ORNL-TM-3595.txt

Contract No. W—74054eng-26

Reactor Division

INDEXED ABSTRACTSIOF SELECTED REFERENCES
ON MOLTEN-SALT REACTOR TECHNOLOGY

D. W. Cardwell and P. N. Haubenreich

 

NOTICE

] This report “was prepared as an account of work

sponsored by the United. States Government, Neither

; the United States nor the United States Atomic 'Energy

| ' Commission, nor any of their employees, nor any of

.| their contractors, subcontractors, or their employees,

- makes any warranty, express or implied, or assumes any

] legal liabitity or responsibility for the accuracy, com-

‘| pleteness or usefulness of any information, spparatus,

_product or process disclosed, or represents that its use
would not infnnge privately owned rights, -

 

 

 

- DECEMBER 1971

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
operated by
UNION CARBIDE CORPORAIION
FOR THE
U. S. ATOMIC ENERGY COMMISSION

1

 

ORNLfTMF3595

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- DISTRIBUTION OF THIS DOCUMENT IS ENUMITED

 
 

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MSTRACT . . - » .
INTRODUCTION . .
LIST OF ABSTRACTS.

KEYWORD INDEX. . .

CATEGORY INDEX . .
INDEX OF AUTHORS .

TABLE

iii

OF CONTENTS

w w =

169
197
205

 
 

 

iv

ACKNOWLEDGMENTS

The authors express their appreciation to Ruth Hofstra, John A.
Carpenter, and A. F. Joseph of the ORNL Mathematicé_Division fof_providing
technical guidance and programming required for file formatting,‘cbmputer
entry and automated printout of the MSR publication abstracts contained
in this report. Appreciation is also expressed to Annabel Legg who pre-
pared all text material for computer entry on our magnetic tape type-
writer/converter and to various staff members of the Molten Salt Reactor

Program who developed the abstracts and assigned keywords.

£

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INDEXED ABSTRACTS OF SELECTED REFERENCES
- ON MOLTEN-SALT REACTOR TECHNOLOGY

D. W, Cardwell and P. N. Haubenreich

ABSTRACT

~ Abstracts are given for 321 reports and articles which
provide an introduction to MSR technology and describe major
developments since 1960, Three indexes are provided: by key-

- word, by author, and by subject category.

 
 

 

 

 

 

 

 

 

 

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INTRODUCTION

This document contains abstracts of 321 selected reports and papers
which collectively provide a good, basic introduction to molten-salt re-
actor technology and describe major dévelopments in the field since the
initiation of the MSRE in 1960. As an aid in locating specific information,
three indexes are provided: by keyword, by author, and by subject category.

The abstracts and indexes, prepared and printed by a computer, were
taken from the file of the Molten-Salt Reactor Information System (MSRIS).
This is a growing file in the IBM-360 computer at ORNL, which can be
searched in various ways from femote terminals. A report is now being pre-

pared to describe MSRIS and how to use it.

LIST OF ABSTRACTS

In the pages which follow, abstracts are listed in the alphabetical
order of their primary subjeCt categories. Each’appears only once, even
though its subject may extend into éeveral other categories. Therefore to
find all abstracts having information on a particular subject, it is neces-
sary to use the category index. |

Each entry in this list consists of the abstract itself plus certain
other information about the reference. The first line is an identification
number, assighed when the_reference was added to the MSRIS file, The three

letters in this number'identify.thé-gtimarg subject category. If the ma-

 terial in the document'extendé'éignificantly into another category, this is

shown in the last line of'thg entry. Authors, title, and originating or-

'ganigation are listed on séparate'lines, then the document identification,

date of publication and'numbers"of_pages, figures, and references are given.
Following the abstract is a list ofrkeywqrds, with the most significant

marked by asterisks.

 
 

 

 

 

 

 

 

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Cateqgory A
Molten-Salt Reactcr Prograss

AAX67C003
Briggs EB
SUMMARY OF THE OBJECTIVES, THE DESIGN, AND A ERCGHEAY GCF DEV ELOPMENT
OF FOLTEN-SALT EREELER RFACIORS ) .
Oak Ridge Naticnal Laboratory, Tenn.
ORNL~TM-1851 (June 1967), 84 p, 20 fiqg, 13 ref.
Molten-salt thermal bkreeder reactors are characterized by
low specific inventcry, moderate breeding gain with low fuel
cycle cost, and high efficiency fcr ccnverting heat irsto
electricity. Studies indicate they should be akle to
produce electricity in 1000-Mw {e) staticns at a ccst that is
as lcw or lower than prcjected for advanced converter
reactors or fast breeder power stations. The fuel
utilization characteristics compare tavorably with thcse of
fast breeders. The present status of thke breeder technclcqy
is being dercnstrated in successtul operation of the MSRE.
A two-region Molten-Salt Breeder Experiunent tc demonstrate
all the basic technclogy for full-scale breeders is proposed
as the next step in the development. Design and
censtruction of the MSBE would be accompanied by a program
of fuels, materials, fuel reprocessing, and ergineerirg
development. Develcgment, construction, and startup of the
breeder reactor is estimated to take abcut eight years and
to ccst about $125 million.
*development + *MSRP ¢ *plans + *reviews + fuel cycle ccsts + MSBE +
MSBR + natural resources + performance + power costs + technology
1
AAX6TO004
Carter WL + Whatley HE
FUEL AND ELANKET EROCESSING CEVELOPMENT FOR MOLTEN SAL1
BREEDER REACTORS
Cak Ridge National Laboratory, Tenn,
ORNL-TH~- 1652 (June 1967) 52 p, 10 fig, 13 ref.
This document descrikes the fuel and klanket frccesses for
the MSBR, giving the current status of the technology and
outlining the needed development. It is ccncluded that the
principal needs are to develop the vacuum distillation and
protactinium removal operations, which tave beer
demcnstrated in the labcratory but not on an engineering
scale, A program to develop continucus fluoride volatility,
liquid-rhase reduction-reconstitution, improved xeaon
control, and special instrumentation shculd alsc be a major
~developmental effort. ' An estimate of manpower and cost for
developing MSBE fuel and fertile frocesses indicates that
it will require 288 manyears of effort over a 6-year period
at a total cost of about $18,000,C00.
*development + *¥SBR + *fprocessing + blanket + costs +
distillation + fuels + protactinium

AAX6T7C00S
Grimes WK .
CHEMICAL RESEARCH AND DEVELOPPFENT FCR FCLTEN-SALT EFEELER

Accession Numker AAX670003 to AAX670C(CS

 
 

Page 6

Category A
Molten-5alt Reactor Progrags | i —

AAX67C005 *Ccntinued* _
REACT CES | | 2
Oak Ridge National Laboratory, Tenn. :
CRNL-TM-1853 (June 1967), 140 p, 26 fig, 69 ret.
Results cf chemical research and development for molten
salt reactors are summarized. These results indicate that
LiF-BeF2-UF4 mixtures are feasible fuels for thermal
breeder reactors, Such mixtures show satisfactcry rphase
behavior, they are compatible with Hastelloy N and
moderator graphite, and they appear tc resist radiaticn and
" tolerate fissicn prcduct accumulation. Mixtures of
LiF-PeF2-ThP4 similarly appear suitakle as blankets fcr
such machines., Several possible secondary coolant mixtures
are available; NaF-NaPBF4 systems seem, at fpresent, to be
the most likely poss=sibility. Gaps in the technology are
presented along with the accomplishments, and ar atteupt
is pade to define the information {and the research and
development program) needed tefore an MSBR cart be ope1ated
witk confidence.
*chemistry + *development + *HSRP + *research + *rev1ewe +
compatibility + fissicn jroducts + fluorides + :
fluoroborates + molten salts + pldns + two-fluid reactcr
OTHER CATEGORIES: <CIXX

€

AAX670006
pcCoy HE + HWeir JR
MATERIALS DEVELOPMENT FOR MOLTEN-SALT BREEDER REACICRE
Cak Ridge Naticnal Labcratory, Tenn.
ORNL-TM-1854 (June 1967), 88 p, 28 fiy, 623 ref.
The materials develcpment program is described for a
two-region MSER with a uranium-ktearing flucride fuel
salt, a thorius-bearing fluoride blanket salt, and a
lower melting fluoride coolant salt, Tke primary struvctural
materials are grarhite and modified Hastelloy N. 1Individual
fuel cells will be graphite tutes, whict must withstarnd
10(23rd) neutrcns/sg-ct and have very low permeatility
to gases and mclten salts. Available graphites and
their properties are descrited in detail., A jfrcgranm
for cbtaining and evaluating improved graphites is
proposed. A progqgram is descrited in detail fcr develcping
modified Hastelloy N, which will be used in all rarts
of the system except the core. Brazing allecys and
a reasonable jcint design have been developed for a
joint between the graphite tukes and the mcdified
Hastelloy N. Needed inspection techniques are ccnsidered.
{This report is cone of a set of 9 on development prograas 3
regquired for an MSER.)
modified Hastelloy N + graphite + *development + *materials +
inspection + *MSPER + Lkrazing + compatiltility +
rechanical prorerties + ccsts + reviews + *KSEP + #fplans : C
THER CATEGCRIES: ELCX + FCX o \sfi

*}

)

AAX67C007

Accession Numtkter AAXfi?OOOS to AAX670007
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Category A
Molten-Salt Reactcr Prograrms

AAX67C007 ¥*Ccntinued*

Scott D + Grindell AG

COMPONENTS AND SYSTEMS DEVELGEMENT FCR !OLTEN SALT EREFELER
REACTCRS

Oak Ridge Naticnal Laboratory, Tenn. :

CRNL-TM-1855 (June 1967), S6.-p, 5 fiqg, £ taL, 19 ref.

- Studied thermal Molten-Salt freeder Fkeactors to 1dent1fy
important design and development proklers. The purpose was
to crganize these rproblems into a progras which would
produce components for use in, a2 Molten-cSalt Breeder
Experiment, The reference-design concept is a two-region
two~-fluid system with the fuel salt separated frcm the
blanket salt by grarhite tubes. The energy produced in the
reactor fluid is transferred to a seccndary ccclant-salt
circuit, which couples the reactor to a supercritical
steam cycle. The specific development froblems tc be
studied include the reactcr core and heat exchanger
hydraulics, pumps for the three salt systems, heat transfer
in the heat exchangers and boiler-superheater, mechanical
valves for salt-flow control, ccntrecl rc¢d and drive,
pressure relief in cooclant system, cell 1nsu1at10n an G
heaters, and the cover-gas.

*compcnents + *develcpment + *MSBR + *MSRFE + *plans +
*reviews + control-rod drives + control rcds + ccres +
heaters + heat exchangers + hydraulics +

pumps + steam systems + two-fluid reactor + valves +
thermal insulaticn

CTHER CATEGCEIES: HXX

AAX67C008B

Tallackson JR + Moore RL + Litto SJ

INSTRUMENTAIION AND CONTRCLS DEVELCEMENT FCER PCLTEN-SPLT

BREFDER FEACTCES

Oak Ridge Naticpal Laboratory, Tenn.

CRNL-THM-1856 (May 1967), 36 p, Z ref. : '
Instrumentation used-in the MSRE is a good basis for
development of the instrumentation for large sclten-salt
breeder reactors. The development would ipvclve rrimarily
the testing and improvement of existing instrument -
components and systems. New cr much improved devices are
required fcr measuring flow= and pressures of molten salts
in the fuel and blanket circulating systems., Nc frcblems
are foreseen that should delay the design or construction
of a breeder reactor experllent. An estimate cf costc of
develoring HSR instruments is given, :

*development + *instrumentation # *MSBHR + *systess ¢+

- components + ccntrol + flcw measurement ¢ FESRE ¢ MSEP +

plans + measurement + radiation measurerent +
temperature measurement + weigh cell
OTHER CATEGCEFIES: JXX

AAX67C008

Accession Number AAX67C007 to AAX6T7CCCS

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Page . 8

Category A
Molten-Salt Reactcr Prograts

AAX67C009 *Ccntinued*

Ferry AP

PHYSICS PROGRAM FOR MOLTEN-SALT BREEDEPR REACTCRS

Cak Eidge Kational Laboratory, Tenn.

ORNL-TM-1€57 (June 1967) 40 p, 4 fig, 11 ref,
The sources of possille error in estimates ot breeding
performance of a Mclten-Salt Breeder Feactor are discussed.
Uncertainties in cross sections may ccntribute ar
uncertainty of about plus or minus 0.026 in breeding ratio.
Other sources of error may arise from assumpticrs regarding
“behavior of fission products, or trox inadequacies in
methods of computation., A reactor physics develcgment
program is cutlined which should provide a sound tasis for
design of a reactor experiment. The prcgrar includes
thecrétical investigation of system dynamic characteristics,
evaluation of alternate core designs, develogprent of
copputational methods, cross-section evaluation, development
of computer codes and experimental physics. Prcgraec
manpocwer requirements and costs are estimated. (This report
is one of a set of nine on development frograss reguired for
an MSEBR.)

MSBR + *breeding performance + *nuclear analysis +

kcross sections + conputer codes + rare earths +

fission products + *MSRP + dynamic characteristics +

neutron yield + costs + *fplans + stability + 3*design data +

calculations + methods :

OTHER CATEGORIES: BIXX

ABX670010

Kasten PR

SAFETY PRCGEAM FOF MOLTEN-SALT BREECER REACTORS

0ak Ridge Naticnal Laboratory, Tenn.

ORNL-TM-1858 (June 1967) 42 p, 6 fig, 3 ref,
Investigaticns required in determining the safety
characteristics of MSBR power fplants are outlined,
and the safety features of the major. plant systenms
are described, Reactivity additicns which need detailed
study include those asscciated with net fuel addition
to the core region, graphite lehavior, changes it ,
fluid flow conditicns, and control rod movement. Reactivity
coefficients which reguire evaluation include thcse
asscciated with temperature, voids, pressure, fuel
concentration, and graphite concentraticn., The integrity
of plant containpment under reactivity incident conditions
and, also under circumstances where reactivity itself
is nct invclved, needs to be evaluated. Stability
analysis of the reactor plant is required. Pkysical
behavior of materials and of equipment under NSEHh
conditions, as they relate to reactor safety, neced
tc ke deterrined exgerimentally. To delineate and
resolve the basic safety proklems associated with
MSBR systems, about $1.3 million is required over

Accession Numker AAX67C00€ to AAX670010

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Page 9

Category A
Molten-Salt Reactcer Programs. -

AAX67C010 *Ccntinuedx
- a period of about elqht years, with most of the effcrt

{$0.9 millicn) occurring during the first four years. .
(This report is one of a set of nine cn develcprent
programs required fcr an MSBR.)

*MSRP + *safety + *analysis + *plans *+ reactivity + MSEER +

accidents + ccsts + containment + statility +

dynamic characteristics + off-gas fiyctems + proce951ng

CTHER CATEGCFIES: EGX

AAX67C011

Blumberg &

MAINTENANCE DEVELOPMENT FCR MCLTEN SALT BBhEBEE BEACTCRS

Cak Ridge National Laboratory, Tenn.

ORNL-TM-16S9 (June 1967), 18 g, 1 fig, 6 ref. :
The maintenance system of the proposed solten-salt breeder
reactcors will be based upon the technology in use and
experience gained from the Molten-Salt Reactcr Experizent.

- The unit replacement scheme, long-handled tools, movatle
maintenance shields, and the means for landlirg ccntaninated:
egquiprent will be similar for many operations. The
techniques must be improved and extendec¢ and rew techrigues
must be developed fcr maintaining some of the larger, more
radioactive components of the ktreeder reactcrs., Renmote
velding is needed fcr major component replacement, Methods

- must be available for replacing the ccre ard fcr the rerpair
of heat exchanger. Finally, a general development angd
design surveillance progran uill’te'required. These
programs are described and their cost is estimated. {This
report is one of a set of 9 on develcyment prcqraus required
for an MSBR.)

*Raintenance + *MSBR ¢ *pldns + develogment + MSRE ¢+
remote welding S .
OTHER CATEGCBIES: KEE

ABXS80001

MacPherson HG

MOLTEN-SALT REACTORS - '

Cak Ridge National Laboratory,,Tenn.

Part II of Fluid-Fuel Reactorh, Addlson Wegley (1958),
PP 563-697.

" The early hlstcry and 1958 development status of molten--
salt reactors is covered in 7 chapters cf this bcck,
prepared for the second Geneva Conference. Chapter tOple
include chemistry, materials, nuclear aspects, keat-
transfer equipment, the Aircraft Feactor Experlment,'and
a conceptual design of a power reactor. The ccncept
presented has a core and blanket, Wlth no moderator other
than the LiF-BeF2 carrier salt,

¥*ARL ¢+ *develcrment + ¥MSRP + *reviews + *technoloqy +
chemistry + corrosion + Eastelloy N ¢ inconels +

Accessicn Number AAX670010 to AEX5E00C1

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Page 10

Category A
Mclten-Salt Reactor Prograss

ABX5800C1 *Ccntinuedx*
molten salts

ABX64C00U4
Briggs ®B
MOLTEN-SALT PO&ER REACTORS AND THE RCLE OF THE M3 EE 1IN THEIF
DEVELCEMENT (FART OF MSRP SEMIANN PROG REPT 7/31/6“)
Oak Ridge Naticmal Labcratory, Tenn.
ORNL-3708 (Nov. 1964), pp 3-21, 7 fiqg, 8 ref,
ORNL studies show the mclten-salt reactor to te the most

promising thermal-neutron thorium-0233 treeder ccncept, In

this paper, a compact 500-MWe two-fluid breeder with
graphite tubes separating fuel and fertile salts is
described and its processing and economics are discussed.
The MSRE was authorized in 1960 tc investigate chemistry,
materials, engineering and operation of the MSE concept.
Success with the MSRE should lead to ccrstructicrn of a
ccnverter reactcr that could -be modified to beccme a
breeder.

*MSRP + *two-fluid reactcer + breedlng performance +

design + development + economics + MSBR + MSRE + plancs +

reviews

ABX670049
MacPherson HG
MOLTEN-SAIT REACTCR SHOWS MOST PROMISE TO CON‘ERVE NUCIEAR
FUELS
Cak Ridge National Laboratory, Tenn.
Power Enqineering 71, 1 and 2 (Jan and Feb 1967), 7 p,
6 fig, 6 ref.
The MSBR promlees tc comblne simplified fuel recycle and
stable fuel in a high-perforrance thermal treeder having
lov rover costs. The present concept of an MSEF has tfuel
and fertile salts separdted Ly qraphite in a 14-f¢t
reactcr vessecl. MSRE experience has shown molten salt
reactors to be practical, A S50-MWe two-fluid breeder is
suggested as the next step.
*breeding performance + *economics 4+ *MSBR +
*natural resources + conceptual design + experieunce + MSRE +
plans + reviews

ABX68003%

MacPherson HG

MOLTEN-SALT REACTORS

Cak Ridge NKNational Laboratory, Tenn.

Proc. Intl. Conf. on Ccnstructive Uses of Atomic Energy,

Wwashington, RNov., 1968, pp. 111-121, 7 tig, L ref. -

Experiserts on feasibility cof molten salts as reactor fuels
started in 1947 in the aircraft reactor prcqrar. The
concept now features rolten fluoride salt containing UF4
and ThF4 circulated through a graphite ccre. Advantages

Accessicn Number ABXSEB0CC1 to ABX6E0C3S

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Page 11

| CatégOry.A
Molten-Salt Reactcor Prograss

ABX680035 *Centinued*
of low-pressure, hlgh-temperature, fluié¢ fuel frcmnote
safety and economy. Research and development have
concentrated on materials, compatibility, comfcrents and
the MSRE. Recent advances include improved materials and
simplified processing,  Conceptual desicn studies of
‘one-fluid molten-salt breeder reactors indicate good
breeding performance and low power costs

*MSRP + *reviews + breeding performance + costs +

development + MSBR + safety + technology

ABX69C007

Haubenreich EN + Eosenthal MW

MOLTEN-SALT REACTORS

cak Ridge National Laboratory, Tenn, -

Science Jcurnal 5 (6) (Jumne 1969), 6 p, S fig, 4 ref.
Breeder reactors are needed to keep power ccsts dcwn
as uranium prices rise. Development emphasis is on
fast breeders, which promise high gain. Therpal breeders
must have fast processing to remove protactinium and
poisons to achieve moderate gain, but fissile paterial
investments can be low. The fluid-fuel molten-salt
reactor with on-site processing promises lcw fuel cycle
cost and acceptable doubling times. MSF development
dates back to 1948 and includes successful cperation
of the MSRE at 650 deg C for over three years. The
molten-salt breeder concept is now a graphite ccte with
circunlating salt ccntaining both uranium and thorlum,
processed bty reductive extracticn . .intc tissutk.

*treeding perfcrmance + *economics + *electrical power +

#MSBR + *natural resources + *reviews + experience +

fuel cycle costs + MSRE + MSRE + processing

ABX690056
Rosenthal MW + Robertson RC + Bettlc ES
MCLTEN-SAIT BRFEELEHE REACTCRS -
Oak Ridge Naticpnal Laboratory, Tenn. '
Nucl. Engrg. Int. Vol. 14, No. 156§(%ay51969), FE- kZC-425
5 t1q.
This article explalnn hcw molten-salt reactcres cfter
lcw-ccst power now and in the future because ¢f good
breeding performance and inherent advantages ¢f gclter-salt
fuel. Brief descrifptions are given or PFSBF materials,
core design, components, and processing scheze. After
discussing MSR maintenance, safety, and costs, the
- authors conclude with an outline cf wc:k requlred tc
develcp a large commercial MSEE.
*MsRP + *reviews + breeding periormance + ccm;onentc +
costs + development + HSBR + safety + technoloqy

ABX700054

Accessicn Number ABX680035 to AEXT7000%4

 
 

Category 2
Mclten-Salt Reactor Erograms

ABX700054 *Continuedx |

Rosenthal MW + Kasten PR + Briggs BB

MOLTEK-SALT RERCTORS =-—- HISTORY, STATUS, AND PFOTENTIAL

Oak Ridge Naticnal Labcratory, Tenn. _ |

Nucl. 2ppl. Tech. 8, 107 (Fek. 1970), 11 p, 3 fig, 1€ ref.
Molten-salt breeder reactors being developed at CENL
promise safe, low—cost power while extending rescurces of
fissicnable material., MSR technology, develciirng since
1947, was adequate for successful construction and
operation of the MSRE which showed thkat circulating
molten fuel is practical, that fluoride salts are statle
under reactor conditions, and that ccrrcsicn is very lovw.
The sivple fuel prccessing necessary for a converter was
demonstrated in the MSRE. Ptocessing methcds being
develcged should permit MSR*s in which UF4 and ThEF4 are
combined in a single salt flowing throuch a graghite
nmoderator to ofperate as economical breeders. Initial
startup can be with U-235, U-233, or Pu-23S. Ccrstruction
costs shculd be about the same as light-water reactors and
fuel costs should ke much lower. Achievement of eccacnmic
MSBR's requires develorment and construction of severaezl -
MSE plants of increasing size.

*MSRP ¢+ *reviews + ARE + breeding performance +

capital costs + design + development @ fuel cycle ccsts +

MSBR + materials + processing + safety + technology

ABX700055

Shaw M + Landis JW ¢+ Laney RV + Rosenthal KW + Layman WH

Us S. SUERVEY: REACTOR LEVELOPMENT PROGRAM

United States Atcmic Epergy Ccmmission

Nucl. Eng. Int. Vol. 15, No. 173 (Nov. 197C), pp. €S9-GC4,

4 fiq.

In the U.S.A. there vwas proliferation of reactcr concepts
in the 1650's eliminaticns in the 1960 's; development
efforts are now concentrated on € concegts: Light Wa ter,
Liquid Metal-cocled Past Breeder, lLight Water Ereeder,
Molten-Salt Freeder, High-Terperature Gas, ard Gas-Cocled
Fast Breeder. This article covers the development status
of each. The molten-salt reactor program, sictce the
conclusicn of the MSRE, includes: design studies, reactor
systems and equipment development, chemical frccessing,
materials, and chemistry. :

*AEC + *development + *electrical pcwver + *reactcrs +

*reviews + foreign _

ABX710020

Grenon M ¢ Geist JJ

IES REACTEUES A SELS FONLUS

Euratcm |

Energie Nucleaire, Vol. 13, No. 2 {Mar.-Apr. 1971)
pp. 86-%3, 1C figq.

Accession Number ABX700054 to AEX710020.

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Category A
Molten-Salt Reactor Erograms

ABX710020 *Continued¥ |
This article (in French) appears in a series on chemical
sciences, The authors, formerly involved in the
Euratcm~USAEC exchange c¢n molten-salt reactors, introduce
the MSR as a potential treeder worthy cf multinational
consideration. They describe the concept, early
developrent, recent progress, proctlems, advantages angd
possible future development. (An English translation,
OFNI-tr-2508, is availaktle from ORNL.)

*development + *econcmics + ¥PFSBR + ¥MSEP +

breeding performance + foreign + reviews

ACAG65C00Y4

Haubenreich PN

HOLTEN-SALT REACTOR EXPERIHFNi (PABT 1 FSRE EEUGI. FFET.
2/28/65)

Oak Ridge Natiocnal Laboratory,,Tenn.

ORNL-3812 (June 1965), pp. 5-603, 17 fig, 29 ref.

- Construction of the salt systems and clcsely asscciated
ancillary systems was ccrmpleted and full-time prenuclear
testing began in Septemter. After leak—-testirg, fFurging
and heating cf the-salt'systems,-flush salt and coolant
salt were loaded, Transfers and circulaticn fcllcwed.
Testing showed the need for modification of radiator doors,
freeze-valve air supplies and controls therral &hield
water piping and scre ccollnq ‘air control valves.
Krypton-85 was lnjected Lnto the fuel system tc test
‘remcval mechanisrms. _

*construction + *éxperience + *MSRE # #startufp + *testing +
drying + freceze valves + krypten + loading + wmolten salts +
thermal insulation '
OTHER CATEGORIES: MXX + KAB .

ACA650010

Hautenreich PN | | | - |

MOLTEN-SALT REACTOR EXPERIflENT {PART 1 MSRP PROGR. REP1.
8/31/6€5)

Cak Ridge National Ldboxatory, Tenn.

CRNL-3872 (Dec. 1965), pg. 7-78, 34 fig, 4C ref.

Prenuclear testing thh flush falt vas cowpleted in
- March after 1000 hours of salt circulaticn., In preparation
fer lew-fpower -nuclear operatxon, nuclear instruments, the
fuel sampler-enricher and one layer c¢f the reactcr cell
roof blocks were 1n=ta11ed ‘and reactor operatcrs received
~additional training., Fuel carrier salt. ccrtainan de fle ted
uranium was loaded and circulated for 10 days in May lefore
additions of entiched U~235 kegan, first intc the draian
“tanks, then through the -pump bowl. Criticality was reached
~on June 1 at very near the predicted loading. - Subseqtent
small additions of 0-235 permitted calikration of the
control rods and measurement of reactivity ccefficients and

Accessicntflumber'an710020 to ACAGE001C

 
 

 

Page 14

Category A
Mclten~Salt Reactor Erograms

ACA650010 *Continuedx
provided enough reactivity to operate for several months at
power. Zero-pover measurements and dynamics tests wele
completed in July ard final preparations for high-power
operation were started,

*criticality + *experience + *MSRE + #*operation +

*startup + control rods + dynamics tests + lcading +

measutrement + gpolten salts + reactlvlty + operators +

training + testing -

OTHER CATEGORIES: MXX + KAB

ACA66 0008

Hautenreich PN

MOLTEN-SAIT REACTOR EXPERIMENT (PART 1 MSKP PROGR. REPT,

- 2/2E8,66)

cak Ridge NXational Laboratory, Tenn.

CRNL-3936 (June 1966), pr. 7-92, 41 fig, 43 ref.
Preparations fcr high-pcver operation were completed.
These included modifying coolant line archcr sleeves,
replacing radiatcr doors, inspecting fuel pump internals,
measuring salt piping stresses, heat treating the reactocr
vessel, sealing and testing secondary containment,
installing new core specimens, improving insulaticn or the
radiator enclosure, and further training of operators.
Nuclear operation resumed in Lecerker ard tests at fposers
up tc 1 MW verified predicted dynamic behavior. The
power ascension was interrupted at 1 MW vhen valves and
filters in the fuel off-gqas system plugged. Investigetion
revealed radiation-polymerized decompcsiticn prcducts of
oil that had leaked intc the fuel pump Lbowl.

*experience + *%MSRE + soperation + analysis + containgernt +

contrcl rcds ¢+ dynamics tests + heat treatments +

off-gas systems + piping + pumps + remote mairtenance +

stability + startup + stress + testing

CTHER CATEGCEIES: MXX + KABR + KBaA

ACAGECC1Y

Haubenreich PN -

MOLTEN~SALT REACTOR EXPERTMENT (PAET 1 KESEE FFGGF. FEET.
8/31/66) |

0ak Ridge Naticnal Labcratory, Tenn.

- ORNL-4037 (Jan. 1967), pp. 1-94, 24 fig, 35 ref.

Power ascension was resumed in April atter a large,

efficient filter assembly was installed to protect the

fuel systea pressure control valve frcm cil deccrfposition

products. Full power of 7.5 ¥R (limited by heat removal

capability) was reached in May. 7Tests at eack stage

verified predictions except that xenon stripping was gore

effective and heat transfer frcm the radiatcr was lcwer

than expected. Restrictions at the fuel off-gas charcoal

bed inlets developed tut were cleared hy backklcwing.)

_Accessicn Number ACA650010 to ACA6E00 14
.

)

Page 15

Category A
Mclten-Salt Reactor Ekrograms

ACA660014 *Continued*
Operation was interrupted briefly by electrical failures in
a component cooling pump and the fuel sampler, a false
indication of containment cell leakage, and failure of a
drive coupling on a main tlouer. Hut ard blades of a main
blower shattered on July 17, forcing a shutdown. Flaws
vere found in the hubs of the other klower ard the spare
and procurement of nev blowets was started. The delay was
used to remove core specimens, alter the radiatcr dcor
seals, irnstall equlpment to handle radiolytic gas from the
thermal shield, repair leaky cell ccolers, ard renove the
off-gas partlcle trap for examxnatlon. Flush salt got into

- some gas lines vhen the fuel pumfp was accidentally

overfilled, and this was nmelted out by temporary heaters.

*experience + *MSRE + *maintenance + *cperaticn +

analysis e blowers + compcnents + cracks + failures +

filters + fission products + heat transfer +

off-gas systems + radlolyclc + remote malntenance +

samplers + startup

OTHER CATEGORIES: MXX + KAB + KBA

ACA670016
Haukenreich PN
MOLTEN-SAIT REACTOR EXPERIHENT (PART 1 MSRP PROGB. REPT.
2/28/€71)
Cak Ridge National Laboratory, Tenn,
ORNL-4119 (July 1967), pp. 1-94, 36 fig, 42 ref.
Replacement blcwers were received in Cctober and oper ation
vas resumed after a 12-week shutdcwn, A restriction which
-appeared in the off-gas line at the fuel pump Lbowl wvas
temporarily relieved Lty heating, tut bhad tc cleared
mechanically in Ncvember, After a successrul 30-day run
at full power, the reactor was shut dcwr in Janvary tc
1nspect the blcwers and to replace air line disconnects
'in the reactor cell whcse leakage had interfered with
measurement of containment cell leaxkace., At the sage tinme
the fuel off-gas filter assembly was replaced with two
parallel particle traps ot 1mproved design, (Ir the
first particle trags expansxon of some parts tended to
throttle the flow upon beating ty fresh ftissicn rroducts.)
A ccuprehensxve reactivity balance (including automatic
computatlon at frequent intervals by the on-lire comptter)
~becane operational and unexplained reactivity changes from
the beginning of operation vere shown tc be crly C.05%.
Full-fover operatlcn was resumed and contlnued through
-February.
*experlence + *HSRE + *operatlon + analyszb + blowers +
off—qas systems + redctivity + remote malntenance +
disconnects
CTHER CATEGCRIES: mxx + KAE + KBA

ACA67C023

Accession Numker ACAG6€C014 to ACA67CCZ3

 
 

Page 16

-Category A
Molten-Salt Reactcr Progracs

ACA67C023 *Ccntinueds

Haubenreich EN

MOLTEN-SALT REACTOR EXPERIMENT {PART 1 FMSRE SEMIANN EFCG

REPT 8/31/67) |

Oak Ridge National Labo:atory, Tenne.

