector tank at an altitude of 195 ft. The
G-E reactor shield was a horizontal distance of
64 ft from the detector tank, and, for these meas-
urements, five l-in.-thick lead slabs were installed
1 ft from the rear face (reactor side) of the tank.
The lead slabs simulated the shielding in the crew
compartment,

A knowledge of the thermal-neutron flux distri-
bution in the detector tank is useful in extending
the fast-neutron dose rate measurements into
regions where the intensity is too low to be meas-
ured with the dosimeter. This distribution is also
necessary for predicting the capture g'amma-rgjy

infensity to be expected in the crew shield. The

4=

.

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.‘} .

‘,\
 

o

4

U

@)
RS

a

€«

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‘»flgr5

- RELATIVE THERMAL NEUTRON FLUX (ARBITRARY SCALE)

flux (Fig. 15.6) along the y axis (coincident with
the reactor-detector axis) exhibits an average re-

" laxation length of 5.0 e¢m in the water between the

rear face and the lead. This indicates that these
neutrons are largely air-scattered neutrons, as
would be expected, since the thick neutron shadow
shield eliminates the direct beam. The apparent
transparency of the lead is suspect, since the
possibility exists that neutrons were streaming in
through the lead from the side of the detector tank.

An x traverse (Fig. 15.7) was taken in order to
indicate the attenuation to be expected from the

_.
<

e L,

S F;g.'|5.5 Thermal-Neutron Flux at Side of Detector Tank as a Function of Reactor-
Reactor in GE-ANP R-1 Mockup.

 

 

PERIOD ENDING DECEMBER 10, 1954

crew compartment side shielding, The apparent
relaxation length, which is not constant, is very
short, 4.2 cm, in the first few centimeters of pene-
tration and increases to 6.0 cm between 25 and
40 cm of penetration. This short relaxation length
is probably indicative of the large number of low-
energy neutrons escaping through the air ducts in
the reactor shield.

In a y traverse (Fig. 15.8) the gamma relaxation

length is approximately 13 cm at the front of the

detector tank and approximately 17 ¢m at the rear
of the tank.

SECRTT
1-01-056-3-3CD

CENTER OF TRIPLET BFy CHAMBER AT:
X —67.5cm
¥y 62.0c¢m
z 0.0c¢m

Dirrw 161

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~ ANP QUARTERLY PROGRESS REPORT

O

- o -
B

1-01-056-3-4(-)2 "R - '
1-01-056-3—142 (-1

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h 195 ft —— CENTER OF TRIPLET BF; CHAMBER AT:
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| 0 25 50 75 100 125 150
¥, DISTANCE FROM REAR FACE OF DETECTOR TANK TO DETECTOR CENTER (cm) -

F:g.'|5.6. f.lgl.termal-Ne.ufron Flux Along y Axis of Detector Tank; Reactor in GE-ANP R-'lMockup. o b _ o

162

i

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RELATIVE THERMAL NEUTRON FLUX (ARBITRARY SCALE)

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PERIOD ENDING DECEMBER 10, 1954

h 195 fi
d 64 ft
p —90 deg

x VARIABLE
y 90.0cm

HORIZONTAL DISTANGE FROM & AXIS TO DETECTOR CENTER (cm B

 

 

SEGRET

{-01-056-3-7(-)1R{

z 0.0 ¢m

1 Moc fl(u.]’.'n-.

CENTER OF TRIPLET BFy CHAMBER AT:

  

163

 
o B i ok e A B B A i B

 

 

 

Bk b ale Rl nmfl’a; b

 

 

 

SECRET
4 2-04—056-3—TI12R{A
10

CENTER OF 900-cc ION CHAMBER AT!

x 0.0cm
¥ VARIABLE

Z 00cm

h 195 ft
d 65ft

¢ 0O deg
1073

FIVE f=in. LEAD SLABS WITH
Y4—in. TRAPPED WATER

10

GAMMA DOSE RATE (mr hr'/wcn”r' front of tank)

 

5 5

2 2
40"7 40“5

148.8 428.8 108.8 88.8 ©8.8 48.8 28.8 8.8

y, DISTANCE FROM REAR FACE OF DETECTOR TANK TO DETECTOR CENTER (cm)

Fig. 15.8. Gamma Dose Rate Along y Axis of Detector Tank; Reactor in GE-ANP R-1 Mockup.

164
£ T ey Y %
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Part IV

APPENDIX

 

 
  

 

 

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ki el e et i e i i
1
¢
'HE AIRCRAFT NUCLEAR PROPULSION PROJECT |
AT t
|
THE OAK RIDGE NATIONAL LABORATORY ‘
DECEMBER 1, 1954 !
I
f
ANP PROJECT DIRECTOR W. H. JORDAN RO i
5. J. CROMER RD
ASSOCIATE DIRECTOR R. |. 5TROUGH ¥
ASSISTANT DIRECTOR A, J. MILLER RD |
D. HILYER, SEC. RD i
ASSISTANTS TO DIRECTORS REPORTS i
A, W. SAVOLAINEN ARE '
L. M, COOK ARE i
H. MCFATRIDGE, SEC. ARE P. HARMAN, SEC. ARE !
A. GIANGREGORIO PW LITERATURE SEARCHES
SEC. ARE
P. HARKAN, A. L. DAVIS ARE
[ |
ARCRAFT REACTOR ENGINEERING DIVISION L
S. J. CROMER, DIRECTOR RD
H. MCFATRIDGE, SEC. ARE SUPPORTING RESEARCH
W. H. JORDAN
A. J. MILLER
ASSISTANT TO DIRECTOR
L. M. COOK ARE
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