CRNL-4191 (Dec. 1967), pp 13-62, 33 fig, 25 ref.
Run 11 continued for 10z days, over 90% at full fcver,
befcre a scheduled shutdown May 10. Fakeup U- 235, ‘ad éed
at power for the first time, mixed in 2 rinutes, The
new offgas particle trap worked sell, but the charcoal
bed inlets occasionally plugqged. A rairp blcwer bearirg
had tc be replaced during the run. LCuring the May-
June shutdown, core specimens were reglaced. 2 remcte
gamra spectrometer was tested and used to scan the
primary heat exchanger. Minor maintepance was also
done and annual tests uere completed., The next run
was 42 days at full power, with e€xphasis or beryllium
additions and fuvel sampling. Shutdown came after the
fuel sampler cable tangled and was severed. 7Tccls vere
develcped and the latch sas retrieved, but not the
capsule. Operations analysis included long-tern reactivity
effects, thermccougle drift, and salt heat transfer,
In preparation for replacement of the uraniunm ir the
fuel with 0-233, neutrcnic characteristics with this
fissile material were calculated,

kexperience + *pmaintepance + *MSRE + #*operation + dnalY81S +

bearings ¢+ components + gamma spectrchetry + keat transfer +

off-gas systems + perfcrmance + reactivity +

remote maintenance + reactivity + temperature preasurefernt +

uranium-233 + sarmplers

CTHER CATEGCETES: MXX + KAR + MLX + MEC

ACA68C012

Haubenreich EN

MOLTEN-SALT REACTOR EXPERIMENT (PART 1 ESRE SEMIANK EFCC

REFT 2/29/68)

Oak Ridge Naticnal Labora tory, Tenn.

ORNL-42¢84 (Aug. 1968), 49 p, 35 fiqg, 32 ref.
Early in the period the fuel sampler was reinstalled
and full-pcwer operation resumed. RAfter the startup
vas interrupted to repair a component ccoling fpurnyp,
there was 6 ponths withcut a fuel dArain. Fuel circulation
was stopped 2 days in Novemker fcr wcrk cn the sampletl
and during a xencn exreriment at low power a tearing
was replaced on a main tlower. Otherwise nc equigment .
protlem interfered with operation. Cperation at various
fuel levels, temperatures and pressures shcwed effects
on xenon stripping, neutrcn noise, and gas in the
access nozzle, Reactivity Lkalances showed slight
drift (0.1%) over the lcng run., An offgas sampler
was installed downstream of tke charccal beds. The

Accessicn Number ACA670023 to ACAG68CC12
Page 17

“Cateqgory 2
Molten-Salt Reactor Prograss

ACA68C012 *Ccntinued*
freeze flange thermal cycle test, stopped after 103 cycle
was resurged. Analycec cf system dynamics with the
proposed U-233 fuel predicted safe and stable operation.

kexperience + *maintenance + *MSRE ¢+ *ofperaticn + aralysis +

hear1ngs + comgpcnents + dynamic characteristics +

noise analysis + off-gas systems + reactivity +

aranium—233 + xenon + samflers + freeze flanges

CTHER CATEGCEIES: MNXX + KAE e MDX + MEC

ACA68C019

Haubenreich PN ) |

MOLTEN-SALT REACTOB EXPERIMENT (PART 1 ESBF SEMIANN EFCG HEET

8/31/68)

Oak Ridge Naticnal Laboxatory, Tenn.

ORNL-U4344 (Feb. 1969), pp. 1-52, 28 fig, 43 ref
A 6-month run, ending in March, ccncluded cperatlcn with
U-235 after 900n eguivalent full -power hours. After _

- shutdown, gamma-spectrometrlc measurenerts were tade cn the
fuel system, core srecimens were replaced, the fuel offgas
line was cleared and two heaters frcr tlke prisary heat
exchanger were repaired. All 15 rod-scram relays vere
replaced and 3 of the new relays failed. A capsuvle diopped
in the fuel sarpler could not be retrieved, tut did not
prevent fuel sampling. The on-site prccessing equipmert was
readied fcr remecval cf the uranium from the salt. After
testing, the sulfur dioxide reaction syster fcr disposal of
excess fluorine was abandoned in favor of reaction with a
caustic solution. In Augqust the flush salt ard the frel
salt were'tlnorlnated, efficiently recovering the U as the
hexafluoride. Corrosion products were ptrecipitated and
filtered in the final step before U-233 loading.
Theoretical analyses of U-233 operaticn, includirg refined
calculaticn of delayed neutron effects, showed that the
system wvould be quite stable, After 2¢t test thermal cycles
-~ of the prototype freeze flange, a,crack was found at the
- -alignment stub.

*egxperience + *flucrinaticn + *malntenancp + 4MSEE + *operatlon +

analysis + dynamic characteristics + freeze flange= + ' :

samplers + uraniur-233 + cff-gas syctems

CTHER CATEGQBIE%' MXX + KAE + MDX 4 MEC + LHX

ACAG9C021

Haubenreich EN ' ' ’ ’ - - o

MOLTEN-SALT REACTOR EXPEBIMLNT (raat 1 MSBE SEMTANK EECG REET

2/28/69) o :

Oak Ridge Naticnal Laho:a{ory, Tenn. |

ORNL-4396 (Aug. 1969) pp. 1-47, 32 fiq, 40 ref.
The MSRFE began nuclear operatlcn with 0—433 ir Cegtemter and
was tbtrought to full pcwer in January. Criticality was

~attained by adding 33 kg of U, as the UF4-LiF eutectic, to

the carrier salt frcm which the original U-23%f had been

Accession Numlker ACA680012 tc¢ ACA690021 o

 
 

Page 18

Category A
Molten-Salt Reactor Erogrars

ACR690021 *Continuedx
stripred. Startup tests included measurements of rod worth
and reactivity coefficients, dynamics tests, and noise
analysis. When beryllium metal was added to adjust the
reducing power of the salt, the entrained cover gas
increased from less thanm 0.1 vol % to 0.6 vol %, apparently
due to slight changes in the physical prcperties cf tbke
salt. During the pcwer ascension, small perturtations in
nuclear pover were seen. Analysis indicated they were due
tc gas in the core, and they did not occur when gas
entrainment was reduced ty siowing the fuel pumf. Befcre
the pcwer ascension, the fuel offyas line was cleared of a
restriction, Shortly afterward a loose cear ir the fuel
sampler forced a 3-week shutdoun, during which time a
control-rod drive was serviced., Therral cycle testing cf the
prototype freeze flange ccntinued and test-stand operation
of a fuel pump with a deeper towl {Mark-2) begar. |

*experience + *paintenance + *MSRE ¢ operation + analysis +

control rods + dynamics tests + irceze flangecs +

off-gas systems + reactivity ¢ uraniuvm-2z33 + ncise apalysis

OTHER CATEGORTIES: MXX 4+ KAB + MDX + MEC _

ACA690028

Haubenreich PN ' . k

MOLTEMN-SALIT EEACTOR EXPERIMENT (PART 1 MSEP SEMIANN PRCG REFORT

8/31/69) -

Cak Ridge National Laboratory, Tenn.

CRNL-4449 (Feb. 1970), pp. 1-38, 19 fiy, 37 ref. |
High-fowver operaticn with 0-233, which kegan in January,
continued until a scheduled shutdcen cn June 1, There were
few equirment problems cther than restrictions in the offgas
lines, and the reactor was critical 5% of the time f1icom
Januvary to June, Fuel samples vere taken periodically to
measure U-233 capture-to-fission ratic and tc study fcel
chemistry. Tests ccocntinved on the behavior of cover gas and
xenon in the fuel systemr at variocus circulaticn rates.
Ccntinuous indicatcrs of reactor pressure and neutron n01sc
levels were installed and used. During the shutdcwn, a new
Ccre srecimen array was installed, a stiff ccntrol rod was
replaced, rod drives were repaired, and the cifgas lires
vere cleared, A rerote gamma-ray spectrometer was used to
measure fission-product distriltutions with the salt drained
and during the startup. Annual containgent tests concluded
the 10-wveek shutdown. Operation resused with exferiments

comparing argon and helium as cover gases. Component
development work included extensicn ci the prctctype freeze
flange thermal cycle test through 400 cycles, and operation
of the Mark-2 fuel pump with a2 high salt level tc redcce
entrainzent.

*experience + *maintenance + ¥*NSRE + *0peraticn + aralysis +

contrcl rcds + ccver gas + freeze flanges + gamma spectrometry +

noise analysis + off-gas systems ¢ uranium-:33 + xercr .

OTHER CATEGORIES: MXX + KAB + MDX + MEC

ACAT700021

Accessicn Number ACA690021 to ACA700021
Page 19

. Category A
Mclten-Salt Reactor Erograts

ACAT700021 *Continued¥*

Haukenreich PN

MOLTEK-SALT RERCTOR FXPERIHENT {(PART 1, MSRP cEMIANN PROG REPI

2/2€,70) '

cak Ridge National Laboratory, Tenne.

ORNL-4548 (Aug. 1970), pp. 1-40, 14 fig, 49 ref.
Nuclear cperation of the ¥SRE was concluded on Lec. 152,
1969, FPFrincipal actlvltles during the final rurs were
studies cf xenon stripring and tritiom distritution, and
sampllng to determine fission product tehavicr. A resote
gamgpa-ray spectrometer was also used during both operation
and shutdown to observe fissicn product distributions
After the final shutdcwn the reactor was placed in standby,
awaiting later examination. A4 swmall leak neatr a freeze
valve during the shutdcwn released some fissicn products
into the containment cell. Refined analyses ctf reactivity
calculations and long-term behavior confirmed the good
ajreement. The prototype freeze flange undergcing thermal
cycle testing was irpspected after 470 cycles, then was run
on to 540 cycles before the test was discontirued. |

*experience + *¥SRE + *operation + analysis + treeze flanges +

gamma spectrometry + leaks + noxse analy51s + reactivity + tritium +

xenon ; ‘

CTHER CATEGCERIES: axx + KAB + MDX

ACAT70C035

Haubenreich EN

MOLTEN-SALT REACTOR EXPBRIMENT (PABT 1 FSRE SEEIAVL EERCC

REET 8/31/70)

Oak Ridge Naticnal Labcratory, Tenn, -

CRNL-4622 (Jan. 1S871), pp. 1-6, 3 fiqg, 12 ref.
The KSEE remained shut down, awaiting pc¢stcreraticn
exaxriration., Procedures and tools were prepared.
Specimens vere cut from the coclant systen Fiting.
Analysis of data taken with the remote gamma spectrometer
‘during the final runs suggested that 'ncble-metal' fission
products quickly migrate, as extremely small fparticles,
‘to metal surfaces or salt-gas interfaces Existing data
on radioclytic flucrlne evcolution from frozen salt indicate
that evolution from the M3RE fuel in stcrage is easily
prevented. MSRE ccrponent development ended with
termination of" pump endurance tests,

MSRE + ‘analysis + experlence + radlolYSLb + fluorlne +

examinations -

OTHER CRTEPORIES: HXX-+_CF

ACA710028

Haukenreich PN ' o - - ,

MOLTEK-SAIT REACTOR EXPEBIMENT tPART[I; MSRP SEMIANN PROG
REPT 272€,71) e -

CaXx Ridge Fational Laboratory, Tenn. o

ORNL-4676 (Aug. 1971), rg. 120, 16 fig, 1€ retf.

Accessicn Number ACAT700021 to ACA710028

 
 

 

Page 20

Category A
Molten-Salt Reactor Frograms

ACA710028 *Continued# ,
Pcrtiocns of the fuel systea were removed for examination as
planned, Control rods, rod thimktles and one ascderator bar
vere taken out and the interior of the reactor vessel was
‘vieweds The fuel sampler cage was cut cut and the punmp bowl
vieved. Porticns of 6 heat exchanger tubes were removed
through a hole cut in the shell. The salt leak was fcund
at a freeze valve and the section was cut out, Conditions
in the reactor vere generally very gqccd. A test cshcwed the
coolant flowmeter had heen reading high and the reactcr
heat balance should have Leen 7.6 MW at tull jcuer.

*examinations + *MSRE + ccres + cutting tools + experience +

flow measurement + heat exchangers + heat fkalance + fumgs +

remote caintenance ,

CTHER CATEGCEIES: MFX

ACB6€CO09

(Staf £ Report) ' -

MSBR DESIGN STUDIES (CHAP 6, FSRP SEEIAXN ERCG REEFT 2/26/66)

Cak Ridge National Laboratory, Tenn. |

ORNL-3936 (June 1966) pp 172-192, 7 fig, 4 ref.
A reference design concert is described for a 1000-MWe
two-fluid MSER with fuel and coolant salts sejarated Ly
grafphite tubes in a 14-ft reactor vessel, Flowsheet,
layouts of the radioactive systewms, and fprccessing by
flucride vclatility and distillation are presented. 2lso
reported are calculated nuclear performance ard ccsts,

*conceptual design + *MSBR + *two-tluid reactor +

breeding performance + costs ¢ flowsheets + layout +

processing

ACB660015

Briqggs RB , |

MCLTEN-SAIT BRFELCER KEACTOR STULIES {PART 2 MSRP SEMIANN

PROG REPT 8-31-66)

Cak Ridge XNational Laboratory, Tenn.

ORNL-4037 (Jan. 1967), fpr. 207-237, 1C fig. € ret.
Design study werk fcr the twc-region, two-fluid 1000 MW (e)
MSBF included adoption of a modular ccncept, using foxr
small reactors to facilitate maintenance, and revision of
the primary heat exchangers to use tent tubes rather than
bellcws, Nuclear fperformances with and without Fa rewoval
are compared. Steam system efficiencies ard ccsts for
700 deq F feedwater are ccmpared to a 5380 deg F syster.
Performance data for other reactcr types are fresented,
including a lead~-cccled MSBR, and eipthermal kreeder, and
a converter with the fertile and iissile materials in a
single strean. Salt prccessing for the fuel and Eklanket
salts is described. The two systeuws are similar, the salt
"being fluorinated, the cff-gas sorbed, aud the uraniur |
tetrafluoride recovered Ly cold-traprinc. A vacuum still-

Accessicn Number ACA7 10048 to ACBGECCIS
Page 21

Category A
Mclten-Salt Reactor Erogranms

ACB660015 *Continued*
separates the rare earths from the remaining salt. Ccncepts

for continuous stills and fluorinaticn units are desciibed.
Liquid-metal extracticn and reductive prec1p1tat10n are
sujgested as alternative methods.
*MSBR + ¥prcgress report + *conceptual design + *proce551nq +
*heat exchangers + *steam systems + *prctactirciur +
*perfcrmance + breeding perfcrmance + electrical power +
thermal power + heat transfer + fuel cycle ccsts +
flowsheets + thermodynamics + design criteria +
*modular design + *lead cool1ng + *two-ilu1d reactcr +
*steamr cycle
CTHER CATEGCRIES: IAC

ACB67C017

Briggs BB '

MOLTEN-SALT BREEDER REACICR DESIGN STUDIES (E2ERT '3 PSEE

SEMIANN EECG EEET, 2-28-67)

Dak Ridge Naticnal Laboratory, Tenn.

ORNL-4119 {July 1967), pp. 174-214, 21 fig, 6 ref.
Design study of the tvo-region, two-fluid 106C( Mk (e) FKSBR
continued. The 250 MW (e) reactor module has a vessel 12 ft
diam with 4-in, diam graphite kalls Letseen tke ccre
elerents and the reflectcr., The revised primary heat
exchangers have the long-shait salt circulatirg fumgs
located abcve the units,  The effect on reactivity of
fissile concentration and fuel-volume fracticr cr the
neutron flux distributicn is explored. An off-gas system
flowsheet is presented and the required gas irjecticn and
remcval system discussed. The effect of xenon removal on
the poison fraction was calculated. Prccessirg cf the salts
in a continucus flucrinator with salt-protected walls may ke
adequate protection against ccrresion. The relative
volatility of the rare earths was investigated and the
equations for buildup of non-volatlles cn vapcrlzan
surfaces are presented.

#MSBR + *progress report +- *conceptual design + ¥prccessing +

*pumps + ¥heat exchanger= + *reactor vessel + *replacement +

*yoid fractions + *volume fractions + pertormance +

graphite + tlanket ¢+ flucrinaticn + corrosion protection +

volatility + xenon + off gas systems + ¥*podular dec:qr +

*two-fluid reactor- _

CTHER CATEGCRIES: IAC

ACB67C024
Brlqgs BB - '
MSBR DESIGN AND DEVLLOP&ENT (EAET 2 FSEE SEFI!hN ERCG FEPT
3-31-67)
0ak Ridge Naticnal Labcratcry, Tenn.
ORNL-4191 (Lec 1967), pp 63-101, 23 fig, 6 ref.
Design study of the two-fluid, twc-recicn, 1C(C ME(e) MSBR

Accessicn Number ACB660015 to ACB67002z4

 
 

Page 22

Categqory 2
Mclten-Salt Reactor Erograms

ACB670024 *Continued*

- using four 250 MW (e) reactor modules involved new cell
layouts to accommodate stresses in piping and pedestal-
mounted equipment due to thermal expansions, The reactor
cell wall construction was studied in mcre detail, The core
grafhite was rearranged to better accommodate dimensicnal
changes due to neutron irradiaticn. Mcre detailed drawings
and data on the fuel and blanket-salt heat exchangers are
presented. Reactor performance was €valuated ir terms cf
the average ccre pouer density, optimized mainly on the
basis of yield, and the fuel-cycle cost estizated. The
useful life of the graphite as a functicn of the neutron
flux is estimated from the LCcunreay Fast Reactci data.
Flux-flattening is discussed and the temperature
coefficients of reactivity calculated. The xencn-135
pbiésgpafhitéipicitedalcDdacédparitaprebleat fenctheidunelp
blanket and coolant-salt pumps are outlined, jatticularly
with regard tc the rmolten-salt journal tearing. ' ,

*MSBR + *progress report + *conceptual cesicn + *pupps +
*heat exchangers + *xenon + *grarhite + *stress + :
thermal effects + tearings + treeding perfcrmance +

design data ¢+ expansicn + fuel cycle ccsts + mass transfer +
neutron flux + neutron fluence + parametric studies +
shrinkage + void fracticns + volume fractions + development +
radiation damage + *modular design + *two-fluid reactcr

OTHER CATEGORIES: IAC '

ACB680013

Briggs RB ,

MSBR CESIGX ANC CEVELOPMENT (PART Z  MSRP SEMIANN PRCG REPT

2-29-68) o

Cak Ridge Kational Laboratory, Tenn.

CRNL-4254 (Aug. 1968), fg. 51-87, 2z fig, 1C retf.
A single fluid concert %as adopted for the two-region
2000-KW (e) MSBR study reference design Lecause it
elirinated the grarhite-tc-metal joints in the two-fluid
concept and kecause means for chenmical frccessirg of a
single salt ncw appeared to be available, Flcw diagrams and
new plant layouts for the single-fluid system are precsented.
Drawings and design data for the single reactor vessel, the
core graphite elements, and the sealt drain tarnk are
included. Tabulated data of reactor physics calculaticns
indicate alrost as gqccd a performance as for the two-fluid
reactor. The effect of use ot ccated qraphite cr the two-
fluid reactor xencn-135 poison fraction is reported. A
conceptual design is shown for a sinqle-fluid MSBR fuel-salt
purp, which dces nct require a salt-lubricated tearing as in
the two-fluid concept. A salt-learing €xperimertal test
lcop and pregram are described, however. Femote main tenance
problems in an MSER plant are discussed. Prelisirary

Accessicn Number ACB670024 to ACB6800 13
Page 23

Category 2 A
Mclten-Salt Reactor Frograms

ACB680013 *Continued* )
results cf analcg ccmputer studies of the dynamics of the
two-fluid MSER are presented

*MSBR + *fprcgress report + *ccnceptual design +

*single-fluid reactors + performance + laycut + dcinpts +

flowsheets + data + reactcr vessel + cores + graphite +°

fuel cycle + neutron physics + coatings + xencn + LURES +

figanlnqschahleténngt;esftrfiene&oymmatntendnce +

CTHER CATEGCRIES: IAEL

ACB68C02C
Briggs KB
MSBR DESIGN AND DEVELOPHENI (EART 2 MSEEF SEEIANN EECG FEPT
8-31-68)
Cak BRidge National Laboratory, Tenn.
ORNL-4344 (Feb 1969) rpp 53-108, 32 fig, 15 takles, 19 ref.
The single-fluid 1000 HW(G) reference plant uses a
confinement building to permit replacement of the reactor
core ‘as an assembly. As shown on new flowsheet and laycut
drawings, the revised reactor has graphite spheres in the
blanket and graphite ccntrol rcds at the center. Details of
a revised primary heat exchanger are presented. Neutron
physics calculations for the revised ccncept weie irmpioved.
Preliminary calculaticns for a 10C-zC0 FW(t) MSER are
reported. The MSBR Xe-~135 poison as furcticn cf bukble
stripping and graphite sealing was calculated ard ccncepts
for a bukble generator and gas ‘separator descrikted. The
MSBE pumping requirements and first cperaticn cf the =scdium
flucrchorate test lcop are discussed, as were the
requirements for a steam generatcr test facility. Analyses
0of the dynanmic respcnse of the MSPR system (and the steaa
generator) indicates general feasibility. Neutrcr decay
after shutdown was calculated., Fesistance thermometers
p0a51bly can be used in the MSBR. Test equipmert fcr .
measuring heat transfer properties of tte =salt are described
"and data given fer therral conductivity and heat capacity.
Test equipment and first data on simulated mass transfer of
xencn to bubbles are covered.
*MSBR + *progress report + *conceptual des1gn + *n—:actcrq +
*heat exchangers + ¥fumgs + *FSBE + *steam generators +
*control + *temperature measurement + *physical Frcirerties +
*gas injection + *gas seraration + *performance +
*heat transfer + bublbles + mass transfer + -
therral ccnductivity + 5§901f1c heat + structures +
malntenance + graphite + control rods + neutrcn phy=1cc +
~xenon + fluoroborates + ccres + reactor vessel +
- single-fluid reactors + test fac111tze= + vcid fracticns +
primary salt e materials testlng + lnstrumentatlon ¢ plplng +
spheres + containment = ' ‘
OTHER CATEGORILS: IAD

ACB590022

Accessicn Number AC8680013 to ACB6Y90022

 
 

 

Page 24

Category 2
Molten-Salt Reactor Frograms

ACB690022 *Continued*

Briggs RB

MSBR DESIGN ANL DEVELOPMENT (PART 2 HSRP SEMIANN PRCG REP1

2-28~-69)

Cak Ridge National Lahoratory, Tenn.

ORNL-4396 (BRug. 1969) pp 49-128, &7 fig, 26 tables, 5( ref.
Design studies of the single-fluid 1000 MW (e) MSEE
continued with emphasis on the reector core and vessel

design, flow and terperature distributions, fission-product -

distribution in the systems, krypton and xench furging, and
the cff-gas system heating loads. The diameter of the
reactor cell was increased and the cell wall ccrstructica
studied in more detail. Changes in the central core
dimensions resulted in increased gqraphite life. Reactcr
afterheat sources, temperature distributions in graphite
core and reflector and in reactor vessel are plctted.
Develcpment work includes methods for bubble generaticn and
gas separation in fuel-salt system. Distributicn of roble
metal fission products is tabulated. Improved values were
~obtained for thermal conductivity cf the fuel salt and an
experimental lccp tc ccnfirm heat transfer relationships has
furnished preliminary data.: Opera icn c¢f the scdiux
flucrcborate test 1cof is described. A successful rerotely-
operated orbital welder for piping is reported. 1The
ccntrcls systen studies continued., EFreliminary drawlngs and
descriptions of the MSEE are included.

*MSBR + *rrcgress regort 4+ ¥conceptual design + *reactors +

*MSBE + *heat exchangers + *pumps + *steam ceneratcrs +

*physical froperties + *ccntrcl + #*gas injection + \

*gas separation + *performance + *heat transfer + *cells +

*test facilities + thermal insulation + bubbles +

mass transfer + thermal conductivity ¢ structures +

welding + maintenance ¢ ccres + reactor vessel +

noble metals + fission products + neutron physics ¢ xencn +

kryptcen + fluorcbcocrates + decay + heat + graphite ¢

¥*steam systems

OTHER CATEGORIES: IAD

ACB690029

Briggs RB

MSBR CESIGN ANLC DEVELOPHENT {PART 2z MSRP SEMIANN PRCG REPT

8-31€9) - :

Cak Ridge kational Laboratory, Tenh. . :

CRNL-4449 (Feb. 1970) pr 39-95, 41 fig, 1Z tatles, 3& ref.
Ccnceptual study cof a single-fluid 1000 MW (e) reference _
design MSBF is essentially corplete. Princigal desigr data
are tabulated. The fplant layout was revised to include a
domed confinement building which provides rpissile prctection
and acts as containrent during maintenance., ? waste storage

cell is also provided. Seismic disturlances sere ccneidered-

in the design., Laycut drawings are shoun tor all building

Accession Numker ACBE€9C0zZz tc ACE6SC(CZ9
Page 25

~ Category R
Molten-Salt Reactcr Prograsgs

ACB69C029 ¥*Ccntinued*
~ levels, The primary drain tank was revised tc use a
'flithium—berylllum fluoride salt-to-water-to-air cooling

system. Nuclear calculations were refined tc irclude effect
of plant size and tc consider alternate reactcr designs.
Gamma and neutron heating was calculated for the tefetence
design gecmetry and alsc for an FSBE with spherical vessel.
The industrial program to develof a stear geheratcr is
discussed., Results cf cperation of the sodium tluorotorate
test are reported., The requirements for the MSBE salt pump
test stand are covered. . Results of heat transfer and salt
physical property studies are reported in scie detail. The
mass transfer test facility is completed and experimental
. 'work started,

*MSBR + ¥[rogress report + *ccnceptual design + *MSEE +

*pumps + *steam generdtors + *drain tanks +

*physical froperties + *heat transfer + *test facilities +

*performance + *containment + *cells + rass transfer +

structures + welding + maintenance + *control +

neutron physics + gas injection + gas separation +

single-fluid reactors + rrimary salt + thermal conductivity +

capture + absorption + earthquakes + dynamic characteristice +

radiation heating + layout + flowsheets + data +

wvaste disposal = :

OTHER CATEGORIES: IAD

ACB700022
Briggs RB : ‘ ~ ' ‘
MSBR DESIGN ANB DEVELOPHENT {PART z MSRP SEMIANN PRCGG REPT
2~ 28-70) . -
Cak Ridge Kational Lahoratory, Tenn. S
ORNL-4548 (Aug. 1970) pp 41-92z, 25 fiqg, 15 takles, 32 ref.
Studies cf the reference design for 1000 MW{(e) single-
fluid MSER were completed and the first draft cf a regort
~circulated. Principal design data are Fresented. Studies
are being made of first- -generaticn tyres of nclter-salt
reactors that would have poorer performance ktut would
require less development, -including a large MSRE type and a
spherical reactcr with graphite ball bed. A primary heat
. exchanger with bhayonet tules is. ccmpared tc ttke reference
. design exchanger, The tritiua distribution in - a 1000 MW (e)
MSBF was estimated and the eifectiveness of varicus methods
~of reducing the arcunts reaching the steam system were -
calculated., The nuclear physics calculaticns were refined,
including estinates of the control rod worth. The stean
‘generator development program is discussed and further
tests from the sedium fluoroborate test loop reported. The
- pump test stand is descrlted and the renetely-cperated
orbital welder for fpiping discussed in some detail.
Simulation studies of dynamic resronse cf MSBR cchntrols
systems are presented. Development work was continuedé on

Accession Numker ACB639CC2S to 'ACB700022

 
 

 

‘Fage 26

Category A
Molten-Salt Reactor: Frograms

ACB700022 *Continued¥
gas bubble generaticn and separation fronm ‘the fuel salt.
Better values for the thermal conductivity cf the salt and
for heat transfer relationships were obtained from the
experimental results,

*MSBR + ¥progress report + #%ccnceptuval design + *MSEE .+

*pumps + #*steam generators + *converters + *pkysical frcperties +

*heat transfer + *test facilities ¢+ #*performance + #*control +

welding + mass transfer + thermal ccnductivity +

maintenance + neutron physics + gas injection + .

gas separation + dynamic characteristics + graphite +

spheres + *tritium + develcprent + components

CTHER CATEGCRIES: 1IAC

ACB70C€036

Briggs RB

4S8R DESIGN AND DEVELOPMENT (EART 2 MSKE SEFTIANN EEFCG EEPT

8-31-70)

Oak Ridge Naticnal Labora tory, Tenn. '

ORNL-4622 (Jan. 1971) pp. 7-59, 43 fig, 11 tables, 35 ref.
With completion of the report draft cn thke single-fluid
MSBR, ORNL directed major attention to FSRF technical
probleas but some studies continued c¢cn a demcrstratior
plant and plans prcqressed for an industrial study of a
large MSEF station. Flowsheets and laycut drawings aire
shown for a 3CC MW(e) demonstration plant witihn low enough
povwer density for the graphite to not regquire rerlacernent.
The primary heat exchangers are mounted horizontally to
permit maintenance from the side, and detailed afterheat
studies cn an empty exchanger are reported. The drain tank
uses a natural convection NaK cooling syster. Nucleal
physics studies continued on cores of lcw power density and
for Th concentrations in the 10-18 rcle % rance with fuel
cycle costs and yields tabulated. Eatch processing wés also
considered. Capture cross sectiocn ratics fcr algpha fcr
U-23% wvwere deterrined experimentally. Eubkile generator and
gas separator testing is reported. Plars prccressed fcr
industrial study of a steam generator. The socdium
fluoroborate loojp testing included water irnjecticrn with
inccnclusive results, Femote welder development emphasized
consistentl]ly good welds without direct cbservaticn cr manual
adjustment, Partial load steady-state tehavior of MSER was
studied., Heat transfer tests and investigaticr cf
thermophysical fprcrerties continued. [ata on transfer
‘coefficients to helium ftulkles are reported.

*MSBR ¢ ¥MSBE + *fprogress regcrt + *conceptual design +

#*physical properties + *gas separation ¢+ *heat transter +

*test facilities + *industrial studies + *performance +

*neutron physics + #*welding ¢+ mass transfer + reactcrs +

heat exchangers + drain tanks + structures + layout +

maintenance + control + decay ¢+ fission products +

Accessicn Number ACB7000zz to ACB7C0C2e
Page 27

‘Category A
Mclten-Salt Eeactor Frogranms

ACB700036 *Continued¥* .

graphite + pumgs + coptainment + earthguakes +
fluoroborates + bubbles + gas separat1on + conponentc +
development ‘ . S

CTHER CATEGCRIES: IAC

ACB710029

Briggs RB

MSBR DESIGN AND DEVLLOPMENT ([AET II, ESRP SE!IANN ERCG

REPT 2/28/71)

Oak Ridge Naticnal Labcratory, Tenn.

CRNL-4676 (BAug. 1971), pp 21-72, 27 fiq, 52 ref.
Ccnceptual design of a 1000-MW MSBR was corpleted and
-design studies of a large, 300-F¥e demonstration reactor
vere started., Flowsheets, layouts and ccmpcnent design
data for this reactcr are presented. Conceptual desicn of
a high-power-density, 150-MWth, molten-csalt breeder
experiment (MSBE) also was pursued to define development
requirements, Development efforts focussed cr ccclant
syster technolcgy and the removal and handling of gaseous
fission products from the fuel. Plans rrogressed fcr
industrial studies cf cteam gone:ators and a 1000-MK
MSBF plant, -

¥conceptual design + *develcpment ¢ *KSFE +# *PFSEF +

analysis + converters + ‘coolants + design data +

flowsheets + gas separaticn + industrial studies +

progress report + steam generators + tritium

OTHER CATEGORTIES: HEX + IAD + IAE

ACC6500086
Lindauer RB
FUEL ERCCESSINKG (PART 7 HSRP PROC RE?T 4/48/6‘)
Oak Ridge Naticnal Labcratery, Tenn, -
CRNL-1812 (June- 1965), pp. 169-171, 2 tig. '
The design, frccurement and c*nstruct1cn ct the MSRE fuel
. processing systen were essentially completed except fcr the
. salt sampler and the uranium akscrpticn equiprert. AT
~ electrolytic hygrometer is being tested for in-line
“'monitoring of the removal of cxide trcm mclter salt by
treatrent with hydrcqen and hydrogen fluoride., 1Initial
" 'results are encouraging, tut they indicate that hyarogen
- flucride will have to te completely removed from the gas
that is bypassed to the apaly:er. Study of metheds fcr the
remcval cf volatilized chromium fluoride irom the offgas
stream during fluorination of rolten salt Ltas begun, Scme
data have been obtained for the sorption of chromium
. trifluoride on sodium fluorxdu pellets at u4C( deg C.
¥MSRE + ¥*fprccessing + *ccnctructlon + absorption +
corrosion products + design + hydroqen compcurds + cx1des +
sodiur flucride + uranium
CTHER CATEGCEIES: LHX + MBX

ACC65C012

Accession Number ACB700036 tc ACC650012
 

 

Page 28

Category A
Molten-Salt Reactor Erograms

ACC650012 *Continued*

Lindauer RB

FUEL ERCCESSING {(EBART 7 HSRP PROG REPT 6/31/6‘)

Oak Ridge Naticnal Lahoratory, Tenne.

CRNL- 3872 (December 1965), p 152, 3 ref.
Construction of the MSRE fuel—processxnc syctem ¥as
completed, the system was tested, and the flush salt was
processed for oxide removal., Operaticn of the jlant ias
generally satlefactcry, and about 115 ppm of ozxide was
removed from the fiflt in reducing the ccncentiaticn tc about
SO ffre. .

*MSRE + *oxides + *processzng + constructlcn + operatlcn +

plant ,

GTHER CATEGCBIES' LBX + MCLC

ACC66C010
(Staff Report) |
MOLTEN-SALT REACTOR PROCESSING STUDIES (Part 7 NSEE Frogre.
Rept 2/28/66) .
Oak Ridge Naticnal Labcra tory, Tenn.
- CRNL-3936 (June 1966) pp. 193-211, 10 fig, 6 :ef. :
A close-coupled facility for processing the fuel and fertile
streags will be an integral part of an FSBF systen, Fuel
will be processed on a 40-day cycle., Uraniur will bb
flucrinated frcm the carrier salt vhich will then be
recovered from fission fproducts ky distillatico. relative
vclatilities between lithium and rare earths have been
measured to be 0.001 to 0.04 at 9CO0 to 1C5C deg C. Uranium
hexafluoride will be abtsorbed in fuel salt containing
~uranium tetrafluoride and then reduced with bydrcgen.
"Flucrinatcr corrosicn can probably be eliminated ty a layer
of frozen salt on the wall., Experimental work with a small
ccuntercurrent continucus fluorinator gave recoveries of 90
to 96% of the uranium. Volatile chrcriuve flucrides can bhe
trarred with negligible uranium losses on sodium fluoride
beds. A& preliminaiy design study of the above facility has
illuminated protlems among¢ which is hancéling Ligh—-hea t-
generating materials. The fixed capital cost fcr the
conceptual plant was $5.3 million; the =salt irvertory ccst
was $C.166 willicn, and the direct operating cost was
$787,790 per year. : -
*4SBR ¢ *frocessing + corrosicn protectlon + costs +
design + distillation + fluorination + lithiup +
rare earths + sodiuam flucride + uranium + volat111ty +
two~fluid reactor ,
OTHER CATEGORIES: LJX

ACC660016

(Staff Repcrt) o

MCLTEN-SAILIT REACTOR PEOCESSIYG STUDIES ;PART ¢ MSRP PROG
REPT 8/31,66)

Accession Numbker ACC650012 to ACC660016
-}

Pagé 29

Category a
Molten-Salt Reactor Prograrms

ACC66C016  *Continued*

Cak Ridge National Laboratory, Tenn. -

OBRNL- 4037 (Janvary 1967), pp. 227-237, § fxg, 2 ref.
The MESBR processing plant would use cycle times cf 40 days
for the fuel salt aprd 20 days for the fertile salt. CUsing
a recirculating equilibrium still relative vclatilities have
beepn cbtaibed which are a factor of £( lower than usirg a
ccld-finger technigque. Uranium recoveries exceeding 99%
have been attained with continuocus fluorinatcrs cnly 48 in,
high. Ccrrcsicn prctection by means of a frozen wall is
being studied. Studies continued on alterrative frccessing
methcds to replace vacuum distillation. Tests were made
with the reduction-coprecipitaticn frccess using beryllium
and with the liquid-metal extraction prccess using solut1ons

' of lithium in tismuth, : : \

*MS3R + *precessing + berylllum + bismuth +

distillation + flowsheets + fluorinaticn +

lithium + reductive extract1cn process + volatllxty +

corrosion protection L , : :

OTHER CATEGORIES: LJX

ACC670018
(Staff Repcrt) -
MOLTEN-SAIT REACTOR PROCESSING STUDIES (PART 1C MSRP PROG
REPT 2/2€,67) , _
Cak Ridge National Lakoratory, Tenn.
CRNL-4119 (July, 1967), pp. 204-213, 5 fig, 2 ret.
Studies cn the flucrlnatlon-dlctlllatlon tlowsheet for HaBR-
processing continued, Fluorinaticn studies with
ncngrctected systems using 1-in.-diam towers have
demonstrated steady state recoveries up tc 6%.5% cf the
uraniur with flucrine utilization of 15%. Studies on
~column protection involve the censtruction cf a S5-in.-diaun.
~nickel tcwer with rrovisicn to generate heat fluxes tc
create a frozen wall of salt., HRelative volatilities
measured at 1000 deyg.C and 0.5 mm mercury pressure were
3 x 10(-3rd), 3 x 10(-4th), 6 x 1C(-4th) and 2 x 10 (- 4th)
- for ceriur, lanthanum, neodymium and samarium trifluorides
© with respect to lithium fluoride. Equifprert is Leing
‘fabricated for the distillation of 48 liters cof MSKE fuel
salt after removal of the uraniurm by flvorinaticr,
HSBB + *rrccessing 4+ distillation # fluorination + '
fuels + MSFFE + rare earths + urdnlum +. volat111ty

ACC67002‘ o
{(Staff EReport)
MOLTEN-SALT PROCESSING AND PHEPARAT[CN (PAET 6 HSFE PFCG
" REBT 8/31/67) ,
Oak Ridge Naticnal. Laboratcry, Tenn.
CRNL-4191 (Lecember 1967), pp 239-253, 6 f1g, 10,ref.'
Most of the effort in this periocd was cn the distillaticn

Accessicn-Numbe:,ACC660016 to ACC670025

 
 

 

Page 39

Category A
Molten—Salt Reactor frograms

ACC670025 #*Continued* - o
step in the fluorination-distillation flowsheet. ‘Rel ative
volatllltles of rare earth fluorides wvere reasured using
both an equilibrium still and the transgpiration method.
Data from the two methods is in gcod agreerent. Egquiment
for demonstraticn cf vacuum distillation using MERE fuel
salt is teing installed in a test facility for
non-radioactive experiments Lefore operation with MSRE salt.

A conputer ccde has been prepared to provide information on

fission product heat generation. An alterpative fprccess to
distillation, reductive extraction of rare earths using
lithium reductant in bismuth is lteing studied.
Modifications are being made to the MSRF fuel processing
facility to permit uranium recovery ky flucriraticn after
only 35 days decay. Design and equipment fabrication is in
progress for preparing 40 kg cf uranium-233 as the :
uranium-lithiur flucride eutectic for replacement of the
present MSEFE uranium fuel.

*M5BR + *prccessing + *distillation ¢ volatility +

rare earths + reductive extraction process + MSRE +

uranium-233 + flucrinaticn + lithium + bismuth

CTHER CATEGORIEFS: LCA

ACC680014

(Staff Refpcrt)

MOLTEN SAIT PRCCESSING ANT PBEPAEATION {PAR1 € MSRP PROG

REPT 2,25/68) -

Cak Ridge National Laboratory, Ienn. - :

ORNL-4254 (Auqust 1968) rf. 241-277, 1€ fig, 18 ref.
Distribution coefficients were measured for uranium, thorium
and rare earths Letwveen molten fluoride salts ard litkium-
bisguth soluticns. Calculations were made for the
isolation of protactinium from a single~ftluid MSBR. Studies
are undervay cn prctecting a continuous fluorinator from
corrosion by freezing a layer of salt cr the sessel wall.
Relative volatility measurements were made for uranium,
rubidium, ceaesium and zirconium fluorides with resgect to
lithium flucride. Fcur non-radiocactive test runs vere made .
with fuel carrier salt in the distillaticn unit tc be used
with the MSRE fuel salt. Reductive extracticn processes for
protactinium removal vere evaluated. Spmall scale
flucrination tests vere made with sinulated MSFE fuel salt.
Preparation of the uranium-233 fuel ccncentrate fcr the MSRE
is underwvay. Decay heat from fission products and
protactinium has been calculated ftor a «CG(-Mw =single-regyion
MSR. : ,

*MSBR + *frocessing + distillation e distritution +

fluorinaticn ¢+ MSRE + rare earths +

reductive extraction process + protactinium + thcrium +

uranium + uranium—233 + vclatility -

ACC680021

Accessicn Number ACC670025 to ACC6800:21
"

Page 31

Category 1A
Mclten-Salt Reactor Irograns

ACC680021 *Continued*

{Staff Refport)

MOLTEX SAILIT PRCCESSING ANE PRFPARATICN (PAET 6 MSBF FECG

REPT 8/31/68)

Oak Ridge Naticnal Laboratory, Tenn.

CRNL-4344 (February 1969), pp. 291-326, 22 fig, 13 ref.
Measurement of distributicn coefficients for the reductive
extraction process continued. The solutbility cf
protactinium and thcriup in bismuth was determined.
Simulated molten salt-liguid tismuth ccntactcr studies were
started with wmercury and water. Equipment is keing
installed for semicontinuous experiments on reductive
‘extracticn. A series of experiments was concluded which
demonstrated the feasitility cf a frczer salt wall for
corrcsior protecticn during fluorination. Equipment is
being installed at the MSRE for demonstraticr cf fuel s=alt
distillation. Relative volatility measurements were made
vith thorium fluoride. Preparaticn cf the urarium 233 fuel
ccncentrate for the MSRE was completed using the two—-step
process., Development of this process is described. Final
labcratory tests were made on several steps in the prccess
for recovery of uranium from the HSRL (descrzbed in Part 1
of this repcrt). o

*MSBR + *processing + corrosiocn protectlon + distillaticn +

distributicn + flucrinaticn + protactipium +

reductive extraction process + MSRE ¢ thorium + urarivm +

uranium-233 + veclatility

ACC6490023
{(Staff Refpcrt)
MOLTEN-SALT PRCCESSING ANL PREPARATION (PART € MSRP PRGSG
REPT 2/28/69)
Cak Ridge FKational Latoratory, Tenn. '
ORNL-4396 (Auqust 1969) fp. 270-299, 2% fig, 33 ref.
- The proposed reductive extractxon processing flcwsheet for
a single-fluid MSBR is descrited. Protactiniur and rare-
- earth reroval is included. A computer code has Leen
" developed to perform the necessary material talance
- calculations. The measurement of distribution coefficients
for the reductive extraction pProcess ccrtinued: The nutual
~ solubilities of nickel and ‘thorium in bismuth were
- determined. Experlments were carrled out usirg guartz
electrolytic cells. Preliminary testing of equipment for
. semi-continuous experiments on reductive extracticn has-
begun, Data is regpcrted cn mercury-water tests in columns
to simulate molten-salt- liquid-kismuth, Ccld tectlnq cf the
distillaticn unit at the MSRE ®as completed prlcr to
'.dxstllllng a portion of the MSRE fuel salt. o '
*HSBR + *prccessing + bhisputh + computer codes +
distillation + distribution + electrolysis + flowsheets +
protactinius + reductive extraction process + MSEE +
solubility + thorlum

Acce551cn Number ACCH3002z1 to ACC690643

 
Page 32

Category A
Mclten-Salt Reactcr Programs

ACC69C023 *Ccntinued*
CTHER CATEGCEIES: LKX

ACC69C030
(Staf £ EFeport)
MOLTEN-SALT PROCESSING AND PREPARATICN (PAFT 6 nsas FECG
REPT 8/31/69) :
Oak Ridge Naticnal Laboratory, Tenn.
CRNL-4449 (Feb. 1970) pp. 214-246, 27 tig. 26 ref,
Measurement of distribution coefficients in fclten salt-
metal systems ccntirued and data is presented for
transuranium elements, rare earths and thoriur. T1he
sclubility of plutcrium fluoride was measured in lithium-
beryllium fluoride salt. Flowsheet analyses nere made of
protactiniux isclaticn, rare-earth removal, thorium
" stripping, fission product comcentrations and heat
generaticn rates. Four runs were made with the semi-
continuous system for contactirg tiswutt with mclten =alt.
Electrolytic cell and salt-metal contactor development
continued, Axial mixing wvas studied in both fpacked atd
bubtle cclumns. Abcut 11 liters of the MSEE fuel carrier
salt was distilled. Design studies were carried cut cn:
(1) beat transfer through the frozen salt walls of an
electrolytic cell, (2) a contibucus salt purificaticn
system and (3) plutcniur capsules for refueling the MERE.
*MSBR + *processing + bismuth ¢+ columns + design + o
distillaticn + distributicn + electrolysis + flowsheets +
heat transfer + MSRE + rlutonium + protactinium +
reductive extraction process 4+ solultility + rare earthks +
thorium

ACC700023

(Staff Refcrt)

MOLTEXN-SALT PRCCESSING ANL PREPARATION (PART € MSEP PRCG

REPT 2/2¢8,70)

Cak Ekidge National Laboratory, Tenn. | . _

ORNL~-4548 (August 1970) rp. <77-33z, 42z £ig, 233 ref.
A new processing flcwsheet for a single-fluid MSEF is
described. Electrolytic cells are elicinated by the tse of
a metal transpcrt frccess for removing rare earths anc
fluorination followed Lty reductive extracticn fcr
protactiniumr isclation. Distribution of thoriur and rare
earths between lithium chloride ard tissuth is being
studied in suppcrt c¢f the metal transfer process. Four more
runs were made with the semicontinuous syster fct
contacting bisguth with molten salt. Eguipment is Leing
prepared for a demonstration of the metal trarsgcrt
process. Contactor and electrolytic cell development is

continuing., D[ata from the distillaticn of MSRE fuel carrier
salt is fpresented. Material and energy balance calculations

and calculations on the effect of chemical frccessing on

Accessicn Number ACC690023 to ACC700023
Page 33

Category A
Mclten-Salt Reactor Erogranms

ACC700023 ¥Continued* |
nuclear perforrance were made for the MSBE processing plant.
Installation of continuous salt purificaticn equigment is in
progress., Specially designed capsules vwere loaded with
plutonium fluoride end added to the MSRE fuel salt.

*M3SBR + *rrccessing + bisguth + columns ¢+ design +

distillation + distribution +. electtolySis + flcwsheets +

metal trarsfer prccess + MSRE + plutonzum + rare earths +

reductive extractlon Process

OTHER CATEGORIES: LEKY

ACC700037

(Staff Regcrt) ' ' b

MOLTEN-SAILT PRCCESSING ANL PREPARAIION {PART £, MSRP

SEMIANN PROG REPT 8,31,70) ' '

Cak Ridge National Lakoratory, Tenn. - o

ORNL-U4622 (Jan. 1971), pp 199-224, 2C fiq, 27 ref.
CalCulations were Made fof a flo¥sheet using
fluorination-reductive extraction for Pa isclaticp and
metal-transfer for rare-earth removal. Calculations were
also made on removal of uranium ty oxide precipitaticr.
More data were obtained on distribution of rare earths and
thorium between bismuth soluticns and mclten =alts, '
Engineering developrent included operation of the
flow-through reductive-extracticn facility, tests on
perfcrmance of packed cclumns wsith two liquids differing
widely in density, demonstration cf the metal-transfer |
process, and experiments with electrolytic cells. Tests of
a continuous salt purification systen were started.

*development ¢ ¥*experiment + *MSBR ¢ #*processing +

bismuth + chlorides + columns + date + distrituticn +

electrolysis + flowsheets + fluorination + / :

metal transfer process + ox1de prec1pxtat1cn froce +

reductive extracticn pxcce ‘

ACD650007
(Staff Repcrt)

"RADIATICN CHEMISTRY (CBAP 5, HSRP SEMIANN PROG KEPI 2/48/65)

Oak Ridge Naticnal Labctatory, Tenn.
CRNL-3812 (June 1965), pp 87-120, 25 fig, 5 ref. :
| ' In-pile cafFsule tests in the MTR were ccrpleted and
-~ post-irradiaticn examinations at CRNI were practically
~ finished., Early tests had showed effects cf flucrine
- -evoluticn., Later tests, shich included gas connections
-~ and external heating during reactcr shutdown, fprcved that
the fuel was stable, with no fluorine evoluticn, under
operating conditions.  {Radiolysis of ccld salt had
- produced the gaseous tfluorine.) Examination of salt,
- graphite and INOR-8 from the capsules slowed rc
radiation-induced ircomratibility.
*examlnatlons + *jin-pile tests + *radiclysis +

Accessicn Number ACC70002Z3 to ACL65CCC7

 
 

 

Page 34

Cateqgory A
Molten-Salt Reactcr Prograts

ACD65C007 *Ccntinued® |
capsules + compatibility + fluorine ¢ graphite +
Hastelloy N + golten salts + frogress report

ACD650011

(Staff Rerort)

CHEMISTRY (CHAF 6, MSRP SEHIANN PROG REPT 8/31/6 )

Oak Ridge Naticnal Labora tory, Tenn.

CRNL—’B?Z (Dec. 1965), pp 111-151, 16 £fig, 40 ref.
Apnalyses of MSEE salts dur1ng precritical testirg, U-235
loading, and zero-fpcwer experiments showed that purity was
maintained and corrosion was very low. Vapcr gtressures, HF
solubility, and iodine removal in LiF-BEef2 systems were
determined. Phase relations in the NaF-NaBF4 syster and
viscesity of NaBF4 fiererdetermined. (This systen is
suggested as an inexpensive, lower-melting breeder
coolant,) Preparations were rade for studyinc Pa oxide
precipitation. Effcrts continued to improve analyticel
methods for MSRE salts and cover gas.

*analytical chemistry ¢+ *chemistry + *experience + *MSEE +

*progress report + data ¢ experiment + flucrides +

fluorcborates ¢+ iodire + polten salts ¢+ oxides ¢+

precipitation + protactinium

OTHER CATEGORIES: MCD + CXX + DXX

ACD650013

(Staff Refpcrt)

RADIATICN CHEMISTEY (CHAP 5, MSRP SEMIANN PROG REPI &/31/65)

Oak Ridge Naticnal Laboratcry, Tenn,

CRNL- 3872 (Dec. 1965), pp 106-110, 2 fig, 2 ref.
Design and development proqressed on an in-pile mclter-salt
experiment to gc in the CER. It consists of a compact
thermal~-circulation loop of INOER-& including a 2-inch
graphite ccre and 85 cc of fuel salt.

description + development + in-pile tests

ACD660O1T1

(Sstaff Refpcrt) |

CHEMISTRY (CHAE S5, MSRP SEMIANN PROG REPT 2,28&/6€)

Qak Ridge Naticmnal Labcratory, Tenn.

CRNL-3936 (June 1966), pp 122-171, 25 fig, 32 ref,
Inproved analytical methods apglied tc MSRE fuel =amplec
showed nc anomalies and excellent purity. Flugging
material in the offgas line proved tc ke o0il deccrposition
products. Studies c¢f trhysical chemistry of molten fluoride
and the chemistry of Pa and fissicn fprcduct extractior
continued. The latter included distillation, reductive
extraction into liguid metals and oxide precipitaticn,
Fabricaticn fprcgressed cn a molten-salt loop to go in the
CRR.,

*analytxcal chemistry + *chem1=try + *experience + 1MSPE +

Accession Number ACL6S00C7 to ACDE6CC11
-

Page 35

Category A
Molten—-Salt Reactcr Prograss

ACD66CO11 *Ccntinuedx

data + distillation + in-pile tests + ligquid getals +
molten salts + off-gas systems + oxides + precxpltatlon +
protactinium + rare earths + reduct:cn ,

OTHER CATEGORIES MCD +. DXX + ICA + IDA + CXX

ACD6G0017 -

(Staff Report)

CHEMISTRY (CHAE 7, MSBP SIHIANN PROG REPT 8/ 1/66)

Oak Ridge Naticnal Laboratory, Tenn. |

ORNL-4037 (Bug. 1966), pp 134-200, 24 fig, 42 ref.

' The table of contents. is as fcllcws, Behavicr cf Fuel
and Ccclant Salts in MSBE. Phy51cal Chemistry cf Fluoride
Melts; Viscosity and Censity of Molten Beryllius Fluoride;
Transpiration Studies in Support of the Vacuum LCistillation
Process; Fstimated Thermophysical Prcrerties c¢f MSBR Coclant
Salt, Separaticn ip Molten Fluorides: Extracticn of Rare
Barths from Molten Fluorides intc Molter Metals; Removal
of Rare Earths frcm Molten Fluorides by Simultaneous
Frecipitation with UF3; Removal ¢f Protactiniur froro
Mclten Fluorides by Oxide Precipitation; Extraction ot
Protactinium from Molten Fluorides intc Molter Metals;
Protactiniup Studies in the High-2lpha PFolten-Salt
‘Laboratory.  -Radiation Chemistry: Xenon Diffusicn and
Possible Fcrmation ¢f Cesium Carbide in an MSEF; Fission
Product Eehavior in the MSRE. Developrent and Evalua tion of
Analytical Methods fcr-Mclten~Salt Beactors: [etermination
of Cxide in Radioactive MSRE Samples; Sgpectrcrhctcmetric
Studies cf Molten-Salt Redctor Fuels:; Voltammetric and
Chronopotentiometric Studies cf Uranium in Mclter
LiF-BeF2—ZrF4;-In—Line test Facility; Analysis of Helium
Blanket Gas,  Development and Evaluaticn ot Ecu1puent and

- Procedures fcr Analyzing Radioactive KSEE Salt Sample

Samples Analyses; Quallty—Ccntrol Prcgram.

*analytlcal chemistry + *beryllium. fluoride + *capsules +

*experiment + #*fission products + *graphkite +

*hydrccarbens + *MSRE + %ncble metals + *oxides + ,

*physical properties + *rare carths + *xenon + actirides +

analysis ¢+ behavior + borcn trifluoride + cartides + cells +

c1rcu1at10n +. compat1h111ty + 'concentration + cores +

corrosicn + corrosicn preducts + cover gas + decay + -

density + deposition + diagrams + fllSSGlVJBQ + d1=t111at1cn +7

electrical progerties + ertrainment + eguilibrium +

examinations + fission + fluorides fluorckcrates +

gamma radiation + gamma spectiometry ¢+ gases # Hastelloy N ¢
inert gases + lithium fluoride + mplten salts +

oxide precipitation process + rphase eqguilibria + '
progress report + protactinium + protactinium tlucrides +
rare gases + research + samrling + specific heat + -
uranium fluorides + vapor pressure + viscosity

ACD67C019

Accession Number ACD6ECC11 to ACD67CCIS

 
 

Page 36

Category &
Molten-Salt Reactor Prograims

ACD67C019 *Ccntinued®

(Staff EReport) : _

CHEMISTRY (CHAP. 7, MSRP SENIANN, FEOGR. REPT. 2/2E,€7)

Oak Ridge National Laboratory, Tenn,

ORNL-4119, (July 1967), pp. 118-166, 13 fig, 39 ref.
The following tcrics are included in the table of
contents: Chenmistry of the MSRE, Fuel salt ccmpcsiticn
and purity, MSRE fuel circuit corrosion, Extent of UF4
reduction duvring MSRE Fuel preparation, Adjustsent cf the

UF3 concentraticn in the MSRE fuel salt, Fission product

behavior in the MSRE, long-term surveillance specimens,
Uranium analyses of graphite specimens, Fuel =alt samiles,
Effect of operating ccnditions, Effect of beryllium
additions, Pump bowl volatilizaticn and plating tests,
Urapiuc cn pumf bowl metal specimens, Freeze valve capsule
experiments, Special pump Lkovwl tests, General discussicn
of tissicn product behavior, Fhysical chemistry of
fluoride melts, The oxide chermistry ¢f LiF-BeFZ—-ZrF&
mistures, Sclubilities cf SmF3 and NdF3 in HMolten
LiF-BeF2 (66-34 mole %), Possikle MSBR Llanket-salt
mixtures, Serparaticns in solten fluorides, Femoval of rare
earths from molten fluorides ty precigpitaticn cr solid
UF3, Extraction of rrotactinium from molten fluorides into
molten metals, Extracticn of rare earths frcr nmclten
flucrides into molten metals, Frotactinium studies in the
high-alpha molten-salt laloratory, Preliminary study cf
the system LiF-ThFU4-PaFl4, Development and Evaluation of
analytical methods for rolten-salt reactors, PDeterminaticns
of cxide in MSRE salts, Determination of U (3+)/U(4+) ratios
in radioactive fuel bty 2 hydrogen reduction methcd, ENMF
measuresents cn the Nickel-Nickel (IT) couple in Molten
fluorides, Studies of the anodic uranium wave in molten
LiF-EeF2-2rF4, Spectrophotometric studies c¢f zoclten-salt
reactcr fuels, Analytical chemistry analyses of radioective
MSRE fuels, Sample analyses, Cuality ccertrcl prcgrarc.

*analysis + *analytical chemistry + #*cells + tchemistry +

*fuels + *graphite + *notle metals + *cxides +

*protactinium 4+ ¥protactinium fluorides +

*reductive extraction process + *sampling +

*surveillance + berylliur flucride + bismuth ¢ capsules +

chemical properties + chemical reactions + cospatibility +

concentraticn + ccrrcsicn + ccrrosion products + o

electrolysis + gamma spectrometry + hot cells +

in-pile tests + liquid metals + lithium fluoride +

materials + materials testing + metals + mclyltdenuz +

MSRE ¢ mnickel + nickel alloys + progress report +

rare earths + secondary salts ¢ testing +

thorium fluorides + uranium fluorides + zirconium flucride

CTHER CATEGCEIES: CXX + [XX +¢ MCD

ACD67C020

Accession Numkber ACD67C01S to ACD67CCZ(C
3

Page 37

7 Category &
Holtefi-Salt Reactor Prograss

ACD67C020 ¥*Continuedx

(Staff Report) : L ' o R - ]

CONVECTION LOOP IN ORR (LHAP. '8, MSEP SEFMIANN, PROGR.

REPT. 2,28/67)

cak Ridge National Laboratory, Tenn. ’ '

ORNL-4119 (July 1967), pr. 167-173, 0 fig, 4 ref.
Irradiation of the first rmolten salt corvecticn 1cof
experiment in the Oak Bidge BResearch EReactor was terminated
Aug. 8, 1966, after development cf 1.1 x 1C(1€)
fissicnsycc (0.27% U-235 burnup) in the LiF-EeF2-Zr FU-UF4
{65.1€6-2€.57-4.90-1.36 wole %) fuel. Average fuel power
densities of up to 105 w/cc were attained in the core,
.which vas made of MSRE Grade graphite. The takle of
contents for the report on these experieents ccrtains the
following topics: Cbjectives and description, First loop
experiment, In-pile irradiation asseskly, Oferaticns,
Cherical analysis ¢f salt, Corrosion, Fission products,
Nuclear heat and neutron. flux,_Hct-cell exariraticn of
compcnents, Evaluation c¢f system performance, Heaters,
Coolers, Temperature control, Sarxpling and additicn, falt
circulation, Second in-rile 1rrad1at10n assemkly, Operation.

*experiment + *fission products + *gammpa radiaticn +

*graphite ¢+ *Hastelloy N ¢+ *leaks + noble metals +

*thermal convection + actinides + analysis +

heryl]ium fluoride + circulation + compatibility + cores +

corrosion ® decay + dismantling + examinaticns + fissicnp +

fluorides + wolten salts + gamma spectrometry + hot cells +

lithium fluoride + materials testing + MSKE +

progress report + radiaticn damage + research + samgling +

stress + stress rupture + uranium fluorides + uraniuvm—-235

ACD67C026
Grimes WR . - - e T | -
CHEMISTRKY (PART 3, MSRP- SEMIANN. PROGR. REPT. 8/731,€7)
Oak Ridge:Naticnal-Laboratcry, Tenn, = ° : .
GRNL -4191 (Pec. 1967), pp. 102-175, 39 fig, 6€ ref.
‘Sampling of the MSRE fuel dnd coolant Ssalts is descr1tea
' and the analyses are interpreted., Results frcr
exarinations of metal and graphite survelllance specisen
from the core and of Specimen expcsed tc pump besl ga ses
are presented The fact that metallic fission products
appear in the cover gas prompted a study of tke chempistry
-and volatilizaticn behavior of the little~known '
intermediate valence fluorides cr molytdenum., Czxide
flucride equilibria in fuel systems was studied in a
research for separation processes. Phkase belbavicr,
~deccmposition pressure-and:corrosiveness of flucrolorate
coolants is descrited. Recovery cf prctactinium and
remcval of fissicn rroducts by reductive extraction is
discussed., Developmental ‘studies in analytical chegistry
directed primarily tc improvements in analyses of

Accession Numker ACD67C0ZC to ACD67CC26

 
 

 

Page 38

Category a
Molten-Salt Reactcr Prograts

ACD67C026 *Ccntinued*
radioactive samples of fuel for oxide and urariug
trifluoride and for impurities in helium offgas from the -
MSRE.
*analytical cheristry + *boron trifluoride + ‘coolants +
*corrosion + *fission products + *flucrcktorates +
*molten salts + *grarhite + *cxides + #*rare eartns +
*reductive extraction process + analysis + Etebavicr +
beryllium fluoride + bismuth o blanket + capsules + cells +
chemical reactions + chemistry + chrewium + ccmpatikbility +
cores + ccrrosicn prcducts ¢+ cover gas + decomposition +
distribution + equilibriur + equipment + experiment +
fissile materials + gas aralysis + gases + helium +
hot cells + hydrocarbons + impurities + inert gases +
inventories + irradiation ¢ liquidus ¢+ materials +
materials testing + mists + gpolyktdenum + MSRE +
off-gas systems ¢ cxide precipitation process +
phase equilibria ¢+ physical properties + prctactiniuz +
protactiniuxm fluorides 4 radiclysis + rare gases +
reduction + sampling + solubility + solidus +
spectrophctcretry + stability + surveillance + testxng +
thorium + thorium fluorides + uranium + uranius flucrides +
Yapor fgressure
OTHER CATEGCRIES: CXX + [XX + MCD

ACD67C027

Bohlmann EG

IRRADIATION EXPERIMENTS (FART 4, MSRF SEMIANMN. EBCGE. FEPT.

8/31/67) |

0Oak Ridge Naticnal Laboratory, Tenn. -

ORNL-4191 (Dec. 1967) pp. 176=-195, 10 fig, 7 ref.
A second thermal convection in-pile lcof ccntairing fission
fuel was terrinated when a crack developed in the core
outlet pipe. The crack was caused ty radiaticn
embrittlement of the Hastelloy K and stresses encountered
during a reactor settack, Sufficient ojeratirg time bhad,
however, heen achieved to proiuvce fissicn product
concentration levels equivalent to equilibriurz in a-
breeder; therefcre an exhaustive evaluation of the
experiment is presented.

*fission prcducts + *gamma radiation + #*graphite +

*Hastelloy N + *in-pile tests + *nclle retals +

*therzal ccnvection + actinides + analysis + :

anal ytical chemistry + beryllium flucride + circulaticn +

compatibility ¢+ ccres + ccrrcsicn + corrosion products +

decay + dismantling + emkrittlement ¢ examinaticns + '

experiment + fissicn + flucrides + aolten salts +

gamma spectrometry + hot cells + leaks + lithium flLCtlde +

materials testing + gclybdenum ¢ MSEE + progress Cefort ¢

radiation damage + rare earths + research + sampling +

stress + stress rupture + uranium fluorides

'+ ACD680015

Accessicn Number ACD67C02€ to ACC6ECC15
)

. Page 39

Category A
Molten-Salt -Reactor Erograsms

ACD680015 *Continued*

Grimes WR ;

CHEMISTRY {EART 3, MSRP SEMIANN. PROGR. REPT. 2/29,¢€E)

Oak Ridge Naticnal Laboratory, Tenn.

CRNL-4254 (Aug. 1¢68), pp. 88-173, 55 fig, 102 ref..
The chemistry of the MSRE is éiscussed from the standpoint
of fuel compositicn and fpurity, corrosion chewmistry, énd
isotopic composition of the uranium in the fuel. Fission
product behavior in the fuel and in the .cover gas 1is
described.  Results on fissicn prcducts found cr satgples
of graphite and metal from the core are given. Cther
topics are "Proton Reaction Analysis for Lithiuz and
Flucrine in ¥SR Grarhite¥ .and "Surface Fhenomena in
Molten Salts", Items pertaining to thke physical chemistry
of nolten salts are the thermodynamics of 1iF-EeF2 melts
from EMF measurements, and electrical properties cf melts,
The chemistry cf silica in 1iF-BeF2 melts is presented. A
Molten Salt Chemistry Informaticn Center is described.
Synthesis and properties of molytdenus and nicbiue
flucrides is discussed. - Reprocessing of fuel Lty reductive
extraction into molten tismuth is described, «ith srecial
emphasis cn prctactinium recovery. The behavior of BI3 and
fluoroborate mixtures is examined fror the starcdpcint of
rhase relaticas, ncn-ldeallty, thermodynam1cs, corrosion
and compatibility.

actinides + berylliurm + beryllium fluoride + bubbles +

capsules + cells + concentration + coolants + corrcsicr +

cover gas ¢+ decowmfosition + distribution + :

electrical properties + electrical conductivity +

electrolysis + equilibriur + examinations + experiment +

fissile materials + fuel preparation + camza spectrcmetry +

Hastellcy N + heat transfer + hot cells +

hydrofluorination + hydrogen + in-pile tests +

interfacial tersion + irradiation + liquidus' +

lithium fluoride + materials ¢+ mists + MSRE + |

off-qas systems + oxide prec151tat10n process + oxldes +

oxidation + phase eguilikria + progreéss repcrt +

protactiniur + fprotactinium fluorides * rare gases +

reaction rates + reduction + research + sampling +

sodiur fluoride + solublllty + - -specific heat ¢+ -

test facilities + testlng + thermal properties + thcz1um +

thorium flucrides + uranium + vuranium fluorides +

vapor pressure + viscosity + volat111ty +

zirconium fluoride + *bismuth +

*boron trifluoride + #*chemical propertxes +

_*chemxcal reacticns + *chepistry + *compat1b111ty +

*corrosion products + *fluorides + *flucrolkorates +
*molten salts + *graphite + *noble metals + . '
*physical properties + *process1ng + *rare earthc +
*reductive extracticn prceces

OCTHER CATEGCEKIES: CXX + fiCE

ACD68CO1E

Accessicn Number ACD6B0015 to ACD6EC(16

 
 

 

Page 40

Category A
Hclten-Salt Reactcr Progra:s

ACD68CO16 *Ccntinued«

Bohlmann EG

IRRADIATION EXPERIMENTS (PARt 4, MSRP SEMIANN. FRCG. FEFT.

2/29/68) -

Oak Ridge Naticnal Labora tcry, Tenn.

ORNL-42%4 (Aug. 1968) pp. 174-182, 2 fig, 8 ref.
The isotope balance on a second in-pile convect1cn locg
containing fissioning fuel is given. Fission product
behavior is descrited. In this lcop the fuel wetted the
graphite, presumably because of trace amounts of moisture
present in the helium used in loading, samplirg and
draining. Acccrdingly a study wvas made of the effect of
moisture on the wetting of graphite Ly MSRE carrier salt.
Alsc presented is a decign for a third in-pile molten salt
convection loop.

actinides + cells + compatxblllty + cover gas +

experiment + fissile materials + flucrides + fuels +

gas analysis + gases + inert gases + laboratory equipment +

materials ® noble metals + MSRE + sampling + *chemistry +

*examinaticns + #fissicn fredevcts + *molten salts +

*graphite + #*in-pile tests ¢+ *thermal ccnvection

ACD68C022

Grimes WEH '

CHEMISTRY {(PART 3, MSRP PROGR. REPFT. 8/31/68)

Cak Ridge Eatloaal Laboratory, Tenn.

ORNL-4344 (Aug. 1968) pp. 109-199, 57 Fiq, 13# fRef.
MSRF chemistry topics include Feasikility cf Fueling ilth
PuF3, Burnup, High Temfperature Fuel-Graphite Compatibility,
and Examination of a Corroded Sarrle Cagsule. Fissior
Product Behavior is discussed in connection with Specimens
from the Core, Analyses for Li and F, ard Surface Phei1cttena
in Molten Salts. Items under Fhysical Chemistry of Molten
Salts include Molybkdenum Ffluoride Chegpistry, Alkali
Flucrcborates, Physical Properties of ThF4-containing
Melts, Electrochemical Studies, Spectroccopy, Oxide
Chemistry, and the Chemistry of Silica in liF-EeF2, Huel
Reprocessing was studied in experiments cn the Reductive
Extracticn of Pa and of Rare Earths intc Biswmuth,
Analytical Studies included oxide determinaticn, U(3+) and
total reducing powver, U(5+¢) in liF-BeF2-2rF4, Ni {o) /Ni(+)
couple, Cr(2+), Hot Cell Spectrophotcreter, Stectra of
U(5¢) and U(6+), a Gas Chromatograph for tke Oft-gas
systenm, Hydrocarbons in MSEE Helium, Gamma Spectroscogy,
Fission Froduct Penetration in Graphite, U-235 Aralyses,
and Deterwvinaticn of U.

*anal ytical chemistry + *Lismuth + *cbemlcal ;ropettlec +

tchemrical reactions ¢+ *chemlstry + #*compatibkility +

#*corrcsicn products + *fissicn products + *xluorldes +

tfluoroborates + *molter salts + *graphite +

*physical properties + *prccessing +

Accession Number ACLE680016 tc¢ ACD680022
@K

Page 41

Cateqgory A
Molten-salt-Reactcr-Prograns

ACDGBCOZZ *Continued®

*reductive extraction process + *SOluhllltY +
*surveillance + berylliur + beryllium fluoride + bukbles ¢
burnup + capsules + carbides + cells + chrecgiuve + :
concentraticn + coolants + corrosion + cover.gas +

density + distribution + electrical properties +
electrical conductivity + electrolysis + equilibrium ¢+
examinations + expansion + experiment + fissile raterials +
fuel preparaticn + gamma spectrometry + Hastelloy N +

heat transfer + hot cells + hydrofluorinaticn + bydrcgen +
hydrocen ccrpounds + interfacial tension + liquidus +
lithium fluoride + materials + mists + MSRE +

‘noble metals + off-gas systems +

oxide precipitation process + oxides + cxidation +
phase eguilibria + plutonium fluorides + precipitation +

progress report + protactinium + prctactinium flucrides ¢

rare eartls + rare gases + reaction rates + reduction +
research + sampling + sodium fluoride + solidus ¢+

specific heat . + spectrophctcmetry + test facilities +
testing + thermal conductivity + therral prcpert1e= +
thorium + thorium fluorides + uranium + uranium fluorldes +
vapor pressure + V150051ty + volatllzty +

Zzircorium fluoride -

GTHER CATEGCRIES: CXX + [XX +¢ HCD

ACD6EBCO23
Bohlmann EG - ' o :
IRRADIATION EXPERIMENTS (PART 4, MSEF EROGB. FEET. 8/31/68)
Cak Ridge Kational Laboratory, Tenn. - -
ORNL-4344 (Aug, 1968), pr. 20C-210, 1 flg,-8 ref.,
Examinat ions of the grarhite from an CFF convection lcop
shoved the salt had wetted the qraphite, ccntrary tc
previcus exreriences in very dry inert gases. Subsequent
laboratory studies show that: extremely rinute :
concentraticns of water (approximately 1 ppm) promote
wetting at points of three phase contact of salt, graghite
and gas, A second Haetelloy~N capsule containing -
NaBF4-NaF (92-8 mole %) was’'irradiated for 1460 hr at
600 'deg C in three successive spent HFIR tuel elesents;
-.nc-deleterious effects vere observed. Fluorine due to the
delayed neutrons by B-10F3 + N to LiF + alpha + F2 was
deered to be tolerably low., The jumper section of the NSRE
off-gas lxne, 2 ft downstream frcr the fumpp bcewl, was
reccvered for examipation. All ‘ipternal snrfaces were
covered with a thin, soctlike film, and no other deposits
were found., A group of fissicn products, largely "noble
metals" (Mc, Ru, Ag, Te) were present in quantities
several hundred times the amcunts expected frce the
inventory of salt present in the deposit; this
substantiated earlier olservations that metal= cculd be
transferred - in the cff-gas,

-

Accession Number ACD680022 tc ACD680023

 
 

Fage 42

Category 2
Molten-Salt Reactor [rograms

ACD680023 *Continued«

*toror trifluoride + *fisscion fproducts + *fluorobo:ates +
*in-pile tests + *noble metals + *off-~gas systems +
*wetting + behavior + cagpsules ¢+ chemical reactions +
chemistry + compatibility + coolants ¢ corrcsicn +

cover gas + delayed neutrcns + molten salts + ‘

gamma radiation ¢ gamma sources + gases + hydrocarbcns +
inert gases + inventcries + materials + materials testing +
mists + MSEE ¢ radiation damage + radiclysis ¢+ rare gases +
sodiur fluoride ¢ testing ¢ uranium + uranium fluorides -
CTHER CATEGCFIES: HCD - -

ACD69CO24

Grimes WR

CHEMISTRY (PART 3. MSRP PROGE. REFT. 2/28/69)

Cak Ridge Bational Laboratory, Tenn, :

ORNL- 4396 (Feb. 1969), pr. 129-196, 32 fig, 122 ref. ;
MSBE chemistry topics include the uranium raterial
balance, ccrrosicn, adjustment of U (3+)/Sigma U, and .
foaming behavior. Fission product dispcsitict in the MSRE
is described under Examination of Graphite frcm the Core,
Distribution of Fission Products, Fissicn Eroduct
Inventory, Off-gas Pnalyses, and Material Reccvered frco
the cff-gas line., Alsc the formation of aerosols fror the
MSRE was studied extensively in a hot cell; additiomally
tracer level studies were also made. The chenistry of the
fluorides of Nk, Mo and Ru was studied ty mass spectrcscopy.
Under Physical Chepistry of Kolten Salts are 15 items
dealing with such topics as CeF3 {a standin fcr EFuF3)
solubility, Zone Melting, Phase Relaticns, Sclubility cf
Th(c) in LiPF-ThF4, Densities, Crystal Structure,
Spectroscopy in a Diamond-Windowed Cell, Distributicn of
U(4+) between fuel and (U-Th)C2 Solid Soclution, Feference
Electrodes, Concentration Cells, Electrical Ccrductance.
Chemistry in sugpcrt cf fuel reprocessing deals with
Teductive extraction of Zr, U, Pa, rare earths, and Th.
Analytical Chenristry is represented by Ietermination of
Oxide and Cxidation State, Eenf, Vecltammetric, ard
Spectrographic studies, Ketting Eehavior, Contasinants in
Blanket Gas from NaEF4 tests,'and the Determitaticn of
Bi in MSRP Saltse

*anal ytical chemistry * stismuth + *chewlcal ;ro;ertxec +

*chemical reactions + *chemistry ¢+ #*corrosion products +

*examinations ¢ *fission products ¢ *fluorides + -

*fluoroborates + *molten salts + *graphite + *llguldus +

*naterials + tnoble metals + *phase equilikria +

*processing + *reducticn ¢+ *reductive extraction preccess +

*s0lidus + #*solubility + #*surveillance + .

actinides + beryllium flucride +

beryllium oxide + blanket + loron tr1f1uor1de + bubhle= +

carbon + cagsules + cells + ccmpatibility + concentration +

Accession Number ACD68C023 to ACD69002Z4
"%

Page 43

Category A
Molten-Salt Reacter Prograss

ACD69€02H *Ccntinuedx

‘coolants + cover gas + dens1ty + dvposztlon + diagrams +
- distribution + electrical conductivity + electrolysis +

equilibrium + experiment + foaming + freezing +

fuel preparaticn + fuels + gamma spectrometry +

gas analysis + gases + hydrofluorinaticn + hydrogen +
hydrogen cozpounds + inert - gases + in-pile tests +
interfacial tension + ions + latoratcry equipnent +
lithiurp flucride + melting + metal transfer process +
mists + MSEE + off-gas systems +

oxide precipitation process + oxides + OX1dat10n +
physical properties + plutonium fluorides +

potassium fluorides + precipitatior + protactinium +
protactinium fluorides + potassium flucrides +
precipitaticn + protactirium fluorides +

rare earths + rare gases + reaction rates + sampling +
secondary salts + sodium flucride + spectrophotometry +
surface tension + thorium + thoriug fluorides +
uranium + uranium fluorides + volatility ¢ zirconium +
zirconium fluoride

OTHER CATEGORIES: CXX ¢+ DXX + MCD

ACD690025

Bohlmann EG '

IRRADIATICN EXFERIMENTS (PART 4, MSRP SEMIANN PROG REE1

2/28/69) '

Cak Ridge National Laboratory, Tenn.

CRNL-4396 (Aug. 1969), ¢ z10. '
The rrograk of in-file rolten-salt loops was suspended.
Laboratory experlments showed that the wetting cf gra jphite
that was seen in the ORF molten-salt loop was protably
caused by traces of moisture in the gas used tc transfer
salt,

experiment + graphite + inert gases 4+ in-pile tests +

molten salts + prcqgress report + wetting

ACD690031
Grimes WR : S C '
CHEMISTERY (FART 3, MSRP PROGR. REPT. 8/ 1/6“)
Oak Ridge Hat:cnal Labgcra tory, Tenn. -
OBNL-QQM9_(Aug. 1969), pp. 9€6-163, 29 fiq, 1CE ref.
| MSRE Cheristry topics 1nclude the .composition of the fuel,
plutonium material lalance, and gas Febavicr. Fissior
- product behavicr is deduced from surveillance sreciwers from
~latcratory studies cf metal fission product chemlstry. A
measurement of the surface tensicn of tte fuel ir the
~reactor is presented. Chemical and physical prcperties of
alkali fluoroborates are-given under 10 toric headlng
- Topics relating to the Fhysical Chemistry of Kolten altq
include phase relatlons, hetrcgeneous equilibria,
liquidus temperature, sclubility of thorium, U(3+)/U(4+)

Accession Numter ACL690024 tc ACD690031

 
 

Page 44

Category A
Molten-Salt Reactor Erograms

ACD690031 *Cont1nued*
ratic, spectrur of UF3, concentration cells, electrlcal
conductance, viscosity, and density. Items c¢f irpterest in
connecticn with fuel regrocessing are the reductive
extraction of Pa, rare carths and thoriur, and also the
separaticn of zirccorium as a platinide. In connection with
analytical chemistry, there is oxide deterrinaticn and
removal, U{3+),/U(4+) determination, electroanalytical
studies, spectral studies, hot cell spectrcphctceetry,
and bismuth determinaticn.

*anal ytical chemistry + *chemical prcperties +

*chemical reactions + *chemistry ¢ #fcoclants + |

¥fission prcducts + *flucrides + *fluoroborates +

*molten salts + *graphite + *physxcal fEoperties +

*surveillance ¢ berylliun +

beryllium fluoride ¢ bismuth + toron trifluoride +

tubbles + cells + concentraticn + corrosion products +

cover gas + dens1ty + distritution +

electrical conductivity + egu111br1um + examinations +

fuel preparation + gamma spectrometry + helium +

hot cells + hydroflucrination + hydrogen + impurities +

inert gases ¢+ interfacial tension ¢ kinetic eguaticrs ¢

liquids + lithiur flucride + materials + melting +

metal transfer process + noble metals +

oxide precipitation process + oxides + phase equilitria +

plutonium fluorides + potassium fluorides + frecipitaticn +

protactinijur + protactinium fluorides + rare earths +

rare gases + reaction rates ¢ reduction +

reductive extracticn prccess + research + sampl1ng +

sodium fluoride + solidus + solukility +

spectrophctcmetry + surface tension + testing +

thorium fluorides + uranium fluorides ¢+ viscosity +

-void fracticns + zirccrniur

CTHER CATEGCRIES: CXX + I[XX ¢+ MCD

ACD70C024

Grimes WH

CHEMISTRY (PART 3, MSRP PROGR. REFT. 2/28/70)

Cak Eidge Kational Lakoratory, Tenn,

ORNL-U4548 (Feb. 1970), pE. 93-187, 50 fig, 153 ref. -
MSRE chenmistry topics discussed are ccrrosicn, afpearance
of NE-95 in the fuel salt, isotopic composition of U and
Pu, and surface tehsion and wetting behavior, Fission
product behavior was demonstrated by sanples frcm the ccre
and from the purp bewl. 1laboratory studies of the metals
that are fission products are presented. Fcuiteen tofics
are discussed under Physical Chemistry of Molten Salts and
six under Properties of the Alkali Flucrcbcrates., These
include the oxide chemistry of Pu in molten fluorides, and
the solukility of the corrosicn product, Na3CrF€, in
flucrcborate melts., Basic chemistry work in support of

Accession Numker ACD69CC31 to ACD7CQ024
L4l

Page 45

Categorf A
Molten-Salt Reactcr Prograts

ACD70C024 *Continued* | '
fuel reprocessing included distrituticn of Ce, Eu, and Sr
between bismuth and LiCl, and 7 other topics related to
‘reductive extraction. ‘The following analytical chemistry
topics are presented: Tetermination of Oxide in MSRE Salt,
Deterrinaticn of U(3+)/=1qma U{4+) Ratios, Spectral
Studies, Tritium in the Fffluent Gases cf the MSRE,
Reference Rlectrodes in Mclten Fluorides, Femoval of Cxide
from NaBF4, Volatile AlCl3 Complexes., -
*chemical froperties + *chemical reactions + *chemlstry +
concentration + *coolants + *corrosicn +
*fission rrcducts + *flucrides + *fluoroborates +
*molten salts + *graphite + *materials + *oxides +
*physical fproperties + *fprogress Leport + *ra:e earths +
*surveillance + =2tritium +
analytical chemistry + barium + bismuth + compatibility +
cesiulb + corrosion products ¢+ cover gas + distributicr +
electrical conductivity + equilibrium + fuels + hydrogen +
ions + lithium chloride + lithium fluoride + reasurerent +
liguidus + metal transfer frocess + ESEE + nokle metals +
oxide precipitation process + phase equiliktria +
plutoniun + plutcniurm flucrides + reaction rates +
reductive extraction process + rescarch +
saspling + seccndary salts + sodium fluoride 4+ solidus +
solubility + specific heat ¢+ spectrcphctometry +
surface tension + technclecqgy + testing + thorium fluorides +
uranium fluorides + uranium-232 e uranium-233 + urarius-235
OTHER CATEGORIES: CXX + DXX ¢+ FCD

ACD70C038B
Grimes WR
CHEMISTRY (PABT 3, MSRP QEHIA&N PRCG REET 8/31/70)
Cak Ridge National Lahoratory, Tenn, -
ORNL-4622 (Aug. 1970), pE 60~-118, 35 fig, 143 ref.
~ Fission product hehavxor in the MSRE is analyzed in terms of
age cf tke rroeducts, time of exposure for short expogures,
surface roughness, flou conditions, and the comparison of
“deposition on graphite with that cn retal. A pcsslble
meckanisr for “"smokes" of metallic fission products is
advanced. The chemlstry of molyidenur and nicbiur
flucrides is treated.  Various properties of alkali
fluoroborates, including tritium retenticn were measured.
‘Phase relaticns, Pu solubllLty, oxide chemistry, :
~ entropies and conductances were investicated. Fissicr
- product separaticon studies were expanded to include
chemistry of molten chlorides, 'Analytical methcds beiny
studied include electrochemistry, studles of NaFF4,
coolant salt, and in-line analyses, |
*analys:s + *analytical chemistry + *blsmuth +
*boron trifluoride + *chemical properties +
*chemical reacticons + *chemistry + %coolants + *deposition ¢

Accession Number ACD700024 tc ACD700038

 

 
 

 

Page 46

Category A
Molten-Salt EReactor Eroqrams

ACD700038 *Continued* ,
*electrical conductivity + *fission products + *fluorides +
sfluoroborates + *graphite + molten salts + *rcble mretals +.
*oxides + %*processing + *physical properties + *rare earths +
sreductive extraction process + *solulkility + :
*surveillance + *thercodynamics + actinides +

beryllium + beryllium fluoride + capsules + cells +
compatibility + concentraticn + corrosion + diagrams +
distribution + electrical properties + entrairment +
equilibrium + examirations + experiment + fissile materials +
fuel preparation + gamma spectrometry + gases + Hastellcy N +
hydrogen corpcunds + inext gases ¢+ inventories ¢+

irradiation + liquidus + lithium fluoride + materials +
measuresent + rists ¢+ molybdenum + FKSRE + off-gas systems +
oxide precipitation process + oxidaticn + phase equ:llbrla +
primary salt + prcgress regort + protactinium +

protactinium fluorides + rare cearths ¢ reacticn rates +
reduction + sarpling + scdium fluoride ¢+ spectrophotometry 4
test facilities + testing + thermal prcperties + thcrive +
thorivm flucrides + vuranium + vuranium fluorides +

vapor pressure + volatility ¢+ zirconiur fluoride

ACE6SC008

{Staff Report) |

METALLURGY (Chap. 4 MSRP Prog. Rept. 2/28/65)

Cak BRidge National Laboratory, Tenn.

OKNL- 3812 (June 1965), ff. 63-86, 13 fig, 12 tef.
Reaction of Hastellcy N thh impure aitrogen apparently
gave a protective oxide film. Plugs were velded to seal
the MSRE heat exchanger after four tubes had Leen removed.
Welds in Hastelloy N showed reascnakly cccd creep and
tensile fproperties. In a fpoorly weldable heat silicon and
aluminum concentrated in a grain-toundary eutectic, EBrazing
allcys were sought for joints betseen various refractcry
metals and graphite or Bastelloy N. Dersity, gas evolution,
and porosity vere measured on ESFE graphite. CGCxidation
accelerated by thermal cycling led to a crack ir a
Hastelloy N baycnet tube in a development test., Plans for
Hastelloy N creep measurements in the ORR and the MSRE
surveillance riqg are described.

drain tanks ¢+ scaling ¢ cracks + tungsten + tantalun +

alloys + ccmpatibility + corrcsion + creep + development +

ductility ¢+ embrittlement + eguipment + exasicaticns +

failures + ricrcstructure + brazing ¢+ graphite +

Hastelloy & ¢ heat treatments + impurities + in-fpile tests +

inspection + irradiation + jcints + materials + :

materials testing + mechanical properties + metallccta;hy +

metallurgy + metals + mclybdenum + nitrogen +

progress report + radiation damage ¢ rugture +

stress rupture + surveillance + testing + welding

CTHER CATEGCRIES: EAX + FEX + CGEX

ACE6E5C014

Accessicn Number ACD700038 to ACE650014
L

Page U7

Category A
Mclten-Salt Reactor Erograns

ACE650014 *Continued*

(staff Refpcrt) | |
METALIUEKGY (Chap. 4 MSRP Prog Rept. 8/31/65)
Oak Ridge National Labocratory, Tenn.

ORfRNL- 3872 (Dec. 1965), pr 81-105, 20 flq. 11 ref.

Molten fluorides were czrculated in loors cf Hastelloy N
and type 304 stainless steel, and molten lead in several
' steels and Nb-1% Zr.  The MSRE surveillance rig is
described. Irradiaticn creep of graphite is discussed for
advanced reactors, Creep of Hastelloy N ics repcrted as
affected by thermal and mechanical treatments, prior.
irradiation, and simultaneous irradiaticn., Wwelding air-
melted but not vacuum-melted Hastelloy » impaired strength
and ductility. Autoradiography and micrcoprcbe aralysis
showed intermetallic precipitation in Hastelloy N.
precipitation + modified Eastelloy N + alloys +
compatibility + ccrrcsion + creep + development +
ductility + emlkrittlement + examinations + failures +
fluorides + molten salts + micrcstructure + graphite +
Hastelloy N + heat treatments + impurities + in-pilec tests +
inspection + iron alloys + irradiation + joints ¢+ lead +

liquid metals + loop + mass transfer + materials +

materials testing + mecharnical froperties ¢+ metallography +
metallurgy ¢+ metals + progress report + radiation darage +
rupture + stainless steels + stress rupture + surveillance +
testing + thermal convection + welding -
OTHER CATFGORILS' ECX + FBX + GXX

ACE660012

(Staff Repcrt) ‘ S
METALIUBGY (Chap. 4 MSRP Prog. Rept. 2,/28/€6)
Oak Ridge Naticnal Labcratory, Tenn.. a
CRNL-3936 (June 1966), pp. 95-121, 18 tiq, 10 ref.
Molten lead was circulated in NB-1% Zr and in a chrcmium-
~molybdenum steel; mclten fluorides in Hastelloy N and type
304 stainless steel, MSRE surveillance specinens were
exarined after prenuclear operation. Irradiation darkened
grain boundarles of Hastelloy N tut did not irduce reaction
with nitrogen. Development of graphite-to-Hastelloy K
- brazed joints included préssure testing with fclten
flucrides Prccurement, ‘characterizaticn, and ¥SER
r@qulrements of. graphlte are descrited., Irradiation
decreased rupture life -and creep ductility of Hastelloy N.
Effects of pre- and postirradiaticn heat treatrerts or
tensile propertles cf Hastelloy X are shown. Attempts were
wmade to 1mprove adverse mechan:cal p:opertlec cf Hastelloy N
welds.

,alloys + compatibility + corrosion- + czeep + density +
‘development + ductility + electrical conductivity ¢

embrittlement + equipment + examinations + exfpansicr +
failures ¢+ fluorides + mclten salts + microstructure +

Accession Number ACE6S5C014 tc ACE660012

 
 

Page 438

Category A
Molten-Salt Reactor Frograss

ACE660012 *Continued=* _ ‘ _

brazing ® graphite + Hastellcy N + heat treatments +
impurities ¢ in-pile tests ¢ inspection + ircr allcys +
irradiaticn + joints + lead + liquid metals + loop +

mass transfer + materials + materials testing +

mechanical properties + metallography + metallurgy + metals +
rnolybdenum + nitrogen + physical profperties ¢ frccuresent +
progress report + radiaticn damage + rupture + :
specifications + stainless steels + stress rupture +
surveillarce + testing + thermal convection + welding +
tensile properties

OTHER CATEGORIES: EXX + FYXX + GXX

ACE660018

Adamson GHM

BSBRP MATEFIALS (Chap. 6 MSRP Prog. Rept 8/31/¢€€)

Oak Ridge Naticpal Laboratory, Tenn.

CRNL-4037 (Jan. [ 967), pp. 97-133, 33 fiq. 20 ref.
Mechanical damage from unegqual expansicr ci ¢raphite and
Hastelloy N required revision and replacement of the MSRE
surveillance rig. Aluminum alloy from a blcwer failure was
removed frow the MSRE radiator; simulated tests showeé no
damage of aluminum to Hastellcy N. Graphite studies
included characterizaticn of several grades, measurement of
creep under irradiation at 700 and 10C0 degq C, trazing to
polybdenum and Hastelloy N, and molten-salt ccrrosion of
brazed joints, Observations c¢n Hastellcy N irclude gccd
veldability with titanium modifications hut poor with
zirconium, postirradiation ductility imfroved by titatrium
additions but not tungsten or niobium, and extecsion cf
creep studies to 987 deg C., MNMclten flucrides were
circulated in lcocps of Hastelloy K, type 304 stainless
steel, and NE-1¥% Zr clad with type 44€ stainless steel, and
Croloy 9M. A Crolcy 9M loop circulating lead plugged aud
was examined.

heat exchangers + lead ¢+ zirconium ¢ tungsten + iron alloys +

aluminum + alloys + brazing + corrcsicn prctecticn + :

columns ® ccmpatibility + corrosion + creep + density +

development + distillation e electrical conductivity +

embrittlergent + duvctility + equipment + examinations +

expansion + failures + fluorides + molten salts +

tluorcborates + graphite + Hastelloy N + heat treatments +

in-pile tests + inspection ¢ irradiaticn ¢ jcints + lccfp +

mass transfer + paterials + materials testing . e '

measurement ¢+ mechanical properties ¢+ wmetallccraphy + .

metallurgy ¢+ metals + modified Hastelloy N + molytdenum +

MSRE + physical properties + procurement + [frcgress repcrt +

radiation dasage + reliability + rupture + stainless steels +

stress rupture + surveillance + testing + oo

alloy compcsition + welding + liquid metals

CTHER CATEGCRIES: EXX ¢+ FXX + GXX

ACE67C021

Accession Numbker ACE660012 to ACE670021
-

Page 49

- Category A
" Molten-Salt Reacter Prograts

ACE67C021 *Ccntinued#
Cook WH + McCoy HE + Kennedy CR + Werner WJ + Litmar AP +
Canonicc DA + Haseltine DN
MOLTEN-SALIT REACTOCR PROGBAH HATERIALS lChap. € MSRP Prcqg.
Rept. 2,28/617)
Cak Ridge National Laboratory, Tenn.
ORNL-4119 (July 1967), FPEe 95-117, <C fig. 11 ref.
The first Hastelloy N surveillance specimens were removed
from the MSRE, and tensile properties, gparticularly
ductility, are shown as functions of temperature and strain
rate, Hastelloy N was unaffected by ccntact with aluminum.
MSBR graphite studies included regquirements, studies of
prospective grades, irradiation plans, and brazing to
Hastelloy N. Loops are descrilked to study ccrrcsicn cf
Hastelloy N, stainless-steel-clad niobium-1% zirconiurm, and
type 3041 stainless steel in mclten salts.
alumipum + brazing + compatibility & corrosion +

emprittlement + ductility ¢ examinations + fluorides +
mclten salts ¢ flucrcbcrates + graphite + Hastelloy N +

heat treatments + in-pile tests ¢ irradiaticm + lccy +
mecharical rroperties + metallography + MSEE +

progress report + stainless steels + surveillance + testirng
OTHER CATEGORIES: EDX + FBX + GAX

ACE670028
McCoy HE + Weir JR
MATERIALS LEVELOPMENT (Part 5 MSRP Prog. Rept. 8,31,€7)
Oak Ridge Naticnal Labcratory, Tenn.
CRNL-4191 (Cec. 1967), pp 196-238, 37 fig, 24 ref.
Creer and microstructural results are given fcr Hastelloy N
MSRE surveillance stecimens; microstructures include weld
metal and modified alloys. Other Hastelloy N studies
include aging mcdified alloy, weldability of zirconium-
modified alloys, precipitate rorprhclogy, residual welding
stress, corrosion by molten salts, tellurium comppatibility,
“titanium diffusion, and compat1t111ty with qrarhite,
Graphite studies include procurement, physical and
- mechanical characterlzatlon.'tatr1cat1cn, ccatirg with
mclybdenum, gas irfpregnation, and start of HFI1F
irradiation. Molten salt corrcsicn tc type 3(C4 L
_stainless steel, Crclcy 9, and a graphite-to-molybdenum
brazed joint were also studled. |
diffusion + iron allcys + zirconium + elasticity +
alloys + brazing + carbides + ccatings + compatibility +
corrosion + creep + density + deposition + develcpment +
electrical conductivity + embxlttlement + ductility +
equipment + examinations + falrication + failures +
fluorides + molten salts + fluoroborates + graphite +
Hastelloy X + heat treatments + impregnation +
in-pile tests ¢ inspecticn ¢ irradiation + joints + lcop +
mass transfer ¢+ materials + materials testing +

Accessicn Number ACE67002z1 to ACE67CCZE

 
 

Page 50

Category A
Mclten-Salt Reactcer Prograns.

ACE670028 *Ccntinuedx

measurement + mechanical properties + metallocraphy +
petals + rcdified Hastellcy ¥ + amolybdenum ¢ FSRE ¢+
physical properties + precipitation ¢+ frocurerent +
progress regort + radiaticn damage + reliability +
rupture + sealing + stainless steels + stress rurture +
surveillance + testing + allcy composition ¢ weldxng +
X-rays -
OTHER CATEGORIES‘r EXX ¢+ FXX + GXX

ACE680017

McCoY HE + Weir JR

MATERYALS LCEVELOPMENT (Part 5 HSRP Prog. Rept. 2/29/68)

Oak Ridge Naticnal Labcratcry, Tenn.

CENL- 4254 (Aug. 1968), Pp. 183-240, 43 fig, 33 ref.
Hastelloy N lnvestlgat10n= include creep of M SRF
surveillance sgecimens, effect of strain rate on ductility
of irradiated and control specipens, creep ¢t rcdified
allcy, electron wmicrcsccpy of effects of silicon and
titaniuom on precipitate morphology, titaniur diffusior,
veld stresses, weld development, oxldatlon, and corrosion
of standard and modified alloys Ly varicus mclten salt
Grarhite studies include p:ocurement. density, , :
microstructure, increase in porosity by cxidaticr, x-1ay
dif fraction, gas impregnation, sealing with mclybdenur,
ultrasonic measurement cf elastic properties, and brazing
to molybdenum and Hastelloy N. Also studied wele
irradiation of brazing alloys, ultrasonic inspection of
brazed joints, and electrical resistivity cf scdified
Hastelloy N. :

alloys + brazing + carbides + ccatings ¢4 corpatibility +

corrosicn + creep + density + deposition + development +

electrical conductivity ¢ emktrittlerent + ductility +

equiprent + examinations + fabrication ¢ failures +

fluorides + molten salts + fluorolorates + ¢rarhite +

Hastellcy N + heat treatments ¢ impregnation +

in-pile tests + inspection + irradiaticn + jecints f lccg +

mass transfer + materials ¢ materials testing + measurementl+

mechanical properties + metallography ¢ wetallurgy + getals +

mcdified Hastelloy N + mclybdenum + ¥SEE + oxidation +

physical properties + precipitation + frocurement +

progyress regport + radiaticn damage + reliability +

remote welding + rupture + sealing ¢+ stainless steels +

stress rugture + surveillance + testing +

alloy composition + welding ¢ x-rays + tensile prorerties

OTHER CATEGORIES: LEXX + FXX ¢ GXX

ACE680024

HcCoy HE + Weir JR

MATERIALS LEVELCEMENT (Part 5 MSRP Prog. Re;t. 8/31/6&)
Oak Ridge Naticnal Laboratory, Tenn.

Accession Numbter ACE670028 to ACE680024
-

Page 51

Categoriil |
Mclten-Salt Reactor Erograms

ACE680024 *Continued*

ORNL-4344 (Feb. 1969), PE. 211-290, 6z fig, 32z ref.
Graphite and Hastelloy N,survelllance Sgfecilers uere
remcved from the MSRE and replaced; creep tests were run on
removed Hastelloy N and two modified allcys., Etffcrts on
several grades c¢f graphite include procurement, density,
resistivity, permealkility, Ltend testing, x-ray diffraction,

- gas impregnaticn, =ea11nq with molybdenum, HFIF irradiation,

and small-angle x-ray scattering., Effects cf titaniuzn
content and irradiation temperature on creep, effect of
aging on tensile properties, weldability, and mclten salt
corrcsion were studied c¢n modified Hastelloy N, Stancard
Hastelloy N studies included resistance to salts and air,
transiticn jcints with graphite, mechanical properties of
velds in irradiated specimens, and fluted tubirg.
Ptecipitate morpholcgies were studied in both., Cther
corrosion studies included Haynes allcy No. 2% in
flucrchorates, stainless steel, and a chemical separatxon
still. Eearing coatings were stucdied by x-ray dlftractlon.

alloys + tearings ¢ brazing + carbides + cermets +

coatings + columns + compatitility + corrosicr + creegp +

density + defposition + develcrment + distillation +

electrical conductivity + emkrittlement + ductility +

cokalt + equipment + exacinations ¢ expansion +

fabrication + failures + fluorides ¢+ mclten salts +

fluorcborates + graphite + hafnium + Hastelloy N +

heat treatments + impregnation + in-pile tests ¢

inspection + irradiaticn + 1c13ts + loop + mass transfer +

materials + materials testing + measurecent + '

mecharical rrofrerties + metallography + metallurgy + metals +

modified Hastelloy N + molybtdenum + MSRE + ‘cxidaticr +

physical fproperties + precipitation + procurement +

progress report + radiation damaqe + reliatility +

remote welding + rupture + sealing + stainless steels +

stress rupture + surveillance + testing + |

alloy comfosition + welding + x-rays ¢ tensxle propertles

OTHEE CATEGCRIFS' EXX + PXX + GXX:'

ACE69C026

 Eatherly WE ¢ EcCoy EE + Weir JR

MATERIALS DEVELOPMENT (PART 5, MSRP SEEIANL EFCG EEET
2/28769)
Oak Ridge National Laboratory, Tenn.-

OBNL-4396 (Aug. 1969), pp. 211-268, 54 flg, 44 ref.

Magnetic particles from the MSRE pump tcwl were cuite
- diverse as shovn by microgrobe analysis and metallography.
 Graphite topics include radiation darage fundarentals
- binder chemistry, hct Lrecsxng, physical characterlzatlon,
thermal conductivity, x-ray diffractien, electrcr
micrcscopy, gas impregnation, and irradiation in HFIR,
Creep of modified Hastelloy N was studied as affected by

Accessicn Number ACE6B80024 to ACE6900Z¢

 
 

Page 52

Category 2
Mclten-Salt Reactor Erograms

ACE690026 *Continued#
titanium and carbon contents, aglng, and irradiation at
various temperatures, Flectron mzcrcsccpy ideptified the

cartide frecipitates formed in Hastelloy N and various

modifications., Corrosion in varicus molten salts and in
~air and welding stresses vwere measured for Hastelloy X\,
Molten fluoride corrosion was ccmpared for Hastelloy K,
the modified alloy, and stainless steel., Cther studies
include brazing Hastelloy N tc graphite, brazing
molybdenum, corrosicn of TZ¥ by fluorides, and thermal
cycling of bearing materials.

alloy compositicn + bearings + bismuth + compat1b111ty +

coolants + corrosion + creep + defects ¢+ develcpment +

ductility + embrittlement + examinations + expansion +

fabrication + failures + ferrocalloys + fluorides +

fluorcborates + forwing + microstructure +

graphite + heat treatments + '

impregnaticn ¢ impurities + Hastelloy K + in- plle tests +

iron alloys + irradiation ¢ mass transfer + caterials +

mecharical properties + metallography + metallurgy +

metals + modified Hastelloy N + molten salts + MSRE +

molybdenur + physical prcperties + processing +

procurement + progress report ¢ radiation damage + Iupture +

sealirg + stairless steels + stress + surveillance +

test facilities + testing + welding ¢ lrazing + x~-rays

ACE69(C0132

Eatherly WE ¢+ NMcCoy HE ¢+ Weir JR

MATERIAL DEVELOPMENT (PART 5, MSRP SEMIANN ERCG BEET E/31/6C)

Cak Ridge Kational Laktoratory, Tenn.

ORNL-4449 (Feb. 1970), pp 165-213, &€ fig, 22 ref.
Preliminary examrinaticn cf graphite and Hastelloy N
surveillance specimens from the MSRE and tensile results on
the Hastellcy N are repcrted. Graphite studies include
x-ray diffraction determinaticn cf anisctrcpy, e€electrcen
micrcscopy, gas impregnation, irradiation erffects to Ligh
fluences, and interpretation c¢f radiaticn darage. For
titanium-mcdified Hastelloy N, creep was studied as
affected by titanium content and aging, and tensile
profrerties wvere related statistically tc several variatbles.
Combinations among Ti, Ef, NiL, and Y additicnes inprcved
- postirradiation ductility; titanium plus hafniur was lest,
Commercial heats are often infericr to labcratcry-nelted
Hastelloy N modifications. Electron microscopy tracec the
improvements to formation of MC-type cartides. GCther
studies include corxpatibility of Hastelloy K with coolant
salts and other fluids, depositinc tungsten, krazing
molybdenum, and remcte selding.

bismuth + coatings + compatitility + ccolants + corrcsicn +

creep + defects + density + deposition + development +

ductility + emkrittlement + equipment + exarminaticors +

Accessicn Number ACE6900z€ to ACE6GCC32
-

Page 53

Category A
Mclten-Salt Reactor Irograms'

ACER90032 *Continued* , |

expansicn + fabrication + failures + fluorides +

fluoroborates + forming + microstructure +

graphite + bhafniur + heat treatments +

impregnation + 1mpur1tles + Hastelloy N + 1n—;1le tests +

irradiaticn + mass transfer + materials +

mechanical properties + metalloqraphy + metallurgy ¥ tetals +

modified Hastelloy N + mclten salts + FKSRE + molybdenum +

physical properties + procurement 4 prccress Iepcrt +

radiaticn damage + reliability ¢+ remote maintenance + =

research + rupture + sealing + steam generatcrs + stress +

surveillance + test fac111tle= + testlng + tangsten +

brazing + x-rays

S~

ACE70C025

Weir JR + McCoy HE ‘ ' '

MATERIALS DEVELOPMENT (PABI S5, MSRE SEMIANN EECG FEET

2/28/10) |

Oak Ridge Naticnal Labcratory, Tenn.

ORNL 4548 (aug. 197C), pp 188-276, 83 flg, 45 ref.
Microstructural changes and creep properties were studied
on Hastellcy N surveillance specimens trom the MSEE.
Graphlte studies include electron damacge; dercsity,
resistivity, and permeability of promising grades; hot
and isostatic pressing; therral ccnductivity neasurement;
x-ray diffraction; electron microscopy; gyas impregnation;
irradiation; and lifetime calculaticn, Effects ¢f aging,
irradiation, and ccmpesition on mechanical properties,
electron miscroscopy, weldatility, and ccrrcsicr by
varicus rolten salts were studied for modiried Eastelloy N.
Back extrusion, weldlng,'and trazing cf molybderur,
compatibility cf alloys with bismuth, coating with tungsten,
and oxidation of steels were studied in sug;crt cf fuel
reprocessing. Also reported are progress in remote welding,
failure analysis of loop compcnents,'and ccm;dtztlllty
testing cf bearings _

:alloy composition + hearlngs + t15¢uth + coatings +
‘compatitility + coolants + corrosion ¢+ creep + deiects ¢

dep051tlon + development + emlrittlement + examinaticrs +
expansicn + fabrication + failures + ferroalloys + -
fluorides + fluorotorates + forming + n:crccttucture +
graphite ¢+ hafrium + heat treatments +

impregnation + impurities + Haatellcy N + ip-fpile tests +

“iron alloys + irradiatiorn ¢+ mass transfer + materials +

mechanical properties + metallography + metallurqy +
metals ¢+ gcdified Hastellcy N + molten salts + NMSEE +
molybdenum + physxcal properties + ‘procurement +
progress report '+ pyrccarbon + radiation damage + .
remote maintenance + rupture + sealing + stairless steels +
steam generatcrs + stress + surveillance + test tacilities +
testing + tungsten + welding + trazing |

ACE70C039

Accession Number ACE69C032 tc ACE700039

 
 

 

Page 54

Category A
Molten-Salt Reactor Prograrms

ACE70C039 *Ccntinued*

Weir JR .

MATERTIALS DEVELOPMEN1 (PART 4 HSBP PBCGE. FEET., 8/31,70)

Cak Ridge Kational Laboratory, Tenn. '

ORNL-4622 (Jan. 1971) pp. 119-198, 7% fiqg, £1 ref.
Tubing and therscccurle wells from the PSFE ccolant circuit
showed very little corrosion. Graphite develctrmernt
included procurement of new grades, determination of
density, resistivity, anisotropy, and wmicrcstiructure,
characterizaticn of pitch, building thermal ccnductivity
apparatus, pore sealing, and irradiation effects cn
density, fporosity, and rore seals. Investigaticns of
Hastelloy N modifications include effects cf titapium
content and aging on hardness, tensile properties, creep,
and postirradiation creep, postirradiaticn creejy of allcys
containing varicus ccmbinations of Ti, Nb, and FE,
weldability, creep of ccmmercially melted allcys, and -
microstructure., Lccp studies of corrosion iancluded type
3041 stainless steel, Hastelloy N and several ncdified
allcys exposed to fuel, blanket, and coclant salts. The
electron microprobe vas used ta study tlte ccrricesicn of
Hastelloy N by power plant steam, and study was started
on a duplex steam-generator tukte made cf Incclcy 80C and
Nickel 28C. Develcpment of processing equipment included
back-extrusion, welding, and trazing cf rolybderur
compcnents, corpatibility of Mo, TZM, MNb, MNE-1% 2r, Ta,
T-111, graphite, FE-5% Mo, and krazed jcints in Mc with
molten biseuth, and depcsition. of coatings by reducticn
of MoF6 and WF6 vapcrs and MoF6 dissolved in molten
fluorides.

krazing + creep + compatibility + contactors + density 4

deposition + ductility + electrical conductivity +

examiraticns + fabricaticn ¢ flvorides + fluoroborates +

bismuth 4 graphite ¢ hardness + Hastellcy N +

heat exchangers + heat treatments + impregnation ¢ iron +

irradiation + loop + molyldenux + modified Hastellcy N +

nickel + fprocurement ¢+ prcgress report + stainless steels +

steam generators + tantalum + thermal ccnductivity +

velding + alloy compcsiticn + microstructure + molten salts +

niobiun e tensile properties

OTHER CATEGORIES: EXX ¢+ FCX + GXxx '

ACX640008 .

(Staff Regcrt)

MCLTEN-SAIT RER2CTCR PROGRLH SEMIANNUAL PROGRECS REPORT FOR

PERIOD ENDING JANUARY 31, 1964

Cak Ridge kational Laboratory, Tean,

ORNL- 3626 (July 1964), 166 p.
Status and progress are reported. Ccntents ate abstracted
and filed in S5 parts: MSRE, Metallurgy, Radiation
Chemistry, Chemistry, and Fuel Frocessing..

Accession Numker ACE70003S tc ACXe6u4CC(8
.

Page 55

Catégory A
Molten-Salt Reacter Procrars

ACX64C008 *Ccntinued*
MSRP + progress report

ACX6uC018
(Staff Eeport)
MOLTEN-SALT REACTOR PROGRAM SEMIANNUAIL ERCGEESS FEECET
FOR PEEICL ENLDING JULY 21, 1964
Oak Ridge Naticnal Laboratory, Tenn. '
ORNL-3708 (Nov. 1964), 39% p. | |
This repcrt is a review 1in depth rather than a report
- of 6 months' progress., It is a ccllecticn cf paters
given at an infcrmation meeting at CENL, August 18-19, 1964,
near the end of MSRE construction, The pafpers ccver '
the backorcund and rerort the status of the techknclcgy
of molten-salt thermal-breeder reactors as of mid-1964.
.Separate abstracts are filed for eack cf 16 fpagers.
MSRP + prcegress report + reviews + technology
OTHFER CATEGCRIES" AEX

ACXG50003-

(Staff Report)

MOLTEN-SALT REACTOR PROGRAM SEFIANNUAI ERCGHESS FEECFT FOB

PERIGC ENCING FEERURARY 28, 1965

Oak Ridge Naticnal Labcratory, Tenn.

ORNL-3812 (June, 1965), 176 p.
Status and progress are reported. Contents are atstracted
and filed in 5 parts: ¥SRE, Fetallurgy, Radiaticnh

, Chermistry, Chemistry, and Fuel Prccessirg.

MSRP ® prcgress refgort |

ACX650009

{Staff Regpcort)

MOLTEN-3SAIT RERACTCR PROGRA% bEHIANNUAL PROGRESS REPCR1 FOR

PERIOD ENDING ADGUST 31, 1965

Cak Ridge National Lalbkoratory, Tenn.

ORNL- 3872 (Dec. 1965), 156 p,
Status and progress are reported. Contents are abstracted
‘and filed in 5 parts: ' MSRE, MNetallurgy, H&dlfitlflfl a
Chemistry, Chemlstry. and- Fuel Eroce551ng.

MSRP + progress report

ACX660007

(Staf £ Report) ‘

MOLTEN-SALT REACTOR PROUGRAN SEFIA&&UAI ERCGFESS EEECET FOB

PERICL ENCING FFERUARY 28, 1966 = =

Oak Ridge National Laboratory, Tenn,

CRNL-3936 (June, 1960), 216 p.”-"- ' : |
Status and progress-are- teported. Contents are ahstrccted
and filed in 5 parts: ESRE,,Petallurcy, Cben:ctry,

MSBE Design Studies, and KSR Processing Studies

MSRP + Eprcgress regort . '

ACX660013

Accessicn Number ACX64CCCE to ACX66CC13

 
 

Page 56

Cateqgory A
Mclten-Salt Reactcr Prograrns

ACX66C013 ¥*Ccntinued*

(Staff BReport)

MOLTEN-SALT REACTOR PROGRAF SEMIANNUAL PROGREESS REPCRT FOR EERIOD

ENDING AUGUST 31, 1966

Cak Ridge National Laboratory, Tenn,

ORNL—- 4037 (Jan. 1967), 242 p. | L
Status and proyress are reported. Contents are abstracted
and filed in 5 parts: HMSRE, MSRP Materials, Chemistry, MNSBR
Design Studies, and ESR ProceSSan Studies ' '

MSRP + progress report

ACX67C015

(Staff Report)

MOLTEN-SALT REACTOR PROGRAF SEEIA&LUB[ EBGGBESS EEECET FOE, EERIOE

ENDING FEEERURFY 28, 1967

Oak Ridge Naticnal Laboratory, Tenn.

ORNL-4119 (July 196p), 219 p. o
Status and progress are reported. Contents are akstrected
and filed in 6 parts : KSRE, Materials, Chemistry, In-Pile
Loops, MSBF Design Studies, and MSR Processing Sftudies

MSRP + prcgrese regcrt v

ACX670022

(Staff Repcrt)

MOLTEN-SALT REACTCR FROGRAM SEHIANNUAL PROGREES REPOR? FOR

PERIOD ENKDING AUGUST 31, 1967

Cak Ridge RKational Latoratory, Tenn,

ORNL- 4191 (Dec. 1967), 260 p. :
Status and progress are reported in six parts with
these titles: Mclten-Salt Reactor Experiment, MSER
Design and Development, Chemistry, Molten-calt Irradiation
Experirents, Materials Development, and Folten-Sait
Processing and Preparation., Separate atstracts are
filed for each fgart.

MSRP + progress report

ACX68C011

(Staff Report)

MOLTEN-SALT REACTOR PROGRAF SEMIANKUARL ERCGHESS FEECEKET.

FOR PERICL ENKDING FEERURRY 29, 1S€8

Oak Ridge Naticnal Laboratory, Tenn.

CRNL-4254 (Aug. 1968), 282 p.
Status and progress are reported in six parts with these
titles: Molten-Salt Reactor Experiment, MSBR Design
and Develogment, Chemistry, N¥olten-Salt Irradiation
Experiments, Materials Levelopment, and Molten-falt
Processing and Pregaration. Separate akstracts are
filed for each part,

MSRP ¢+ prcgress tTeport

ACX680018
(Staff Regcrt)
MOLTEN-SALT REACTCR PROGRAM SEMIANNUAL PROGRESS REPCRT FOR

Accessicn Number ACX6€0013 to ACX6ECCIE

~
*)

»

Page 57

| Category A
Mclten-Salt EFeactcr EFrograms

ACX680018 *Continued* ,

 PERIOD ENDING AUGUST 31, 1968

Cak Ridge National Laboratory., Tenn."

CRNL-4344 (Feb. 1969), 333 . - |
Status and progress are reported in six parts with
these titles: Molten-Salt Reactor Experiment, MSBR
Design and Development, Chemistry, Molten-Salt Irradiation
Experiments, Materials Cevelopment, and Molter-falt
Processing and Preparatlon. Separate akstracts are-
filed for each part. h | |

HSRP + prcgress regort

ACX690020

(Staff Refpcrt) ' '

MOLTEN-SAIT KEACTOR PROGRAM SEMIANNUAL PROGRE‘S REPCRT FOR
PERIOD ENDING FEBRUARY 28, 1969

Cak Ridge National Laboratory, Tenn.

ORNL- 4396 (Aug. 1969), 3C7 p.

Status and progress are reported in six parts with
these titles: Molten-Salt Reactor Experiment, MSEX
Design and Development, Chemistry, Molten-Salt Irradiation
Experiments, Materials Development, and kolten-Salt
Processing and Preparation. . Separate alkstracts are
filed fcr each fpart, '

MSRP + progress report

ACXh9C027

(Staff EReport)

MOLTEN-SALT REACTOR PROGRAY SEMIANNUAL ERCGFESS FEECET FOR

PERICD ENCING AUGUST 31, 19¢€9

Oak Ridge Naticnal Labcratory, Tenn.

CRNL-U4449 (Feb. 1970), 25z p.
Status and prcgress are reported in six parts with
these titles: Molten-Salt Reactcr Experiment, MSBR
Design and Developrent, Chemistry, Molten-5alt Irradiation
Experiments, Materials [evelopment, and Molter—calt
Processing and Preparation. , Separate atstracts are
filed for each part. except Irradzatxcn Experinerts
‘where there was no acthLtY.

VMSRP + progress report

ACX70C018
(Staff Report)

MOLTEN-SALT REACTOR PRUGRAK anIAthAI EHCGEESS EEECFT FOB

PERICC ENLCING FEBRUARY 28, 1970
Oak R:dqe Naticnal Labcratcry, Ienn.

ORNL-4548 (Aug. 1970), 338 p.

Status anrd progress are teported ih six parts Ulth these
titles: flfiolten-SaIt,Reactor Experiment, MSBK Design

and Develorment, Chenistry, Molten-Salt Irradiation
Experiments, Materials L[evelopment, and Molter—ctalt

Accessicn Number ACX680018 to ACKX7CCC1E

 
 

Page 58

Category A
Mclten-Salt Reactcr Prograis

ACX70C018 *Ccntinuedx L
Processing and Preparation. Separate atstracts ate
filed for each part excert Irradiation Experiments,.
vhere there was no activity.

MSRP + progress report

ACX700034

(Staff Rerport)

MOEDENPSRITCREKCINE BBGGBZHBSEH]!BBUAL PROGRESS REPOR?

Oak Ridge Naticnal Labcra tory, Ienn.

CRNL-4622 (Jan. 1971), 230 p.
Status and progress are reported in five pa:tc vith
these titles: Molten-Salt Eeactor Experiment, MSER Desan
and Levelopment, Chemistry, Materials Develcpuert, and
Mclten-Salt Prccessing and Preparation. Separate atstracts
are filed for each part.

MSRP + prcgress rerort

ACX710027

(Staff Refpcrt)

MOLTEN-SAIT REACTCR PREOGRAM SEMIANNUAL PROGREES REPGRI FOR

PERTOD ENDING FPEBRUARY 28, 1971

Cak Bidge National Laboratory, Tenn.

ORNL- 4676 {Aug. 1S$71), 277 . ,
Status and progress are€ reported in five parts with these
titles: Moltenr-Salt Reactor Experiment, MSEF LCesign and

- Development, Chemistry, Materials Develcpment, ard

Molten-Salt Prccessing and Freparation. Separate atstracts
are filed for each part. -

MSRP + prcgress regcrt

ADX640021

Kasten PR

THE MCSEL REACTCR CONCEPT

Kernforschurgsanlage Julich, Geraany

froc. 3rd Int. Conf. in Peacefnl Uses cf Atcmic hnequ.

- Geneva, (Aug. 31 - Sept. 9 1964), Vol. 6, pp. 363-3€9,

1 fig. 23 ref.

The Mclten Salt Eplthetmal (MOSEL) reactor ccrcejft featurec
a ccre fluid cf UF4~NaF-BeF2 and a blanket fluid of
ThF4-NaF-BeF4 separated ry nickel alloy tubes. Processing
is ty flucride vclatility. Farasitic aktsorption of
neutrons (predominantly epitherral) is noderate ard
breeding ratios between 1.08 - 1.22 appear attainable. The
core concentration is atout 300 ¢ fissile/liter ard fiscile
inventcry ccst is a major item. Nevertleless fcwer ccsts
appear reascnable,.

*breeding performance + *conceptual design + *mpolter salts +

*reactors + costs + fcreign + flucrination ¢+ Hastelloy N +

inconels + inventories + neutron sgectra + nickel +

Accessicn Number ACX700018 to ALCX640021
.}

at

Page 59

Category A
Mclten-Salt Reactor Frograas

ADX64 0021 *Continued*
sodiur flucride
CTHER CATEGCRIES: IAF-

ADX67COUE
Gat U
COOLING CONCEPTS FOR A COMPACT MOSEI (MCLTEN SALT) FEACTOF

Kernforschungsanlage Julich, Cermany

Nucl,., Engrg. and Design 5 (1967), pp. 113-122, 10 tig,

23 ref. _

This review of englneerlng possibilities of the MCSEL
reactor considers cooling by direct ccntact cf c=alt ard
molten lead, either internally or externally to the ccre.
Internal cooling reduces the fuel inventcry ard fuel cycle
costs but concenttate= englneerlng problems in the core
ZOBE.

*conceptual design +_*lead coollnq + *reactors + cores +

foreign + heat transfer + inventories

OTHER CATEGORIES: IAF :

ADX690063
Jensen RJ + Swanscn E N
UTILITY AFFLICATION CF MOLTEN-SALT BREELCER REACTOR
Northern States Pcwer Co, Minpearolis + Black and Veatch,
Kansas City
Proc. American Power Lonf. 31 (1969;. pp. 222-230, & figqg,
7 ref.
"The MSBR concerpt, supported by the successful MSFE, is under
study by utilities as a contender with fast breeders for
lcng-range power., 1Its low breeding ratio is offset by
low fissile inventory and-it can use U-233, U-238, cr
Pu as initial fuel. On-site processing is required. A&
schedule for 1976 startup ot a ?CC— to £C0-M%e prctctyge
is rreserted.
*electrlcal power + *zndustry t *MSER ¢+, *protcty;ec +
*utilities + atchltect-enqlneerlng + breedlnq pertcrmance +
converters + economlcs + natural re@ources -

Accession Number ADX64C0Z1 tc ALX6SCCE3

 
 

60

Page
Category B
Reactor Analysis
BAX68C006
Perry AN

INFLUENCE OF NEUTRON DATA IN THE DESIGAN CF CTEEER TYEFES CF
FCWER REACTOEFES

Oak Ridge Naticnal Laboratory, Tenn.

0RNL-TH-2157 (karch 1968), 23 p, S fig, 10 ref,

This repcrt was presented at the Second Conference on

Neutron Cross Sections and Technology, Washingtcr, Naich

4-7, 1968. The effects of cross-sectior uncertainties
on estimates cf breeding perfcrmance and of pcwer
cost for a molten-salt Ltreeder reactcr are shcwrn to

be small. Uncertainty in breeding ratio due to cross-
section uncertainties is less than fplus cr rinus C.C2,
and the uncertainty in rover costs is less than plus
or minus 0,3 mills/kvhr {e). Similarly small effects
are shown fcr the high-temperature gas-cooled reactor.
The need for further refinements in nuclear data is
related primarily tc the calculation of temperature
coefficients of reactivity.

*breeding rerfcrmance + *cross sections + #*design data ¢

*arrors + MSPR + nuclear analysis + fpower ccsts + reactivity +

reactcrs + thoriur 4 uranium-233
CTHER CATEGCEIES: PBFX

BAX700008

Kasten PR ¢+ Craven CWw + firtht RC

CROSS-SECTICN ANC NUCLEAR-CONSTANT LCATA FOR HEAVY METAL
NOCLIDES (FUELS)

Cak Ridge hNational Lalkoratory, Tenn.

OBRNL-IM-2851 (Rev.) (Apr. 1970) 21 p, 18 £ig, 0 ref,

Cross sections and nuclear constants cf fissile ard
fertile paterials and cf higher isotopes are summarized
in graphical form, tased on ENDF/B data. The resultirg
fiqures fermit visual appreciation of nuclear data

in present use, and reletive ccmparison cf data fcr

the different fissile, fertile, and higher-isotoge
materials, Nuclides considered are Th-:232, U-23E,
Pa-233, Ng-239, U233, U-235, Eu-239, EFu-24%1, U-234,
J-23¢, Pu-240, and Fu-242. The revisicr differs from
the original report only in the quality of rejrcducticn
of grarhs.

¥neutron physics + *cross sections + *pneutrcn yield +
fissicn ¢ abksorption + *fissile materials + thorium +
uranium + plutonium + neptunium + *¥*data + isotcpes +
capture + *fertile materials :

BBX670012 -

Carlspith RS + Bennett L1 ¢+ Edison GE + Gift EH + Thomas WE +
¥elfare FG

REVIEXN OF MOLTEN SALT REACTOR PHYSICS CALCUIATICKS

Cak Ridge Kational Latoratory, Tenn.

Accessicn Number BAX630006 to BEX670012
»

Page 61

Category B
Reactor Analysis

BBX670012 *Continued*

ORNL-TM-1¢46 (Aug. 1967) 59 p, 13 fzg, 41 rerf.
A set of calculatlons was made to check the reactivity
and breeding ratio ¢f the reference design of a two-
region, two-fluid MSER. The review ccvered crcss
secticn selection, fission product treatment, multigroup
cell calculations, 'two-dimensional reactcr criticality
calculations, equilibrium depletlon calculations and
startup depletion calculations., Inscfar as pcssible,
the cross secticns and calculations methods were made
independent of those used previcusly. The reference
composition gqave a k(eff)_of 0.95. This discrepancy
was traced to use of a low value for thcriur rescnance
integral in previous calculations. When the reactor
was made critical by the addition of 14% mcre U0-233,
the bteedlng ratio was 1, 062 compared with 1,(C54 in
the preV1ous ‘calculations. ' Beoptimization cf the
comp051tlcn would prokatly decrease tkis dlffetence
in treeding ratic.,

*calculations + MSEBF + *two-fluid reactor +

*breeding perfcrrance + reactivity + burnup +

criticality + cross sections + fissicn prcducts +

computer ccdes + nuclear analysis + #*design data +

*errors + *reviews + thorium '

OTHER CATEGORTIES: BFX

EFX680009

Perry AN + Smith OL + Kort HT .

NEW DEVELCEMENTS IN MSR PHYSICS

Oak Ridge Naticnal Laboratecry, Tenn.

Summary of paper presented at ANS Winter Meet:ng,

Wasbington, D. C., Nov. 10-15, 1968, ANS Transactions 11,

(2) 619, 2 pe. L ' - '
Develcpments which rermit separation of protactinium and
fission products from-MSR fuel and from each cther raise the
possibility of a sirgle~fluid ¥SEF, The f£luid which

contains uranium and thorium is rade tc functicr as bcth
fuel and fertile materlal by adjustlng the degree of neutron

~moderation that occurs in varicus regicns cf a zcned ccre.

- Zoned cores have higher. yleld= and lower fuel inventories

~ than uniform cores. Power casts can ke reduced (.1 tc 0.3

- milsskwh(e) below those fcr’ comparable uniform cores.

- Temperature coefficients of react1v1ty in zcred-ccre
sincle-fluid reactors are such that dynamic claracteristics
are expected tc be acceptable. (Ereprints of this parer are
not available but similar, and mcre recent, data are : '

- presented in Nucl. Appl.tand Tech. 8, 208 (Felk. 1970). 3See
BFX700016.) _

*MSBR + *single-fluid rpactor= + *breedlng performdnce +

reactivity + dynamic characteristics + specific inventory +

nuclear analysis + neutron th:lCS + klanket ¢+ mcderatcrs +
fuel cycle ccsts

BFX700016

Accessicn Number BBX67C01z to BFX7CCC1e

 
 

 

Page 62

Category B
Reactor Analysis

BFX70C016 ¥*Ccntinued¥

Perry AM ¢ Pauman HF

REACTOR PHYSICS AND FUEL CYCLE ANAIYSES

Cak Ridge National Laboratory, Tenn,

Nucl. Appl. Tech. 8, (2) 208 (Feb. 1970), 12 g, 12 fig,

5 ref. :

' General nuclear characteristics, breeding performance and
fuel-cycle costs are discussed for a reference design,
single-fluid MSBR crerating on a thoriur -- uranium-233 fuel
cycle with full chemical processing. Tlis desigr has a
breeding ratio near 1.06 specific fissile inventcry of
1.5 kg/Mw(e), fuel doubling time of 20 yr, amd a fuel cycle
cost near 0.7 mil/kwh({e). Either enricked urarium cr
plutconiup may be used as a startup fuel. 1f chemical
processing for Pa isolation and rare-earth rescval is
oritted, the design has a conversion ratio of 0.8 to (.9.
The fuel cycle cost penalty for cperaticn as a ccnverter is
around 0.1 zil/kwh (e) |

*MSDR + *single-fluid reactors + *fuel cycle costs +

*breeding perfcrrance + nuclear analysis + thcrium +

uranium-233 + plutonium + processing + specitic invertcry +

rare earths + ncoble metals ¢+ fission products + ‘

rare gases # protactiniunm

BFX70C056

Carlsmith FS + Lane JA

POWER REACTORS FOR THE FUTURE ~-- AN EVAIUATICH

Cak Fidge National Latkoratory, Tenn. ,

Proc. American Pcwer Conf. Vol. 32 (1970) pp. 98-104,

4 fig, 16 ref. | |

A review by the AEC c¢f the U.S. civilian nuclear power
program consisted of twec phases: detersinaticn cf the
characteristics of varicus reactor types and simulatica of
optimal growth patterns using these reactors. . 1This
paper refports cn the first phase including uranium, thorium
and plutonium usage, separative wc¢rk requirements, and
carital, crerating, fuel, and total power costs for 7C
fueling variations in 7 reactcer ccncepts. 1Tke figqures
repcrted are those that sere available in 1967, 2An MSER
with 1.1 kg/MWe specific inventory, 1.07 breeding ratic and
capital ccsts around those for liqgqht-water reactors had a
total power cost slightly lower than any other reactor.

*breeding performance + *power costs ¢ *reactcrs + AEC +

capital ccsts + econcmics + fuel cycle costs +

optimizations + natural resources

BGX67C04E

(Statf Report)

SAFETY STUDIES FOR MSBR (FART 5 NUCL SAFEY EECG ANN ERC
REET 12/31/67) , _ -

Qak Ridge National Labcratory, Tenn.

Accession Number BFX700016 tc EGX670045
»

a)

-

Page 63

Category B
Reactor Analysis

BGX670045  *Continuedx*

ORNL- 4228 (April 1968), pr. 287-307, 8 fig, & ref.
These studies, the first reported as part cf the Safety
Prcgram, are aimed at information needed for satety
criteria for MSPBER's, 1Indications are ttat MSER systens
have favcrable inherent safety and stability
characteristics, that fission-prcduct tehavicr strongly
influences emergency cocling requirements, and that MSBR
materials are compatitle.

*analysis + *design criteria + *MSBF + #*safety +

afterheat o compatibility + depositicn +

fissicn prcducts + molten caltc + react1VLty + stability +

dynamic characteristics :

 

1
|

MAccession Number BGX67C04E to BGX67CCUS
 

 

Page 64

Category C
keactor Chegristry

CAX68C032

Thoma BRE

CHEMICAL FEASIBILITY OF. FUELIMG MOLTEN SAILT EEALTGES WITH

POF3

Oak Ridqge National Laboratocry, Tenn.

ORNL-TM~-2256 (June 1968), 37 p, S5 fiq, <0 ref.
The feasibility of starting molten salt reactcrs wlth
‘PuF3 was evaluated with respect tc cherical ccrpatibility
within fuel systems and to resoval of plutonium from the
fuel by chemical reprocessing after Pu-239 burncut.
Ccmpatibility within reactor containment systems is
moderately well-assured but requires ccrfireaticr of FuF3
sclubility and oxide tclerance before tests can te made
using the MSRE, Although separation cf plutcricr and
protactinium in the chemical reprocessing plant, as wculd
be desirable in a large treeder reactcr, has rct yet Leen
demcnstrated, cconcegrtual designs of processes for
effecting such separations are availaktle fcr development.

*chemistry + *cowmpatibility + *dissolving + tfuels +

*plutonium fluorides + *primary salt + *processing +

*separat1cn= + actinides + berylliup fluoride +

carriers o chemical properties + copcentraticr +

corrosicn + fluorides + graphite + Hastelloy X +

lithium fluoride + oxides ¢+ physical prcperties +

reactors + replacement + stability + thorium fluorides +

uranium fluorides

OTHER CATEGORIES: LDA

CAX690052

Fredricksen JA + Gilpatrick IC + Barton CJ

SOLUBILITY CF CERIUM TRIFLUORILCE IN MOLTEN MIXTURES OF

LIF, BEF2, AND THF4

Cak Ridge National Laboratory, Tenn,

ORNL-TM-233%5 (Jan. 1969), 25 . 9 £fig, 7 ref.
The Solubility of CeF3 was determined at varicus
temperatures irn six mixtures >f LiF, BeF2, ThF4 of the type
that wmay be used to fuel a molten salt kreeder reactor,
Comparison of earlier data on the solubkility c¢f PuF3 and
CeF3 in fluoride sclvents maka2s it possible to predict that
the solubility of PuF3 in single-region fuel ccrpcsiticns
at reactor operating temperatures will te more than
.adeguate. The solutility data as a functicn cf sclvert
compcsiticn vwere best ccrrelated by a model that assumes
BeF2 to Lbe complexed as the BeFY4 (~) icr and ThF4 as the
ThFZ(1-) icn.

*beryllium fluoride + *dissolving + *fluorides +

*liquidus ¢ *lithiur flucride + *phase equilikria +

*plutoniun fluorides + *rare earths + *solutkility +

*thoriuc fluorides + actinides + chepistry + compatikility 4

fissile materials + fuels + mixtures + MSBE e MSBR +

replacement + sclidus

CAX690053

Accessicn Number CAX63C032z to CR2X6S0CE3
a9

Page 6%

Category C
Reactor Chemistry

CAX690053 *Continuedx*
McDuffie HF + McCoy HE + Robertson EC + Scott [ ¢+
Thoma RE
ASSESSMENT OF MOLTEN qAI.'Ic AS INTEEREEDIATE CCCLANTS FCE
LMFERS

Oak Ridge Naticnal Labcratory, Tenn.

ORNL-TM~-2696 (Sept. 1969), 29 p, 7 figq, 23 zef,
Several molten salts were considered as intermediate
coolants for LMFBR's, Included were flucride, chloride,
carbecnate, nitrate-nitrite and fluoroborate salts.
Chemical reactions that could cccur Letween scdiur and
flucrchorates lead to the conclusion that cartonates
might be a ketter choice for LMFBR's, Use c¢f carbonates
avoids the safety ccnsiderations and related costs thet

" arise from the reactions of sodium with water if stean

generator fails and with air if a coolant pipe ruptures.
In the absence of these safety ccnsideraticns, scdium is
clearly sugericr tc the mclten salts as an intermediate
coolant for LMFEBER's because the lcwer tlermal ccrductivity
and higher visccsity of the salts would result in higher
equipment costs.

*carbcnates + *chenmical p:opertles + *chemical reactions +

*chlorides + *coolants + *fluorides + *flucrctorates +

*liquid metals + *LMFBR + *Nak + *physical properties +

*secondary salts + accidents ¢+ afterbeat +

applicaticns + behavicr + borcn trifluoride + compatitility +

concentration + containment + corrosion + deccmpcsiticr +

density + econcmics + emergency cooling + failures +

- heat exchangers + heat transfer + leakage ¢ leaks +

liguidus + lithium chloride + lithium fluoride + mixtures +
phase equilibria + potassium fluorides + safety + '
sodium flucride + solidus + specific heat + stability +
steam generators + thermal conductivity + v1=c051ty

OTHER CATEGOKIES: CCX + CEX

CAX690061

Thoma RE + Ricci JE R

FRACTIONAL CRYSTALLIZATION REACTIONS IN THE
SYSTEM LiF-BeF2-ThF4 ST

Cak Ridge National Laboratory, Tenn,
'ORNL*TH 2896 (July 1969), 33 g, 16 fig, 9 ref.

'Equlllbrlum and non- equ111tr1um crystallizaticn react:cnc
'in the systen LiF-BeF2-ThF4 are analyzed in relation to
their potentlal appllcat1on tc mclten salt reactcr fuel
reprocessing. Hetercgeneous equilibria in the temperature
range from the liquidus at 5S9C deg C to the cclidus at

350 deg C are described: guantltatlvely and in detail ly

. means of ten typical iscthermal secticns and ty three

‘temperature-compcsition cections, The implications of
metastable fractionation in this temperature interval are
discussed as a pecssible feed control step in reductive

Accession Numker CAX6900F%: to CAX64(C(CH1

 
 

Page 66

Category C
Reactor Chemistry

CAX69C061 *Ccntinued%
extraction reprocessing of molten salt treeder reactct
fuels.

crystallization + data + fluorides + freezing +

measurement + fprocessing ¢+ serarations

CAX710023

Mailen JC + Smith FJ + Ferrics LF

SOLUBILITY CF ELUTONIUM THRIFLUORILE IN MOLTEN 2 LITHICUM

FLUCRIDE~-BERYLLIUM FLUORIDE

Cak Ridge National Laboratory, Tenn.

Je. Chem. and Enyg, Data, 12 (Jan. 197Y%), 2 p, 1 £ig, 7 ref.
The solubility of plutonium trifluoride in mclter 2 lithium
flucride-beryllium fluoride was determined over the
temperature range of 55C-660 deg C., The results can te
expressed by the least-squares equation: 1log S(mole %
plutonius trifluoride) = 3.Z305 - 3C9€,1(deq K). The
s0lid phase present at equilibrium was probably pure
plutonium trifluoride. _

*molten salts ¢« actinides + flyorides +

*plutonium fluorides + *solulkility ¢+ MSRE

CCX680033

Kohn B8W

BUBBLES, DECES, ANLC ENTRAINMENT IN MOLTEN SAL1S

Oak Ridge Naticnal Laboratory, Tenn,

ORNL-TM-2373 (Lec. 1968), 21 p, S fiq, 42 ref,
The authcr describes prcduction ot droplets from
splashes and bursting Ekukbles and repcrts exgerirents with
mclten salts which shcwed that jet drops could
preferentially remove a surface film. He ccncludes that
this ghenorencn could ccntribute to remcval of metallic
fission products from the fuel salt in the MSRE pump Lowl.

*bubtles + *entrainment + *fission products + *molten salts +

tgas injection + *gas separation + #interfacial tensicp +

*mists + *MSRE + *ncble metals + *surface tension +

beryllium fluoride + chemistry + circulation + ccver gas +

experiment + fissile materials + fluorides ¢+ foaming +

fuels + gases + inert gases + lithiur flucride +

molybdenur + off-gas systems + fumpgs + sprays +

viscosity + void fractions

OTHER CATEGORIES: CFX + CJX

CCX680038

Cantor S.

PHYSTCAL ERCFEERTIES OF MSER EUEL, COOJLANI, AND FLUSH SAL1S

Oak Ridge Naticnal Labcratory, Tenn..

ORNL-TN-2316 (Mug. 1968), 49 p, 2 fig, 49 ref.
Experimental values or estimates are given for EtOPCEtleb
of seven salts of interest for MSBR's. Prcgerties
include viscosity, thermal copductivity, electrical

Accession Numker CAX69C061 tc CCX6€0(C38
 

"

u}

Page 67

Category C
Eeactor Chemistry

ccx68C038 *Ccntinyedx
conductivity, phase transition tehavicc, specific heat,
heat of fusion, density, expansivity, compressikility,
vapor pressure, surface tension, and gas sclukilities.
*data + * fluorides ¢ #flucroborates + *physical properties +
density ¢ solubkility + specific heat ¢+ surface tensicr +
therrxal ccnductivity + viscosity

‘CDX670035

Malinauskas AP + Rutherfctd JI + EvanslIII) RE ,
GAS TEANSECET IN MSRE MOLERATOR GRAPBITE., I, REVIE& OF
THEORY AND COUNTERDIFFUSION EXPEERIEENIS

Cak Ridge Rkational Labtoratory, Tenn,

ORNL- 4148 (Sept. 1967), 39 p, 7 fiq, 6 ref.
The authors develop equations descritinc gas trarspcrt in
porcus media. The experimental findings are limited lLut
significant. Under MSRF conditicns it afppearts gulte
justifiable to igncre ncrmal diffusion effects in gas
transport computations so that all thke casecus-diffusicn
infcrration necessary tc correlate flelon-product-
migration data can te gained through simple permeability
measurements, The wcre complex interdiffusion experiments
are not required. Thus a comfplete flow-prcrerty survey
of all MSRE moderator materials can be performed with a
minimum expenditure of time and effort.,

*diffusion + *xfissicn ptcduct= + *gas flow + *grapnxte +

*MSRE + *xenon + analysis + rehavior + concentraticr +

deposition + design criteria + distribution + films +

fluids + inert gases + materials + mpeasurement +

moderators + physical frererties + rare earths + testing

CTHER CATEGCEIES: FEX B

CEX640018

Blankenship FF

EFFECTS OF RADIATION ON THE CCNPATIBILIIY CF MSRE MPTIEIALS
(PAET CF MSRE SEMIANN PEOG REPT 7/3 1/6&)

Oak Bidge Naticnal Labcratory, Tenn.

'cauL-37oa (Nov. 1964), pp 252-2€67, 1€ tig, 4 ref.

Capsules cortaining fuel =alt graphite, IKCF-& and
~ molybdenum were irradiated in the MTE ard later exalined
~at CENL. Fnhanced attack and other ancsalcus effects
‘appeared tc be due to fluorine that was produced Lty -
tddlolySLS of frozen salt at low temperature. Much of this
~article is ccncerned with investigation of this phenotwenon.
Typlcal radiolytic yleld of flucrine trcm frczer fuel salt
vas 0.02 molecules per 100 eV absorbed energy. Internal
' recombination was sufficient to prevent any evcluticn of
gaseous fluorine at temperatures above about 80 deg C.
capsules + compatibility + experiment + flucrine +
graphite + in-pile tests ¢+ irradiation + paterials +
molten salts + molybdenum + MSRE + radioclysis +

Accessicn Number CCX680038 to CEX64CC1E

 
 

Page 68

Category C
Reac tor Chemzstry

CEX640018 *Continued%*
recomtination

CLX700010

Haukenreich PN , ,

" FLUORINE ERCCEUCTION AND RECOMEINATION IN FROZEN MSR SAL1S

AFTER REACTOR OPERATION

Cak Ridge National Laboratory, Tenn.

ORNL-THN-3144 (Sept. 1970) 36 f, 9 fig, 12 ref.
Exposure of capsules of MSR fuel =alts in the MNTR betseen
1961 and 1964 showed that vhen the salt was chilled
below about 80 deg C, F2 was produced by radiolysis
at a rate of 0.02 molecules/1CC ev. Other experiments
ccnfirmed the radiclysis cf frozen salt and provided
data on the effect of temperature on recombinaticn.
The data on yield ard reccmbination have recently been
revieved and used in ansvering questicns invclved 1in
storing and disposing of irradiated salt from the MSRFE
and future molten-salt reactors. The €erergy scuice
in the MSRE salt is low enough that no fluorine evolution
is expected for over a year after heatitrtg tc induce
recorbination. Salt from a high-power FSR can te stored
in kare cans vwith no fluorine evcluticn if the suvrrcurdings
are kept at abcut 260 dey F.

*fluorine + *polten salts + *radxoly51= + %*stcrage +

*wyaste disposal + afterheat # analysis + experiment +

heat transfer + MSRE + primary salt ¢+ reacticr ratecs +

recomkinaticn

CXxo40020

Grimes WR _

CHEMICAL BASIS FCF MOLTEN-SALT REACTOBRS (PAKT OF MSEKP

SEMIANN PROG REPT 7/31,/64)

Cakx Ridge National Latoratory, Tenn.

ORNL-3708 (Nov. 1964), pp. 214-251, 25 ref.
Requirements of high-tenmperature fluid-fuel reactors
are best mwet by mixtures of fluorides including UF4 and
ThF4, which have lcw vafpor pressure, good heat transfer
properties, little parasitic atscrpticr ci neutrcros,
and immupity tc radiaticn damage. The selection and
characteristics of MSRE fuel and cooclant salts are
discussed. -

chemistry + coolants + fluorides + fuels + mclten salts +

MSRE + MSRP + rhase equilibria + physical progerties +

reviews

CXX70C049

Grimes WR

MOLTEN-SALT REACTOR CHEMISTIRY

Cak Ridge National Laboratory, Tenn. |

Nucl. Appl. Tech. 8, 137 (Feb. 1970) 19 p, 8 fiq, 58 ref,

Accession Numter CEX640018 to CXX700049

 
u}

Page 69

Category C
Reactor Chemistry

CXY¥70C049 ¥Ccntinued* '
Considerations leading to the chcice cf MSR fuel
comfpcsiticn are discussed under the headings: FEhase
Behavior Amonq Fluorides, Oxide Fluoride Phase Bebavicr,
MSRE and MSBR Fuel Ccmpcsiticns and Choice of Coolant.
Physical properties of fuels and coolant are takulated. 1In
connecticon with the chemical compatibility of MSER
materials, topics included are: Thermcdynamic Data fcr
Molten Flucrides, Oxidation (Corrosion) of Metal and
Compatibility of Graphite with Fluorides. Cherical
behavior in the MSRE is discussed in terms of Behavior of
the Fuel Components, the Corrcsicn Froducts and the Fission
Products, Separations chemistry is treated in terms f
Separaticn cf Prctactinium and of Fission Eroducts ty

- several methods including Reducticn. Wkile much research

and develorment remain to be accomplished, it is
demonstrated that there is no fundamental chermical
difficulty with design and operation of a single-fluid
molten salt breeder systen. |

*behavior + *bismuth ¢ *chemical properties +

*chemical reactions + *chemistry ¢ *cormpatiktility +

*coolants + *corrcsicn + *corrcsion products +

*equilibrium + *fission products + *fluorides +

*fluoroborates + *fuels + *graphite + *FSRE +

*noble metals + *oxide precipitaticn prccess + *cxides +

*protactiprium flucrides + *seraratiocns + actipides +

teryllium fluoride + berylliur cxide + boron triflucride +

chromium + components + concentraticn + density +

deposition + dissclving + experiment + fissile materials +

gases + Hastelloy N + liquid mretals + liquidus +

lithium flucride + mass transfer + melting + mists + MSFE +

MSBR + phase equilibria + physical properties +

primary salt + prctactinium + radiation damage +

rare earths + rare gases + reaction rates + fuel cycle +

sampling + single-fluid reactcrs + solidus + solubility +

thermal conductivity + thorium fluorides + '

uranium flucrides + uranium=-235 + vapor pressure,+

"viscosity + zirconium fluorlde-

Accessicn Number CXX70004S to CXX7CCCu4c

 
 

Page 70

Category E
-Graphite

EBX69C039

Greenstreet WL + Smith JE + Yahr GT + Valachcvic RS

THE MECHANICAL BEHAVIOR OF ARTIFICIAL GEAFHITES AS

PORTRAYED BY UNIAXIAL TESTS '

" 0ak Ridge National Laboratory, Tenn. -

ORNL-TM-2727 (Pec. 1969), 46 p, 27 fig, 5 ref.
Tensile and corpressive stress-strain curves were ‘measured
and combined with previous measurements to shcw behavicr of
several specimens of reactor-grade graphite, principally
AGOT, under several conditions of cyclic lcading and
unlcading. Hysteresis uas considerable but diminished
on successive cycles, Lecoming very small after several
cycles.

graphite + compressive properties + fat1gue +

tensile prcrerties + testing

EBX700041

Yahr GT

DETERFI NATICN CF BELATIVE THERMAL RUPTURE RESISTANCES OF

GRAPHITES

Cak Ridge National Laboratory, Tenn.

ORNL~- 44€7 (Jan. 1970), 47 p, 18 fig. 42 ref.
Polycrystalline graphite has remarkaktle resistarce to ,
therral-stress-induced fracture. Nevertheless, selection
of a particular grade ot graphite for certain afprplicaticns
must include ccnsideration of this property. <Currently the
type and grade of graphite'are often selected cr the kasis
of elastic analyses, since thermal shock tests are toc
expensive. for screening devices., This repcrt describes a
test rapid and econcmical enough for screening candidate
materials. Thin disks of graphite are teated at the center
with an inert-gas shielded-arc nonconsumable electrode
velder, each at a different, Ltut constant, fpcwer level.
The rinirum power input to the welder that will consistently
cause the graphite to fracture is detercined. The graphite
that regu1res the highest power level to produce a _
fracture is the one most resistant tc thtermal shcck. This
test ranked 21 grades, cr types, of graphite, In
appendix contains mechanical and therrcal fprcrerties of
the sgecimens, obtained from the literature for
‘determination of figures of merit ratincs cf rcrmal shcck
resistance., Ncne cf the standard figures ot merit gave
‘reliable predictions.

*graphite + elasticity + experiment + physical properties +

*rupture + *thermal effects ¢ *testing +

thermal rprcrerties + thermal shcck

EBX700042

Derqunov NN + Barbancv VKN + Kurakln VK + Zaitsev GG +
Strokov VI + Abrakhimuv U ,

SHORT-TERM STRENGTH, CREEP, AND DUCTILITY CF CFAEHITE AT

Accession Numkber EBXE€SCC3IY tc EBX7CCC42
”

"y

.

Page 71

Cateqbry E
Craphite

EBX70C042 ¥Ccntinued¥
300 TC 3500 TCEG K

Not given. ‘ . 7 .

LA-4462-TR (Sept. 1970), £ p, 4 fig, 1C ref,
A facility and proceduxe for studying the tensile,
impactr and compressive strengths, cree; and ductlllt) cf
graprhite at terperatures from 3CC to Z2(0 deg K are
described. Experimental data on the mechanlcal
characteristics and variations of the creep rate and
duct;llty of three grades of graphite ditfering in
grain size in this tempelature arLe presented. The
tensile and compres=1ve properties depended in a comp lex
fashion on grain size, temperature, and orientaticn; the
coarse-grained material was weakest. At about 3000 deg K,
the tensile behavior changed from brittle tc ductile and
the temperature derendence of creep increased greatly.
This document is a translation of Paper No., S ir the
Transactions of the Fifth All~Union Scientific-Technical
Conference, Kiev, Ukranian SSR, Decerker 19€¢7,.

creep + ductility 4+ #*graphite + experiment + %equipment +

microstructure + tensile properties + compressive fpIcferties

EBX70C043 \
Fontana A + Winand R
STUDY OF 7THE WETTABILITY CF GEAFHITE BY DIFFEFENT MCLTEN
SCDIUM-FLUCRIDE-ZIRCONIUM TETRAFLUORIDE-ZIRCONIUM DICXIDE
MIXTURES IN THE PRESENCE OF VARICUS GASECUS ATECSEEEFES
Universite libre de Eruxelles, Belqium
J. Nucl. Mater. 35, 87 (Agr. 1970), 5 p, 4 £fig, & ref.
The wettablllty of graphite at 1050 deq C by nclten scdium
flucride containing ug to 25% zirconium fluoride and 3.75%
zirconium oxide was studied by observing the contact éngle
of drops of the molten salt under a kct-stage micicsccge
in the presence cf different gases. The wettability of
- ..graphite by these mixtures teccme less as the ZIF4 ccrtent
“was increased and the 7102 content reduced. The
"wettability of qraph1te ty zirccnia-free mixtures
increased when an argon atmosphere was replaced*by COz,
~'while for mixtures contalnan 2rC2 the change of’
- atmosphere had no effect., CF4 and CO did nct appedt to have
any significant 1nf1uence on the wettability cf graphite by
 the mixtures examined. Polished graphite surfaces were
muck less well wetted by NaF-ZrF4-2rC2 mlxtures than
lachlned graphite surfaces. (auth) S
wettlnq + graphite + cxldes 4+ molten salts

BCX710011 - '

Chang SJ + Carpenter JA + Altcm DN o _

VISCOELASTIC ANALYSIS OF IRRAEIATED GRAPHITE EITH VARIABLE
CREEP COEFFICIENT

Cak Ridge National Laboratory, Tenn,

Accessicn Number EBX700042 to ECX7100 11

 

 
 

Page 72

tateqory E
Graphite

ECX710011 *Continyed*

ORNL-TMN-3242 (May 1971), 31 p, 5 fig, 3 ref.
This report is an addendum to ORNI-TM-2407 ccncernlnq a
methcd of stress amalysis for irradiated gragkite which may
be used for MSBR ccre design., To provide a refined
analysis, the present method includes tle effect cf a
variable creep coefficient which is caused by the
nonuniform temperature distriltuticn., Tc facilitate a
simple formulaticn, it is assumed that the temperature
dependence of the elastic resgcnse of tle gaterial is
approximated tc be inversely proportional to the creep
rate. It is shown that the prctlem reduces tc the
sclution cf several asscciated (fictitious) elastic
problems which have a common elastic rodulus inversely
propcrticnal to the creep rate of the irradiated graphite.
‘Numerical examples in the previcus report wvere recalcrulated
based on the present thecry. It shows, for larqge dose
values, an improvement to the previocus rethcd. A
computer program is written for the purpose and can
include the previous solution as a sgecial case,

stress + analysis + graphite + radiation damage +

elasticity + MSEE + creep

EDX64C016

Cook WH

MSRE GRAPHITE {IN MSRP PROGR. REPT., 7/31/64)

Cak Ridge National Laboratory, Tenn.

ORNL- 3708 (Nov. 1964), pp. 373-389, 10 fig, 18 rei.
The graphite purchases for the MSRE, grade CGE, is a rew
nuclear graphite that is basically an extruded getroleun

" coke bonded with coal tar pitch heated to 2300 deg C. Low

permeation is oktained through a series cf infreqraticns and
heat treatments., The final heat treatment was at 2800 deg C
or higher. Experimental equipment and frocesses were used
on a ccmmercial scale fcr the first time. The graphite was
produced as a 2z-1/2-in.-square€e x 72-in.-long kars, shich
vere machined to the required shapes. The average bulk
density is 1,86 grams per cukic centimeter, 1Its matrix is
not permeated by mclten salts under conditions more severe
than those expected in the MSRE. It exceeded all the
requirements specified for the MSEE except that it had
longitudinal cracks. Tests indicated tlat the cracks
should nct have any significant adverse effect on the
operation of the MSRE. The skrinkage of the craphite at
30 to 475 deg C under exposures between 0.60 x 10{20th)
to 1,40 x 10 {20th) neutrons/sg-cm (E greater than 2.9 MeV)
indicated that this should not create any important adverse
effects on the operaticn of the FSRE.

*graphite + *MSRE + cracks + density + elasticity +

examiraticns + fabricaticn ¢ inspecticn + intrusion +

irradiation + physical properties + progress repcrt +.

Accessicn Number ECX7 10011 to FLX64CC16
"

")

. Page 73
Category E
Graphite

EDX640016 *Continued* |

procurement + reviews + sfpecific heat + specificaticns +
thermal conductivity + heat treatments + tensile prcperties +
flexural prcperties + comfressive properties + micrecstructure
CTHER CATEGCEIES: ACE

EDX68C031 _
Kasten ER + Bettis ES + Cook WH + Fatherly &P +
Holmes DK + Kedl RJ + Kennedy CR + Kirslis S5 +
McCoy HE + Perry AM + Bohettson RC + Scott T +
Strehlow RA
GRAPHITE PBPEHAVICR ANB ITS EFEECT% ON MSBR PERFORMANCE
Oak Ridge Naticnal Laboratory, Tenn..
ORNL-TM-2136 (Feb. '1968), 97 p, 22 fig, 43 ref.
Grarhite behavior under MSBER conditions is reviewed and its
influence on MSPER performance estimated. The dejcsiticn of
. fissicn froducts cn grarhite does not agppear to lte large
(10 to 35 % of the nokle-metal fissicn products based on
MSRE experience). Taking into account graphite replacement
every two years, fission product depositiorn 1educes the MSBR
breeding ratic by about 0.00Z. Also, it appears that xenon
poisoning can be keit at a 0.5% fraction poisoning level Ly
using pyrolytic carkon as a pore impregrant tc seal the
surface andyor by efficient gas stripping of the fuel salt
fluid by xnjectlon and removal of heliur bubbles
Published in slightly abbreviated form in MNucl. Eng.
Design 9, 157-95 (1969).
+MSBR + *graphite + *reviews + irradiation +
mechanical properties + creep + develofrent +
kreeding perfcrmance + xenon |
EDX690051 ) ,
Kasten PR + Bettis kS + Ccck ¥H + Eatherly WE +
Holmes CK + Kedl RJ + Kennedy CR + Kirslis S5 +
McCcy HE + Perry AM + Rcbertccn RC + Scott [ +
Strehlow EA
GRAPHITE BEHAVIOR AND I'Ic EFFELIS Ch MSER LERECREANCE
Cak Ridge National Laboratory, Tenn,
Nucl. Enq. Design 9, 157 (Feb. 1969), 39 p, 18 fig, 40 ref.,
- Graphite behavior under MSBR ccnditions is reviewed ard its
- influence on MSBR perfcimance estimated. The deposition of
- fission products on graphite does nct afrear tc be laige
(10 tc 3%% cf the ncble-metal fission products tased on MSRE
" experience), Taking into acccunt gragphite reglacenment
~every tvwo years, fissicn product deposition reduces the MSBEK
"breedlnq ratio by atout 0.002. Alsc, it appears that xenou
- poiscning can be kept at ‘a 0.5% fracticr pciscricg lewl by
using pyrolytic carben as a pore impregnant tc seal tte
surface and/or by efficient gas stripricg cf the fuel salt
fluid by injecticn and removal of heliur bubkles.
Published with somewhat more detail as CRNL-T14-2316.

Accessicn Number EDX64C01€ to FLX6SCCE

 
 

Page 74

Category E
Craphite

EDX69C051 *Ccntinued*

#*MSBR + *graphite + *reviews ¢ irradiation +
mechanical properties + creep + development +
breedlng performance + xenon

- EXX70C048
Engle GB + Erice RJ + Eokros JC + thte JL
RESEARCH ON GRAPHITE -- THREE-YEAR SUMEARY EFEECET

May 15, 1967, through May 14, 1950

Gulf General Atomic, San Diegoc, Calif.

GA-9975 (June 1, 1970), 111 p, 36 fiqg, 47 ref.
A detailed sumgary is given cf work at GGA related to the
formation, properties, and irradiaticn jerfcrmarce cf
grarhite; other forms of carbon were studied to cowmplement
work on graphite. The seven main topics are Mcrfpholcgy of
the Carbcnaceous Mesothase Fcrmed in the Eyrolysis of
Coal~Tar Pitch, Catalytic Gréephitizatice of Pyrclytic
Carkcns, Petroleum Cokes, and Graphites, Characterization
of Graphites, Model Materials, Annealinc¢ of Irradiated
Pyrclytic Carbons, Highly Oriented Graphites, and
Artificial Commercial Graphites., Refpcrts giving zore
detail are listed.

carbon + coke + creep + density + expansion + fabricaticn +

graphite ¢ irradiaticn + fhysical fproperties + pyrocarkon +

thermal conductivity + x-rays + lattice + heat treatments +

picrostructure

Accession Numkter EDX690051 to EXX700048
-

-}

Page 75

Cateqgory F
Hastelloy N and Related Allcys

FAX62C004
DeVan JH + Evans (I1T) RB
CORROSION BEHAVIOR OF REACTOB MATERIALS IN FIUCRILCE S2LT
MIXTURES
Oak Ridge Naticnal Labocratory, Tenn.
CRNL-TM-328 (Sept. 1962), 35 p, 10 fig, 12 ref.
The rerort discusses (1) corrcosion experiments dealing with
fluoride salts in support of the FKSRE, and (2) analytical
methcds employed tc interpret corrosicn and rass-transfer
‘behavicr. The rrodicts of ccrrosion are solutle in tthe
molten salt; accordingly passivation is precluded and
corrosion degends directly on the thermodynamic driving
force of the corrosion reactions. Conrpatibility, therefore,
~demands salt ccnstituents that are not appreciatly recuced
by useful structural alloys and container raterials whcse
compcnents are near thermodynamic equilibrium with the salt
medium, Utilizing information gained ip ccrrcsicn testing
of ccmmercial alloys . and in fundamental interfpretations of
the corrosion process, CRNL developed a high-strength
- nickel-base allcy ccntaining 17% Fo, 7% Cr, and 5% Fe.
Several long-term corrosion lcops and in-pile capsule tests
ccrrleted with thicallcy demonstrate the excellent
corrosion resistance to fluoride salt mixtures at high
temperatures. Therrmodynamic methods are presented for
predicting corrosion rates. Radiotracer studies ccnfirmed

- the ccrrcsicn model. Also puiblished as pp. 557-579 in
Corrosion of Reactor Haterlals, IAEA, Vienmna, 1€62,
Vol. IT,

alloys + compatibhility + #*corrosion + *gdevelciment +
tluorides + *Hastellcy N + locp + nickel + fuels +
thermodynamics + molten salts

OTHER CATEGORIES: FBD

FAX620005

DeVan JH + EBvans{IIT) PB ,

CORROSICN EEHAVIOQOR OF REACTGR MATEHIALQ IN FLCORIDE qALI

‘MIXTORES . :

Cak Ridge Latlonal Laboratory, Tenn.'_ |

Corrosion cf Reactor Mate:1a1-,,vo1. IT (Proc. Conf.

June 4-8, 1962) IAEA, Vienna, p. 557, 23 p, 10 tig, 12 ref,
“This paper covers the =ame mdterlal as OdNL 1H-328

- (FAXe2C0CH) . |

alloys + compatibility + *COIIOClon + *development +

fluorides + *Hastelloy N + 1ocp + n1cke1 + tuels +

thermodynamics + #*molten salts _

OTHER CATEGORIES" FBD '

FAX690035

DeVan JH

EFFECT CF ALLCYING ALLDITICNS ON CORROSION BEHAVIOR OF
NICKEL-MOLYBLENUM ALLOYS IN FUSED FIUCRINE PFIXTUGFES

Accession Numker FAX6zCOCUL tc FAX6SCC3S

 
 

 

Page 76

Category F
Hastelloy N and Related Allcys

FAX69C035 ¥Ccntinued®

Cak Ridge National Laboratory, Tenn.

ORNL-1IM~-2021 Vcl. 1 (May 1969), 45 p, 13 fig, 16 ref.
Corrosion properties of nickel-wclykdenum allcys with
varicus solid-solution strengthening additions were
tested in thermal convection loofps, which circulated -
salt rixtures betvween 815 and 650 deg C. The alloys
contained 17 to 20% Mo and verious percentages cf Cr,
A1, Ti, Vv, Fe, Nb, and %. Lcops of individual alloys
were exposed to the salt mixture NaF-LiF-KF-UF4
{(11.2-45.3-41.0-2.5 mole %) for 500 and 1000 hr. The
corrosion susceptikbility of alloying additicrs ircrea sed
in this crder: Fe, Kb, V, Cr, W, Ti, and Al.
Metallographic examinations showed relatively light
attack fcr all allcys except those containing ccmkined
aluminua and titanium or alusinum and ctrowiuz. A
nickel-base allcy ccntaining 17% ko, 7% Cr, and 5% Ffe,
designated Hastelloy N, had the Lest ccrbinmaticr ct
strength and ccrrosion resistance among the ccmpositions
tested., ‘

alloys + compatibility + corrcsion + development +

fluorides + fuels + molten salts + HBastellcy N + lccp +

nickel allcys ¢+ allcy compositicn

CTHER CATEGCEIES: FEL

FAX69C04<

Devan JH

EFFECT OF ALLOYING ADDITICNS CN CCHECSICN EEHRVICE CF

NICKEI-FCLYBLENUM ALLOYS IS FUSED FLUORIDE MIXIURES

Oak Ridge Naticnal Labcratcry, Tenn. .

ORNL-TM-2021 Vol. 1 (May 1969), 45 p, 13 fig, 16 ref.
corrcsicr properties ctf nickel-molybdenum allcys with
various solid-solution strengthening additicrs were tested
in thermal convecticn lcops, shich circulated salt mixtures
betvween 815 and 650 deg C. The alloys ccntaired 17 tc 20%
Mc and varicus percentages of Cr, aAl, Ti, V, Fe, N, and W,
Loops of individual alloys were exposed to the s=alt mixture
NaF-1LiF-KF-UF4 (11, 2-45.23-41.0-2.% mole %) tor 5CC and
1C0C hr. The corrosicn susceptibility c¢f alloying .
additions increased in this order: FE, NB, VvV, Cr, &, 1i,
~and Al. Metallocgrathic examinations showed relativeiy light
attack for all alloys except those ccntaining ccrbired
aluminum apd titanium cr aluminum and chromium. 2 nickel-
base alloy containing 17% Mo, 7% Cr, ané 5% Fe, desigrated
Hastelloy N, had the best ccmkination of strength and
corrosion resistance among the ccriositions tecsted,

compatitility + cocrrcsion ¢+ develcpment + fluorides +

fuels + Hastelloy N + loop

OTHER CATEGORTIES: FBD

FBA660020

Accessicn Number FAX620035 to FEA66002Z0
)

Page 77

Category F.
Hastellcy N and Belated Alloys

FBA660020 *Continued*

McCoy HE

STUDIES OF THE CARBON DISTRIEUTION IN EASTELLCY N

Dak Ridge Naticnal Laboratory, Tenn.

OBRNL-TM-=1353 (Feb. 1966), 24 p, 14 fig, 6 retf.
A srall heat of Hastellcy N was prepared in which a
portion of the carkton atoms were tagged as catbhcn—14, The
resgcnse tc heat treatment was studied to determine
whether the changes in mechanical properties cculd be
correlated with the cbserved changes in the carton
distribution, Although markedé segregation resulted,
the changes in wmechanical properties did not appear
to be related. A second okjective was to detersine.
‘whether the relatively large precipitate particles
in this alloy were carbides. These precipitates, 1n
bcth their stringer (lcu-temperature) and lamellar
{high-temperature) forms, were fcund'to be as lcw in
~carton as the ratrix cr lower. It is hypothe51zed
that the other alloying elemepts reduce the sclukility
of rolybdenur 1 n nickel so that the precigpitate is
‘basically nickel-nmclybdenum intermetallic compounds.

*heat treatments + weiding + *metallography + *Hastellcy N +

carbon + rechanical frcperties + ¥microstructure

FBA68Q029 -
Gehlkach RE + McCoy HE
FHASE INSTABILITY 1IN HASTELLOY N
Oak Ridge Naticnal Laboratory, Tenn.
Int, Sym. on Structural Stability in Superallcys,
Seven Sgrings, Pa., Sept. 4-6, 1968, Vol, I1, pp 34¢-366.
Available from Dr. John Radavich, AIME High-Terfperature
Allcys Ccrrittee, Micrcret Laloratories, West Lafayette,
Ind., 21 g, 14 fig, 0 ref.
Though Hastelloy N is tasxcally a solid-scluticr alloy,
‘thermcrechanical treatments change its mechanical properties
and microstructure, Identifying and characterizing
precipitates involved microscopy, extraction reglication,
-~ x-ray diffraction, and electrcn prolke micrcanalysis.
"Cherical analysis with a microprobe attachment tor the
electron microscope and _electron diffracticn were erployed
to identify 1ndLV1dual,partzcles,‘aqqlome:ates,,and grain-
boundary films on. extraction replicas withcut interfelernce
fror the matrix., The microstructure is characterized by
strxngers of massive primary prec1p1tdtes cf the NildMc3C
type. Exposure between 500 dnd 1000 deg C precigates
_pdrtlcles of the Ni2Mou4C type in the grain bcundaries,
In ajir-melted heats with C.,6% Si, the preciritates are
enriched in siliccn and are not dissolved at high
temperatures but melt and transfcrm tc a ncncartide plase.
In vacuum-melted heats with low silicon contents, cartides
go into solid solution. The only fprecipitates that fcrm in

Accessicn Number F8A6560020 to FEA68Q0029

 
 

Page 78

Category F
Hastellcy N and BRelated 2Alloys

FBA680029 *Continued#
air-melted alloys at as high as 1180 deg C are the |
Ni3(Mo,Cr) 3(C,Si) and Ni2 (Mo ,Cr)4 (C,Si) types. The arnount
and kehavicr are highly silicon dependent; this izpurity
stabilizes the particles. The delta-NiFo intermetallic is
probably responsible for the increased embrittlement at high
annealing temfperatures.

heat treatments + Hastelloy N + precipitaticn + picrcstructcre

FBB6ECO18

Donnelly &G

TUBE PLUGGING IN THE MOLTEN-SALT REACTCHF EXEEFIMENT

FRIFARY HEAT EXCHANGER

O0ak Ridge Naticnal lLaboratory, Tenn.

CRNL-TM-1023 (Feb. 1965) 11 p, 6 fig, U ref.
To reduce the gressure drop through the shell 51de cf the
MSRE primary heat exchanger, it was decided tc recove fcur
of the cuter U-tubes. 1This required sealing the eight tube
stubs produced. A plug design and seal weldirg frocedure
wvere develcged tc assure a high-integrity seal Letween the
molten fuel salt on the shell side and the ccclart salt on
the tube side cf the heat exchanger. The plugs had a
slight interference fit ({0.000C to C.CC(C2 in.,) with the
tubes and were machined for edge-welding. The plug
material was Hastelloy KN, as was the entire heat exchanger.
The tube end was manually velded to the plug with a gas
tungsten—-arc tcrch. The conditions were adjusted to
provide weld metal penetration equivalert tc at least the
thickness of the tube wall. Visual, dye- penetrant, and
radiographic examinations of the uelds cave every '
indicaticn that high-integrity selds had been made that
would successfully isolate the fuel salt frcm tke ccolant
salt during the planned operation of the heat exchanger.

fabrication + Hastelloy N + heat exchangers + MSRE ¢ welding

OTHER CATEGCEIES: HCX

FBB6€6C021

Gilliland G + Venard JT

ELEVATED TEMPERATURE MECHANICAIL FRCEERTIES CF ¥ELLS IN A

Ni-Yo-Cr-Fe ALLCY

Oak Ridge Naticnal Laboratory, Tenn.

ORNL-TH-1341 (Jan. 1966), 35 p, 18 fig, 6 ref.
The ccntents of this TIM appear in Welding J. {(N. Y.) 45,
103-s-11C-s (1966) AC FBB660022.

*Hastelloy K + *welding + heat treat:u:(—:nt‘= + creep +

ductility + ternsile rrcperties

CTHER CATEGCEKIES: FEC

FBB6€C0Z2
Gilliland G ¢ Venard JT
ELEVYATED TEMPERATURE MECHANICAL PRCEERTIES CF WELLS 1IN 2

Accession Number FBA68002S to FBB66CCZ2

C
»)

Page 79

Category F
Hastelloy N and Related 2llcys

FBB66C022 *Ccntinued*
Ni-Fo-Cr-Fe ALLOY - -
Oak Ridge Natiocnal Labcratory, Tenn. :
welding J. (N. Y.) 45, 103~-s (Mar. 1966), & .p, 18 fig, 6 ref,
Tensile tests c¢cn transverse weld samples of Hastelloy N in-
the as-welded and annealed conditions skow a ¢gccd.
comkinaticn. of strength and ductility from 70 to 1800 deg F.
Tensile properties of these ccmpare favcrably with thcse of
the base metal. Stless relieving at 1600 deg F for 2 hr
lovered the tensile yield strength, Creep-rujpture tests at
2200, 1300, and 1500 deg F showed significant improvement
in strength and ductility at 1300 deg F trom stress relief
in hydrogen. In creep-rupture tebavior, bcth as-welded angd
stress-relieved specimens were as yood as the base metal.
the nil-ductility temperature, as deterxined Lty simulated
heat-affected zone therwmal cycle tests was 23((C deq F.
Reascnable recovery of mechanical properties followed a
- simulated welding cycle with a 23C0 deg F raxicurn
temperature. The ccntents of this paper also appear as
CRNL-TM~-1341, AC-FEE660022.
*Hastelloy N + *welding + heat treatments + creep +
ductility + tensile properties
OTHER CATEGORIES: FBC

FBR690040
McCoy HE + Cancnico DA '
PREIRRALCTATICN ANC POSTIRRAEIATION MECHANICAL PROPERTIES
OF HASTELLOY N WELDS
Cak Ridge Rational Laboratory, Tenn.
ORNL~-TM-2483 (Mar. 1969), 43 g, 20 fig, 16 ref.
Welds wvere made by the TIG proccess in several heats of
" Hastelloy N. The mechanical fproperties of transverse weld
samples and the base metal were ccmpared in tencsile tests
over the range cf 75 to 1600 deg F and in creep tests at
1200 deqg F. The as-fatricated welds extibited lcwer
fracture strains than the base metal under all test
~ conditions, but the properties of the welds wuere 1mprcved
- markedly by post-weld heat treatments. The pcstirradiation
tensile and creep properties of the welds and base mretal at
elevated temgeratures vere about the same, althougyh the
properties were widely different tefore irradiaticn.
Mechanical properties of all specimens tested are takulated
~in an Appendix. The sameimaterlal withcut the apiendix is
~published in welding J.f(N. Y.) 48, 203-3-211-5 (1969),
AC ~-FBE690041,
*flastelloy N + *yelding + c:eep + ductlllty + *1rrad1atlon +

tensile properties

CTHER CATEGCFIES: FEE

FBB69C0041T
McCoy HE + Canonico CA

Accessicn Number FBB6600ZZ to FEE6SCCUI

 
Page 80

Category F
Hastelloy N and Related Allcys

FBB6900 41 *Continued*
PREIREADIATICN AND POSTIRRADIAT ION BECHANICAL PROPERTIES
OF HASTELLOY N WELDS

Oak Ridge National Laboratory, Tenn. -

Welding J., (N.Y.) 48, 203-s (May 19€%), 9 p, 18 fig, 16 ref.

- This fpaper presents the same material as CFNL-TM-2483
(FBE690040) except that it does nct include the appendix
vhich lists the rrcperties of all specimens.

*Hastelloy N + *yelding ¢+ creep + ductzllty ¢+ ¥jirradiation ¢

tensile prcrerties :

GTHEH CATEGCEIES' FEE

FBB70C028

McCoy HE + Gunkel RW + Slauqhter GY

TENSILE PROPERTIES OF HASTELLCY K WELDEL AFTEE IEEAIIATION

Cak Ridge National Labkoratory, Tenn.

ORNL~TM-2858 (Apr. 1970), 24 g, 7 tig, 8 ref.
Fusion welds affecting 75% of the cress secticn were
made in small tensile samples 0., 125 in. in diameter)
~of Hastelloy N irradiated to thermal tluences up to
9.4 x 10 (20th) neutrons/sq-cm, All cf the unirradiated
samples and 67% of the irradiated samples were
satisfactorily wvelded by a specialized technicue. Suiface
contamination is suspected to cause the unsuccessful welds
in the irradiated samples. The welded irradiated samgles
generally had as gccd temsile properties at 25 and 65C deg C
as the irradiated Lkase metal. The weld metal defcrmed
appreciably at 650 deg C and mwade a significant contribution
to the overall fracture strain. The fracture location in
the irradiated samples tested at €50 dec C shifted frcm the
weld metal to the base metal following the post-veld anneal
of 8 hr at 870 deqg C, Porosity near the fusicr line of the
irradiated sawmgples frobably resulted frcm helium Lbutbles,
but this did not seem to atfect the locaticn c¢f the
fracture,

welding + irradiation + Hastelloy N e gductility +

tensile prcgperties

FBB700031

Cepolina AG + Cancnicc DA

THE MFEASUEKEMENT OF RESIDURL STRESSES

Oak Ridge Naticnal Laboratory, Tenn.

ORNL-THM-3113 (Cct. 1970) 32 p, 12 fiqg, 16 ref.
A mcdification of the Sachs 'boring-out' methcd for
determining residual stresses perrits deterpiraticn of
the distributicn of stresses and their levels over extremely
short increments of distances, This technique was used for
measuring residual stresses in gas tungsten-arc welds made
in Hastelloy N. Circular welds 6 in, ir diarecter were
simultanecusly depcsited on both flat faces of a 1/2-in,-
thick plate, 12 in. in diameter. The raximur tangential
residval stress was fcund to be about =(,0CC psi and was

Accessicn Number F336900u1 to FEE7000231
*;

Page 81

Category F
Hastellcy N and Related Alloys

FBB700031 Continuedx*

not particularly affected by either the shielding gas or
heat input. Stress relieving at 160C deqg P fcr 4,5 hr
proved to be the c¢rtimum heat treatment and reduced tte
tangential residual stress to akcut 5C0C psi. Lcwering of
the maxigum residual stiess to about 10,000 psi was
achieved at 1400 deq F after € br; however, lcwer
temperatures even fcr times as long as 100 hr only reduced
the maximum residual stress Ly aktcut 2z5%.

Hastelloy N + stress + heat treatments + testing

FBC590001

Carlscn RG

FATIGUE STULIES OF INOR-8

Battelle Memorial Institute, Columbus, Chio -

BMI-1354 (June 1959), 16 p, 8 fig, 1 ref.
The tepmperature and freguency dependence of fatigue
properties of Hastelloy N were deterwmined by rctating-
beap fatique tests. Stress-lifetime data were oktained
for temperatures of 1100, 1300, and 15CC F, ard cyclic
frequencies of 100, 600, and 3000 cpm. The fatique
strength decreased with increasing terxperature. No
aprreciable frequency effect uas found up to 1300 £,
At 1500 F, the fatigque strengths ¢f srecimens tested
600 and 3000 cpm were egual, while the fatigue strength.
at 100 cpm was substantially lower. A critical frequency
is associated with each temperature, abcve which frequency
has no effect, but belcw shich fatigue strength decresses
with decreasing frequency. Fatique strength was higher
fcr finer grained material.

Hastelloy N + microstructure + fatigue

FBC61C001

Swindeman F¥W

THE MECHANICAL PROPERTTES OF INOR 8

Cak Ridge .National Laboratory, Tenn., -

ORNL-4780 (Jan. 1961), 76 p, 45 fiq, 20 ref.
Tensile, creep, and relaxation tests were ;erfcrued cr
INOR-8, (This is the-alloy later called Hastelloy N.) The
mechanical properties;are,summarized and discussed in
relation tc the composition, micrestructure, and :
environment. The results indicate that the mipimum strength

~ profperties of INUR-8 are sufricient to permit the use of °*

workable design stresses up to 130 deg F, althcugh certain
areas are pointed cut uherp addltlonal informaticn 1s
desirable.

Hastellcy N + creep + ductlllty + heat treatments +

tensile properties + alloy compositicn

FBC64CO017
McCoy HE

Accessicn Number FBB700031 to FEC640C17

 
 

 

Page 82

Category F
Hastellcy N and Related RARlloys

FBC640017 *Continued*
INFLUENCE OF SEVERAL METALLURGICAL VAEIRABLES (N THE
TENSILE FFOPERTIES OF HASTELLOY N

Oak Ridge Naticnal Laboratory, Tenn, ' |

ORNL-3661 {Aug. 1964), 63 p, 35 fig, 10 ref.
The tensile properties ¢f Hastelloy N were measured after
various heat treatments. One vacuum-kelted ard fcur air-
welted heats were studied. The vacuum-wmelted material
exhibited good ductility after all heat treatrerts,
Annealing the air-melted material to temperatures in e€xcess
of 2150 deg F significantly reduced tite sinimur fractrire
strain. Holding at about 1600 deqg F for an extended period
recovered the fracture ductility. Aginc in tke 110C tc
120C deg F range material that had been previocusly annealed
at 2150 deg F significantly reduced tke ductility. These
changes in ductility cccurred with very small changes in
tensile strength., These effects can ke exrlained in terms
of the fcrmaticn of a brittle grain boundary layer alcng
vhich a crack can propagate easily at elevated terperatures.
Interrupting the ccrtinuity of this layer by overaging or
cold working recovers good fracture ductility. 1Tke
fcrration of this layer is associated with the fresence of
trace alloying elements.

Hastelloy N ¢+ heat treateents + ductility + metallography +

t9n51le properties

FBC65C017

Venard JT

TENSILE AND CREEP PROPERTIES CF INCE-8 FOR THE HCLTEN-SALT
REACTCR EXFERIMENT

Oak Ridge Naticnal Labhoratory, 1Ienn.

CRNL-TM-1017 (Feb. 1965), 22 p, 19 fig, 6 retf.

Tensile and creep-rurture testing has been carried out on

three heats of Hastelloy N selected frcu thcse used fcr the

MSRE. The primary aim was to collect strength information

representative of the constructicn material ard tc corgare

the data on these ccmmercial heats with that from early

experimental heats., The data repcrted are ultimate tensile

.strength, 0.2% cff-set yield strength, elongation, angd

reduction in area from room temperature to 982 deg C .

(1800 deg F). Creeg-tupture behavior was investigated at

493, 7C4, and 816 deg C (1100, 1300, and 1500 deg F). 1In

general, the commercial MSRE ccnstructicn material shcus

greater strength and ductility than did earlier heats,

Additional confidence in the MSRE design strergth valves is

tbus in crder.

*Hastelloy N + creep + ductility + MSRE + tensile rrcierties +

alloy ccmfpositiorn

FBD6900 36"
Koger JW + Litezan AP

Accession Number FBCEU0017 to FBDOHS(CO36
 

 

*}

Page 83

Category F
Hastelloy N and Related Allcys

FBD69C036 *Ccntinued*

'COMPATIBILITY CF HASTELLOY N ANL CROLOY 9M WI1TH

NaBFU4-NaF-KBF4 (90-4-6 mole %) FILUCFCEQORATE SALT

Cak Ridge National Laboratory, Tenn. |

ORNL-TM-249C (Apr. 1969), 41 ¢, 20 fig, 15 ref.
The compatibility of relatively iwpure (greater than
3000 ppw ippurities) NaBF4~NaF-KEF4 (90-4-€¢ mole %)
tested with Hastelloy N and Crcloy 9¥ was tested in
natural circulaticn loops at a maximum temperatuie
of €CE deg C with a temrerature difference of 145 degqg C,
The Croloy 9M-loop was completely plugged after 144C br
and the Hastelloy N loof was three-guarters plugged
after 8760 hr (one year)., All major alloying elements
mass transferred as the result of nonselective attack
by virtue of the initial oxygen and wvater ccntamiraticn
of the salt. Saturaticn concentrations of 700 ppn
Fe and 470 ppm Cr were determined ior tke flucicbcrate
salt at 460 deg C. Initially, solulle metal flucride
compounds formed in the hot leqy, The reverse reaction
in the cold leg causes the metal to depcsit and tc
diffuse into the ccld leg. This continues until an
equilibrium concentration of cne cr more metal fluorides
is reached in the salt at the cold-leqg temperature
and these compounds start depositing cr the eqguilibrium
constant of the reaction changes so much with temperature
that the pure metal is deposited,

*iron allcys + *Hastellcy N + compatibility + *corrosion +

*fluoroborates + *impurities + mass transfer + sclukility +

coolants + thermal ccnvecticn + molten salts

CTHER CATEGCFIES: GEX

FBE6EC01S
Martin WR + Weir JR : '
EFFECT OF ELEVATED TEMPERATURE IRRADIATION CN THE STREINGTE ANL

DUCTILITY CF THE NICKEL-EASF ALLOY HASTELLOY N
Oak Ridge Naticnal Labcratory, Tenn,

 0RNL-TH -1005 (Feb., 1965), 17 p,-7 fig, 17 ref.

The ccntents of this. 14 are the same as an artlcle with the
same tltle in the Aprxl 196‘ Nuclear Appl;catxcn which
S€e.,
*Hastelloy N + ductlllty * *ten51le propertles + 1rtadzat10r
OTHER CATEGORIES: FBC. o .

' FBE650016

Martin WR + ¥Weir JR

EFFECT OF EILEVATED TEHPEBATURE IRRABIAIION ON IHE ST?ENGIH AND

DUCTIILITY OF THE NICKEL-BASE ALLCY HASTELICY N
Cak BRidge Natiomal Laboratory, Tenn, -
Nucl. Appl. 1: 160 (1965) 8 §, 6 fig. 20 ref. '
The tensile properties of Hastellcy N have been deterinined
after irradiaticn at 600 deg - to 7 x 10(20th) n/sq-crk (E
greater than 1 MeV) and 9 x 10 (2Cth) n/sqgq-cm (thermal). The

Accessicn Number FBD690C3E to FBEGZIC0C16

 
 

 

Page 84

Category F
Hastelloy N and Related Allcys

FBE650016 *Ccntinued*
strength and ductility wvere determ1ned fcr the range 20 to
900 deq C. 1The stress-strain relationship is not affected
by irradiation at 700 deg C. Cuctility, as measuvred Ly the
true unifcrm and fracture strains, is reduced for
deformation temperatures of 500 deg C and abcve, The lass
in ductility results in a reduction in the true tensile
strength, especially in intecrcranular failure, suchkh as low
strain rates and elevated temperature. Fost irradiation
annealing does not improve ductility. 1bese data are
ccmfpatible with heliur generation from the (n, alpha)
reaction of boron as the cause of low ductility. (Also
published as ORNL-TN-1005.)

*Hastelloy N + ductility + *ten511c Frcperties + 1Irad1at10t

OTHER CATEGORIES: FBC

FBE660D19
Martin WR + Weir JR
PCSTIFRADIATICN CREEP ANLC STRESS RUPTURE OF HASTELLCY N
Oak Ridge Naticnal Laboratcry, Tenn, '
ORNL-THE-1515 (June 1966) 31 p, 12 fiqg, 15 ref.
The ccontents of this TIM are the same as an article
vith the same title in the March 1967 Nuclear A;;llcatxcn
vhichk see.
*Hastelloy N + *creep + ductility + irradiaticn +
microstructure
CTHER CATEGCFIES: FEC

FBE67C029

Martin ¥R + Weir JR ‘

POSTIRRADIATION CREEP AND STRESS RUETURE OF HRSTELLCY N

Cak Ridge National Latoratory, Tenn.

Nucl. Appl., 3, 167 (Mar. 1967), 11 p, 10 fig, 17 ref.
The creep ductilities of irradiated Hastelicy N at
650 deq C have been determined at several neutrcn
exposures, gflevated-temperature irradiation emtrittlement
Jreatly reduces the stress-rupture strenqtkt acs seasured
in pcstirradiation uniaxial stress tests. The reduction
in ductility to values as low as C.4% is due tc an
irradiation effect related to the process of intergranular
fracture., TIntergranular cracks, once fcrmed, prcgaga te
with greater ease in the irradiated alloy as compared
with a sample exposed to a lesser radiaticn expcsure.
(Also published as CRNL-THM-151%E,)

*Hastelloy N + *creep + ductility + irradiaticn +

microstructure

OTHER CATEGORIES: FBC

FBE670030
McCoy HE ¢ Weir JR
IN- AKD EX-REACTGCR STRESS-RUPTURE PROPERTII-.c OF HASIELLOY N TUBING

Oak Ridge Naticnal Laboratcry, Tenn.

Accession Number FBEG650016 to FEE670030

1.
.}

»

Page

Category F
Hastellcy N and Related Alloys

FBE670030 *Continued*

ORNL-TM-1906 (Sept. 196p), 27 p, W fig, 28 ref.
The stress-rupture properties of two heats cf Hastellcy N
tubing have been determined at 7€C deg C in the irradiated
and unirradiated conditions, Irradiation reduced the
rupture life and the rupture strain kut gprcduced rno
detectable effects cn the creep rate, Small variations in
behavior of tutular specimens tested during irradiaticm and
srall rod specimens tested after irradiation are expléeined
on the basis of differences in stress states and sizes cf
test sections., The effecte of irradiation are rationalized
on the basis of the kehavior ¢f bheliur which is fcrmed in
the metal as a result of the reaction ot boron-10 with
thermal neutrons, (Also putlished with scre ccndensa ticn in
Nuclear Applications ([MSRIS Accession FEE6EC(C25),

Hastelloy ¥ + creep + ductility + 1rrad1atlcn

OTHER CATEGORIES: FBC ,

FBE670031

McCoy HE

AN EVALUATICN CF THE HOLTEN SALT REACTOR EXPERIMENT

HASTELLOY N SURVEILLANCE SPECIMENS -- FIEKST GERCUE

Cak Ridge National Laboratory,  Tenn.

ORNL-TM-1¢97 (Nov. 1967), 57 p, 35 fig, 12 ref.
Gives test results on the effect c¢f varicus variables
(temperature, strain raté, prestrain, etc.) on the
tensile ductility of irradiated and unirradiated Hactelloy
N. Specinmens removed frcm the ¥SEE after 7823 MWh had Leen
at 645 deq C for 4800 and accumulated 1.3 x 10(20th)
neutrcns/sqgq-cm (thermal). The high-temperature ductility
vas reduced sirilarly tc that observed for the same
materials in the ORR in heliux»., No ccrrcsicn was cbserved,
but a 1 to 2 mil carbon-rlch layet was noted where specimens
touched graphite .

*MSRE + *surveillance + *Hactelloy N + *compdtlblllty +

‘fluorides + creep + corrosion + irradiation +
=ten511e ptcpert1es + tensile ;ropentles

FBEOGOOZS ,
McCoy HE + Weir JR

_STRESS-RUPTURE PBOPEBTIES OF IRRADIATED "AND UNIRRADIA?ED HASTELLbY

TUBES
Cak Ridge National Laboratory, Tenn,

~Nucl. Appl. 4, 96 (Feb. 1968), 9 p, &6 fiyg, 2& ref.

The stress-rupture properties of two heats cf Hastellcy N
tubing have been deterpined at 7€C deg C in the irradiated
and unirradiated reacticn of soron-10 with therwmal neutrons.

(Feported in more detail in ORNL-TM-19Cé6.) '
Hastelloy N + creep + ductility + irradiation
CTHER CATEGCRIES: FEC

FBE68C0Z26
McCoy HE
EFFECTS OF IRRADIATION ON THE MECHANICAL ERCFERTIES CF TWO VACUUM-

AccesSion Numker FBE670030 tc FEEBBOOZG

85

 
Page 86

Category F
Hastellcy N and Related Alloys

FBE680026 *Continued*
MELTED HEATS OF HASTELLCY K

Cak Ridge National Laboratory, Tenn.

ORNL-TN-2043 (Jan. 1968) , 43 ¢, 24 fig, 18 ref.
The mechanical behavior of twc vacuum-melted heats cf
"Hastelloy N was tested at 650 and 760 deg ¢. The material
vas subjected to several therral-wmechanical treatments and
then irradiated at 650 and 760 deg C to a thermal dose of
2.3 x 10{20th) neutrons/sq-cm., Tke resclts ate ccmpaxed
with those for unirradiated specimens that were given a
‘similar thermal treatment., The varicus thermal-recharical
treatrents bhad scme relatively small effects on the tensile
properties of unirradiated material, ftut the creeg
progerties were very similar. The primary effects of
irradiation were reductions in the creeg- rupture life and
the rupture ductility in both creep and tensile tests,
These observations are explained cn the basis cf heliim
producticn in the metal by the boron-10 (n, alpha)
transmutation. |

*Hastelloy N + creep + ductility + heat treatments +

tensile properties + irradiation

OTHER CATEGORIES: FBC

FBE690034

McCoy HE

AN EVALUATICN CF THE MOLT EN-SALT REACTOR EXPERIMENT .

HASTELLOY N SORVEILLANCF SEECIMENS -- SECCNL GECUE

Cak Ridge National Labtoratory, Tenn.

ORNL-TM-2359 (Feb. 1969), 69 g, 45 fig, 22 ref.
Two rods of standard Hastelloy N from tle surveillance
pcsiticn cutside the core vessel were exposed to nitrogen
plus 2 to 5% oxygen for 11,000 hr. The allcy was
comgpatible with this environment, shovwing only superficial
oxidation and no nitriding. These sarples were expcsed to
a thermal fluence cf 1,3 x 10(19th) neutrons/sq-cm, and both
tensile and creep tests showed significant changes in
mechanical prcperties, rarticularly the strain at fracture.
These changes are in good agreement witl thcse tcr material
irradiated in heliur in the CRE. Tvwo rods of modifiec
Hastelloy N containing swall additicns cf titapiur and
zirccniur frem the ccre with a thermal fluence of 4,1 x
10(20th) neutrons/sq-cm, showed slightly irprcved
postirradiation mechanical propertles and acceptatle
cCorrosion resistance.

*MSRE + *surveillance + *Hastelloy M + #*podified Hastelloy K +

*compatibility + nitrogen + oxygen & fluorides + creep +

rupture + ccrrosion ¢+ irradiation + irradiation +

tensile properties + mlcrostructure

OTHER CAIEGORIES' FCE

FBE690044
McCoy HE

Accession Numker FBE680C26 to FBE630044
-

Page 87

Category F
Hastelloy N and Related Allicys

FBE69COQH *Ccntinued*
VARIATICN CF THE MECHANICAL PROPERTIES OF IRRADIAEED
HASTELLOY N KITH STIRAIN RATE

Cak Ridge National Laboratory, Tenn.

J. Nucl. Mater. 31, 67 {(May 1969), 19 p, 12 fig, 44 ref.
The postirradiation mechanical prcperties cf several heats
of Hastelloy N, both vacuum- and air-melted, have been
measured after exposure to thermal fluences cf 2 to
6 x 10(2Cth) neutrcns/sq-cm, At strain rates normally
encountered in tensile tests, the fracture strair is cuite
sencsitive tc strain rate in the range of 5C0 tc €50 deg C.
At €5C degq C a minimum fracture strain wWwas okserved at a
strain rate of approximately C.1% hour; the strain
increased rapidly with increasing strain rate and increased
gradually with decreasing strain rate. Althcugh the

 fracture strains at high strain rates differed significantly

for test temperatures of €50 and 760 dec C, the strairs were
the same under creef conditions, A titanium-modified alloy
showed improved resistance tc irradiaticn damage.,
Qualitative explanations are jiven for each of these
observations..

irradiation + Hastelloy N + mcdifiel] Hastelloy N +

ductility + creep + tensile propertles + allcy ccmpcsition

OTHER CATEGORIES: FBC -

FBE700027

McCoy HE

AN EVALUATICN CF THE MOLTEN-SALT REACTOR EXPERIMENT HACTELLCY N
SURVEILLANCE SPECIMENS -~ THIRD GEFCUE

Cak Kidge National Laboratory, Tenn.

ORNL-TM~-2€47 {Jan. 1970), 88 p, 56 fig, 8 ref.

We examined the third group cf hastellcy N surveilla rce
samples froms the MSKE. Standard Hastelloy N was exposed in
the core to a thermal of 9,4 x 10420th) neutrcrsysg-cnr over
15,269 hr at 650 deg C and outside the reactor vessel to 2.6
X 10(19th) neutronsysq-cm over 20,789 hr at 650 deg C. The

, fctmer sabples were expcced to the fuel salt and the latter
to nitrogen rlus 2 to 5% oxygen., The material seemed quite
compatible with hoth environments, Pcstirradiaticn tests

'~ showed that the fracture strain uas reduced at 25 deyg C and

~above 500 deqg C. The reducticn at 2n deqg C is likely due to
cartide precipitaticn and that above £0C deg C is due to
helium from,borqn-10(n;alpha}. Accurulated tesults allcw us

- to fcllow changes in fracture strain witk therral fluence
fror 1.3 x 10(19th) to 9.4 x 10(20th) neutrons/sg-cm. Two
heats of modified Hastelloy N were irraciated in ‘the ccre to

. a thermal tluence cf 5.3 x 10 (20th) neutrons/sy-cm over 9769

- hr at 650 deg C. The postirrediation prcperties were better
thap those c¢f standard Hastelloy W,

*MSRE 4+ *surveillance + *FKastelloy N + *modified Hastelloy X +
*compatibility ¢+ nitrcqgen + oxygen + fluorides + creer +
corrosion + irradiation + tensile properties ¢ picrcstruwcture

Accessicn Number FBE6Y0044 to FEE70CCZ7

 
 

 

Page 88

Category F
Hastellcy N and Related Alloys

FBE700027 *Continued#*
OTHER CATEGORIES: FCE

FBE710017

McCoy HE + Gehlbach RE

INFLUENCE CF IEKRADIATION TEHPERATURE ON THE CBEEP—RUPTURE

PROPERTIES OF HASTELLOY K

Oak Ridge National Laboratory, Tenn. :

Nucl. Technol. 11, 45 (May 1971, 1€ p, 17 fic, 15 ref.
The variaticn c¢f the rostirradiation creep-rupture
properties with irradiation temperature has beer determined
for air- and vacuum-melted Hastelloy k. The air-melted
material vas high in silicon and formed a stable carbide of
the ME€C type. 1The froperties of this material wvere not
dependent upon the irradiation temperature cver the range
studied. The vacuur-melted alloys formed a M2C-type
carbide whose size and morpholcgy depended markedly ujcn
the irradiation terperature., When the carbides were finely
dispersed by irradiation at atout 65C deqg C, the
postirradiation prcierties were equivalent to those of the
air-melted material, Irradiation at abcut 7€6(C deg C

_gresulted in coarser disgersions of the M2C carbide ané

inferior postirradiation properties.

creep + ductility + Hastellcy N + irradiation +

thermal effects + alloy compositicn ¢ ricrcstructure +

carbides '

CTH¥YR CATEGCHRIES: FCE

FBE71C018

McCoy HE

AN EVALUATION OF THE MOLTEKR-SAIT REACTCEF EXEEFIMENT

HASTELILCY N SURVEILLANCE SPECIMENS -—- FOURTH GROUP

Oak Ridge Natiocnal Laboratory, Tenn.

CRNL-TM-3036 (March 1971), 91 p, 67 £fiq, 14 ref,
Two heats of standard Hastelloy Kk were removed from the
core of the MSRE after ¢4,533 hr at 650 deg C, a thereal
fluence of 1.5 x 10(21) neutrons per square centimeter,
and a fast fluence ({greater than 50 keV) cf 1.1 x 10(21)
neutrcns per square centimeter. The mechanical properties
have systematically deteriorated with increasing fluence.,
However, the change in fproperties is due tc thke heliur
produced by the B-1C{n,alfha) 1i-7 transmutation and can
be reduced by changes in chemical comfpositicn. Scme heats
with modified ccompcsiticn have been exposed to the core of
the ¥SRE and shov improved resistance tc irradiaticn. 1Ine
~corrcsion of the Hastelloy N has been largely due to the
selective removal of chromium, The rates cf temcval are
much as rredicted from the measured diffusion rate of
chromium. Other superficial structure todificaticnes have
been cbserved, but they likely result from carbide
precipitation along slip tands that were fcrred during

Accessicn Number FBE700027 to FEE7 100 1€
 

o

&k

=

Page 89

Category F
Hastellcy N and Related Alloys

FBE710018 *Continued*
machining. - -
rodified Hastelloy N + MSRE + surve111ance + corresicr +
creep + ductility + fluorides + Hastelloy N + irradiation +
alloy composition + molten salts + mpicrcstructure +
tensile prcrerties o
CTHER CATEGCEIES: FEC + IBL + ICE

FBX64C015

Taboada A ' |

METALLURGICAL DEVELOPMENTS {IN HSRP EECGR. REET., 7/31/64)

Cak Ridge National Laboratory, ‘Tenn.'

ORNL~- 3708 (Nov. 1964), pfp. 33C-372, 27 fig, & ref.
Metallurgical develcpments in support of the FKSFE show that
Hastelloy N is satisfactory. Physical, tensile, creej, and
fatique properties are given, Compatibility is excellent
with circulating molten fluorides, graphite ir gclten
fluorides, and air to 1800 deg F. Fabrication is described,
including a complex heat exchanger, Anrealing is needed
after cold work, and attention to composition is required
to ensure weldablllty. Irradiaticn causes scre lcss cf
tensile strength and ductility. Inconel-clad gadolinia-
alumina control rod elements were fakricated.

* Hastelloy N + brazing + ceramics + compatibility +

control rods + corrosion + creep + density ¢ ductility +

electrical conductivity + elasticity ¢+ erosion +

expansion + fatrication + fluorides ¢+ graphite +

heat exchangers + heat treatments + inconels + irradiation +

joints + loop + machining + melting + casting +

metallography + MSRE + oxidation + physical properties +¢

procurement + progress report + rare€ earths + revieus +

specific heat + specificaticns + surveillance ¢+ testing +

thermal conductivity + welding + micrcstructure +

tensile prcperties + fatxgue + molten =alts

CTHER CATEGCEIES: ACE - Lo :

FCCGQCOQB ,

Sessions CE ¢+ Lundy TS : ’ -
DIFFUSION OF TITANIUM IN BODIFIED HASTEILOY N
cak Ridge National Laboratory, Tenn, |

ORNL-!H ~2392 (Jan, 1969), 24 f, 7 fig, 13 ref.-

piffusion coefficients of titanius-44 irv tltdtlLfl-mOdlfled

fHastelloy N were determined over the range ECC to
125C deg C by serial sectioning by latheing cr grindirg and

- counting by gamma—cpectrcccdpy. ‘The data were fitted tc

D = {15.3 plus or minus 2.2) exp(-73,000 plus or minus

3300/BT) sq- cm/sec. " Results were used tc predict the
- maxicum loss by diffusicn of titanium from the alloy in a

‘typical molten-salt breeder reactcr at 700 deg C.

Expected increases c¢f titanium in the mclten salt are no

more than 5 to 10 ppm for two years cf cperaticr, based on

Accessicn Number FBE710018 to FCC6Y004¢€

 
 

 

Page 90

Category F
Hastellcy N and Related 2alloys

FCC690048 *Continuedx o ' :
a sirplified diffusion model., Also published as J. Nucl.
Mater. 31, 316-22 1969). : '

diffusion + titanium + modified Hastelloy N

FCC69C049

Sessions CE + lundy TS

DIFFUSION OF TITANIUOM IN MODIFIED HASTEILOY N

Cak Ridge Kational Lakoratory, Tenn.

Je Nucl. Mater. 31, 316 (July 1969), 7 p, 5 f£ig, 13 ref,
Diffusion coefficients of titanius-44 in titarium-
modified Hastellcy K were determined over the range of
800 to 1250 deg C Lty serial secticning ty lathkeing cr
grinding and ccunting by gamma-spectroscopy. The date were
fitted to © = (15.3 plus or minus 2,2) exp(—-73,CCC flis or
minus 33C0/R1T) sg-cm/sec. Results were used to predict the
maximum loss by diffusion of titaniur frcom the alloy in a
typical rolten-salt breeder reactcr at 700 deg C. Exjyected
increases of titanium in the molten salt are no more than
S to 10 ppm for two years of cperaticn, based or a
simplified diffusicn model. Also published as CENL-TE-2392.

diffusion + titanium + modified Hastelloy N

ECC?OCOQO

Sessions CE

INFLUENCE OF TITANIUM ON THE HIGH—TEPPEEATUEE DEFCEFMATICN

AND FBACTURE EEHAVIOR OF SOME NICKEL BASED ALLOY° (1HESIS)

Oak Ridge Naticnal Labcratory, Tenn.

CRNL-4561 (July 1970), 189 p, 44 fig, EFf ref.
Adding 0.5% titaniumw to nickel-1z% molytdenum-7%
chromium-0.06% carbon decreased thke creep rate, ifncreased
stress-rupture life and ductility. Increasing carbcn from
0.0C3 to C.3% increased rupture life and decreased creep
rate four orders of magnitude and increased ductility
threefold for varicus stresses. Fracture at 650 deg ¢
changed from intergranular at low cartcr tc rixed trars-
and intergranular at high carbon for similar heat
treatments. Increasing titanium up tc 1.2x favcred
fcreation of an MC-type carbide during aging at €50 and
760 deq C rather than the M2C that fcrrs at lcwer
titanium, A heavy distribution of ¥C at grain toundaries

. resulted in superior ductility in both creep and tensile

tests, presumakly ty reducing gqrain-ktcurdary =hearirg and
lipiting grcwth of cracks., #hen alloys with 1,2%
titanjum vere solution annealed at 1260 deq C ard aged, MC
cartides precipitated opn dislocations, causing qro#th of
stacking faults, which increased strength but impaired
ductility. Small titanium additicns imfproved ductxllty of
pure nickel at 600 deg C.

*nodified Hastelloy X + *creep + nickel ¢ *ductility +

*heat treatrpents + carbides + precipitation + microstructure +

Accession Numler FCC690048 to PCC700040

 

 
L]

Page 91

Category F
Hastelloy N and Related Allcys

FCC70C040 *Continued*
*alloy composition
OTHER CATEGORIES: FCa

FCC700044
Sessicns CE
INFLUENCE CF TITANIUM ON THE EIGH-TEMPEBATURE DEFORUA?ION
AND FRACTURE BEHAVIOR OF SCFE NICKEL-EASEL 2LLCYS
Cak Ridge National Laboratory, Tenn, | '
Scripta Met. 4, 795 (Oct. 1970), 4 p, 1 ref.
This publication is an atstract of repcrt ORNI-4561 (MSRIS
"Accession FCC700040).
*modified Hastelloy N + *creep + nickel + *ductlllty +
*heat treatments + carbides ¢ precipitation +
microstructure + *allcy ccmposition

FCC710010
Sessions CE + Stansbury EE
THERMAL STABILITY OF TITANTUM- HODIFIED FASTELLOY N AT
65C and 76C DEG C |
Cak Ridge National Laboratory, Tenn.
ORNL~-TN-3321 {July 1971, 43 p, 17 ftig, 19 ref.
. The influence of small titaniur additicrs ¢n the thersal
stability cf Ni-12% Mo-7% Cr-0,07% C was investigated.
The mechanical properties at €50 deg C (tensile tests at
0.002/min strain rate and creep tests at 40,CCC gsi stress)
vere measured for four heats cf this allcy with titanium
contents fror 0.15 tc 1.,2%., Solution annealing
temperatures were 1177 or 126C deg C, ard subscecuent
precipitation heat treatments were conducted at 650 ana
760 deg C. Titanium increases the stability ¢f a ccmjlex
MC-tyre carbide. At low titanium levels the FKC carktide is
stable at 650 deg € but is unstakle at 760 deg C, whele an
- M2C-type carbide is fpreciyitated, resulting in infericr
, properties.- For the higher titanium ccrcentraticcs the MC
~carbide is stable cr aging.at 760 deqg C and results in
excellent properties after a solutlon arneal at 1177 deg C.
~ However, high-titanium alloys are significantly less
ductile it they are solution annealed at 126C deg C atd
- aged at either 650 cr 760 deg C. The heat with the lowest
. carbon content (0.04% C) was most resistant tc prcperty
changes cn aging up to 10,000 hr at kcti €5C and 76C deg C.
*podified Hastelloy N +,*develcpment + %alloy composxtlon +
*heat treatments +. aging + .creep + ductility + :
microstructure + tensile: ;ropertles + carbldes +
thermal effects , e ,

FCD71C01€

Evans {ITI) EE + Koger Jw + DeVan JH

CORROSION IN POLYTHEBHAL LOCP SYSTEMS I1. A SCLIC-STITE
DIFFUSICN MECHANISHM WITE ANL WITHOUT LIQUID FILM EFFECTIS

Accessicn Number FCC700040 to’ FCL710016

 
 

 

Page 92

Category F
Hastellcy N and Related Alloys

FCD710016 *Continued*

Oak Ridge National Labcratory, Tenn. '

CRNL-4575, Vol. II (June 1971) 74 p, 16 fig, 49 retf.
The corresion of alloys exposed to nonisothersal
circulating liquids is important in systems with liguid
coolants or coclant-fuel combinations., Mathematical
descriptions vere developed to explain transpcrt cf |
constituents of allcys. This report specializes to cases

in which solid-state diffusion in the allcy dcmirates the

corrcsion. Equations are derived for hoth transient and
steady-state cases; transients are negligible.
Applicability is demrcnstrated by comparison of fpredicted
values with experimental results for twc systets, The
first involves hot-tc-ccld-zone transfer of nickel in
Inconel 600 pumped loops circulating sodiux. Actual
corresion is much higher than predicted; this suggests that
the true corrosion reaction overrides a slcw diffusior
process., The second system is transfer of chromium in
Hastelloy N thermal convection locps with zclter salt.,
Three examples are considered; (1) corrosion at all pcints,
transfer to salt only; (2) hot-tc-cold-zone trarsfer; and
{3) ccld-tc-hot-zcne transfer. - Early Cr-51 tracer
experiments (example 1) suygest that solid-state diffusion
applies to certain molten-salt systenms.

*corrcsion + *gpodels + *diffusion + analysis ¢

computer codes ¢+ coolant loops ¢ fluorides + Hastellcy N +

inconels + liquid metals + locp + mass transfer +

mathematics 4+ sodium + models )

OTHER CATEGORIES: GCX

FCE690043
McCoy HE + Weir JR
DEVELCPMENT CF A TITANIUM-MOLIFIED EASTELLOY N WITH
IMPROVED RESISTANCE TO FRADIATICN CANMAGE
Cak Ridge KNational Laboratory, Tenn.
Irradiaticon Effects in Structueral Alloys for Thermal and
Fast Feactors, ASTM STP 457, Am, Soc. for Testing ard
Materials, 196G, p. 290, 22 p, 11 fiq, 19 retf.
The effects of neutron irradiaticn on tke bkiglk-temperature
mechanical rrcperties of Hastelloy N are generally thet
the creep-rupture life and ductility are reduced. 1he
ductility is a strorg function of the strain rate and shovs
a minimum at a minimum creep rate of atcut 0.1 %,/hour.
The resistance to radiation damage can ke enhanced greatly
by adding titanium., The postirradiaticn creef~rugture
ductility and strength rise sharply as the titanium ccntent
is increased above 0.3X. Postirradiatich creep—-ruptuire
- tests at 650 C cn specimens irradiated to a thermal fluence
of 5 x 10(20th) neutrons/sq-cy» indicate that a duct111ty
minimum still exists as a function of strain rate,
However, the minimum strain i3 3 to 5% as ccmpared with

Accessicn Number FCD71001€ to FCE6SCCu3
 

L

-

Page 93

Category F
Hastelloy N and Related Allcys

FCE69C043 ¥*Continued* -
0.5% for the standard alloy. . In-reactcr creefp-rLUfpture
tests indicated the same improved properties.
irradiation + ductility + creep ¢ Hastelloy N +
modified Hastelloy N + allcy composition =

FCE71C004

¥cCoy HE

INFLUENCE OF TITANIUH, ZIRCONIUF, AND HAFNIUMN ADEITICLS ON

THE RESISTANCE CF MOLIFIEL EASTELLOY N TO IRRADATION
DAMAGE AT HIGH TEMPERATURE - EHASE 1

Cak Ridge National Laboratory, Tenn. ,

ORNL-TH-30€4 (Jan. 197%1), 146 p, 117 fig, 12 ref.
The influence of small additicns cf Ti, Zr, ard Hf cn the
mechanical fprogerties cf a modified Hastelloy N with the
nominal composition Ni~12% Mo-7% Cr-0.2% Mn-0.5% C is
described, Test results are frcm numercus, ssall,
labcratory melts and several 100-1b melts fror commercial
vendors. Additions of Ti, Zr, and Hf igprcved the
properties of the alloy both unirradiated and after
irradiation., 1Irradiation termperature had a marked ef fect
upor the properties of all alloys investigated. Generally,
good properties vere okserved when the irradiaticn
temperature was 650 deg C or less and poor when the
temperature was 700 deqg C or higher. W%e attributed this
large effect of irradiation temperature to coarsening
of the carbide structure at the higher temperature.

modified Hastelloy N + irradiation + microstructure +

creep + ductility + tensile properties + allcy ccmpcsition

OTHER CATEGORIES: FCC

FCX690033 ‘ .
McCoy HE + Weir JR + Beatty RI + Cook KH + EKennedy (F +
Litman AF + Gehlkbach RE + Sessions CE + cher J 4

‘MATERIALS FOR. MOLTEN~-SALT REACTORS
- .Cak Ridge National Laboratory, Tenn. _
ORNL-TM~-2511 (May 1969), 43 p, 13 fig, 33 ref.

’ The contents of this THM are the same as an article
in the Feb. 1970 Nuclear Applications, which see.

MSRE + *graphite + Hastelloy N ¢ *modified Hastellcy N +

alloy composition + mechaplcal properties + sealzng +

fluoroborates + corrosion + compatitility + iron allcyc +

*reviews 4 frogress repgort
CTHER CATEGCBIES: ELX ¢+ FCX

*chvocoze

¥cCoy HE + Eeatty RL + Cook WH ¢ Gehlbach RE + Kennedy CR +
Koger J% + Litman AP + Sessions CE + Weir Jb '

NEW DEVELCEMENTS IN MATERIALS FOR MOLTEN-SALT REACTORS

Oak Ridge Naticnal Laboratcry, Tenn,

Nucl. Appl. Tech., 8, 156 (Feb, 1970), 14 p, 1: fig, 34 ref.

Accessicn Numhe; FCE69C043 to FCX7CCC26

 
 

 

Page 94

Category F
Hastelloy N and Related Allcys

FCX70C026 ¥*Ccntinued#
Operating experience with the Molten-Salt Reactcr Experiment
(YSRE) has demcnstrated the excellent compatikility
of the graphite-Hastelloy N-fluoride salt syster at
650 deg C. Several improvemeats in materials are needed
for a molten-salt kreeder reactor with a basic plant
life cf 30 years; specifically, (1) Hastelloy N with
improved resistance to emirittlement Lty thermal neutrcns,
(2) graphite with better dimensicnal stability in a
fast neutron flux, (3) graphite that is sealed to ottein
very low surface permealility, ané {4) a seccrdary coclant
that is inexpensive and has a melting point of albout
400 deq C. A brief description is given of wcrk in
progress to satisfy each of these requirements. Significant
improvements are being wade in each area. (Ib1= papex
was also published as CRNL-T¥-2511,) '

MSRE + graphite 4+ Hastelloy N + modified Hastellcy N +

alloy composition ¢+ mechanical frop2rties + sealing +

molten salts + fluoroborates + corresion + conpat1h:l1ty +

iron alloys + reviews + rrogtress report

CTHER CATEGCEKIES: ELCX + ECX

FXX69C047

McCoy HE

THE INOR-E STORY

Cak FRidge National Lakoratory, Tenn,

Review (Oak Ridge Naticnal laboratory) 3, 35 (Fall 1969)

15 p, 9 figqg.

Semitechnical language reviews the development of
Hastelloy N, and descriltes the current frograr tc imprcve
irradiation stability. Studies of compatibility with
molten tluorides, oxidation resistance, strength and
fabricability led tc the basic nickel-base alloy
containing 15-18% Mo, 6—-8% Cr, 5% Fe, 1% Mn, and 1% Si.
Mcre recent studies of wicrostructure and mechanical
properties as intluenced Ly irradiation shcw the need for
reducing siliccr and mcolybdenum and adding small amounts of
titanium, hafniuw, and niotiugm.

revievs + Hastelloy N + mcdified Hactelloy Kk + development +4

alloy composition :

OTHER CATEGORIES: FCX

Accessicn Number FCX790026 to FXX690047
")

-

Page 9%

Category G
Materials Other than Hastelloy N and Grafphite

"GRX670033
"Stiegler JO + ¥Weir JR

EFFECTS OF 1EREDIATICN ON DUCTILITY
Oak Ridge Naticnal Laboratory, Tenn.
Chap, 11, p. 31t in Cuctility, Papers Presented at a feminal
of the American Society for Metals Gct 14-15, 1967, ASH,
Metals EFark, OChio, 1968, 3Z p, 19 fig, 58 ref.
The mechanisms and effects of radiation damage to metels are
presented, with emphasis on effects on tensile elcnga ticn.
Displacement cascades from fast neutrons and transmutation
effects, including the introducticn of telium, are treated,
Many examples show effects of several variables. Most
results are shown for type 304 stainless steel, ircluding
titanium-modified material, Some results are given fcr
Hastelloy N, molybdenum, and tungsten. Electicr rmicrcgraphs
shcw bubbles and other damage. Also puklished as
CRNI-TM-2019, AC-GAE680028.
*ductility + *stainless steels + Hastelloy N + molyktdenum +
tungsten + *irradiation + tensile properties + micrcstructuie
OTHER CATEGORIES: FBE

GAX680028 :

Stiegler JO + #Heir JR

EFFECTS CF IRRRDIATICN ON CUCTILITY

Oak Ridge Naticnal Laboratory, Tenn. |

ORNL-TH¥-2019 (Jan. 1968), 55 p, 19 fiy, 58 ref.
The mechanisms and effects of radiation damage to metzls are
presented, with emphasis on effects on tensile elcnga ticn.
Displacerxent cascades from fast neutrons and transmutation
effects, including the introductiocn of keliun, are treated.
Many exapples show effects of several variables. Host
results are shown for type 304 stainless steel, including
titanium-nodified material. Some results are given fcr
Hastelloy N, molybdenum, and tungsten. Electicr rmicr<qgraghs
show bubbles and cther damage. Also putlished as pp.
311-342 in Cuctility, Papers presented at a Serirar of the
‘American Scciety fcr Metals Cct 14-15, 1967, A5k, Hetals
Park, Chio, 1968 - AC-GAE670033,

*duct111ty + *stainless steels + Hastelloy » + molytdenum +

tungsten + *1rrad1at10n + ten511e properties + m:cxccttuctuxe

‘OTHER CATEGORIES: FBE

GAX?OOOHS | '
Koger JW. + Litgan AP

- CATASTRCPHIC CCRECSION OF TYPE 104 STAINLE c'I‘BEL IN A

~ SYSTEM CIRCULATING FUSED SCDIUM FIUCECBCFBTE

cak Ridge National Laboratory, Tenn,

ORNL-TI4-2741 (Jan. 1970), 22 ., 5 fig, 12 ref.
‘A type 304 stainless steel liquid level prcbe ccntacted
sodiur flucrcbcrate cortaining 8 mole X sodium fluoride in
an Inconel 600 pump loop at ccnstant tenperatures in the
range S4C tc 690 deg C for 192 hr., The probe exhibited

Accession Nfimbér GAX670033 tc GAX7CCC4S

 
 

 

Fage . 96

Category G
Materials Other than Hastelloy N ard Graghite

GAX?OCOQS *Ccntinued*
heavy attack, evidence ly severe leaching c¢f chrcrium, iron,
manganese, and siliccn from the alloy. Equivalent unifornm
attack was about 4 mils/day. Corrosicn of the stainless
steel, wvhich is inferior to nickel-base allcys in fused
flucrides, becare catastrcphic in this system due to
dissimilar-metal effects.

inconels ¢+ stainless steels ¢+ corrosion ¢ fluoroborates +

liquid level measurement + loop + mclten salts

GCX610002

Adamson G¥ + Crouse RS + Manly WD

INTERIM REPORT ON CORROSICN BY 2IRCCNIUF-BASE FLUCFILES:

Cak EKidge Kational Laboratory, Tenn,

ORNL-2338 (Jan. 1961), 60 p, 34 fig, 3 ref.
The mixture Na¥F-ZrF4-UF4 (50-46-4 mole %), was circulated
in thermal convecticn lcogs for £C€0 to £00C hr at a het-
leg temperature of 1500 deg F. In Inconel-600 loops,
subsurface voids were formed ty selective leachirng cf
chrerium. After 500 hr of oparation the voids were found
to depths of about 10 mils, and the depth increased atout
4 mils/1C0C hr. The effects H>f time, hot-leg temperature,
temperature drop, fluoride purity, lccr size and shape, and
inhibitors on the depth of corrosion wvere studied. The
attack was reduced when a portion of the uraniuvm was
trivalent. A few tests were carried out in loops
constructed from nickel, stainless steels, ircn, Hastelloy
B, rclybdenur, and niobium. A linmited amount of work vas
done on Inconel loops circulating alkali-metal-kase
mixtures (NaF, LiF, KF, UF4) «ith portions of the uranium
in the trivalent state., Reduced attacks were fcurd.

*corrosion + *fluorides + *inconels + iron + metallography +

molybdenumr + thermal convection + 1lJop + nickel ¢+

stainless steels + *polter salts + aiobium i

GCX680030

McCoy HE 4+ McElrcy DL

ELECTFICAL RESISTIVITY ANOMALY 1IN NICKEL*BRSE ALLO!c

Oak Ridge Naticnal Labcratory, Tenn, o

Trans. ASM (Am. Soc. Metals) €1, 73C (Dec. 19€8) , 12';,

16 fig, 17 ref.

The electrical resistivity of eigkt nlckel-bace allcys
‘containing ircn, chricmium, and molybdenum was measured to
1000 €. Alloys with more than SC wt % Ni shcwed a rajid
increase in resistivity betse¢en 400 and 600 C and a
decreasing resistivity from akocut 60C tc 1CCC C. Fcr
these alloys the resistivity below 600 C can ke changed by
annealing and by cold working. The resistivity c¢f alloys
with less than 50 wt % Ni increased with temperature uith a
slope decrease between 400 and 6CC C. The etfects cf
annealing and cold weorking were relatively minor for these

Accession Number GAX7CC04S tc GCX68CC3C

.
 

-

»

Page 97

Category G
Materials Other than Hastelloy N ard Grafphite

GCX68C030 ¥Ccntinued*
alloys. The re51st1v1ty variations dc not un:quely derend
~on any one alloying constituent although there is a weak
correlation with the total nickel contert., Electron
micrcscope results indicate that these changes may ke
associated with short-range crder.
electrical conductivity + nickel + inconels + Hastelloy N +
heat treatments

GDX69C032

Koger JW + Litman AP :

COMPATIBILITY OF MOLYBDENUN-BASE ALLCY TZM WITH LITEIUN
FLUCRIDE-EERYLLIUM FLUOFRILE-THORTYUM FLUORIDE-URANIUN (IV)
FLUCRIDE (68-20-11.7-0.3 mcle %) AT 1100 deg C

cak Ridge National Laboratory, Tenn.

ORNL-TM-2724 (Dec. 1969), 16 f, 2 fig, 9 ref.

The T2ZM alloy (Mo-0.5% Ti-0.08% 2r-0.02% C) showed very
little attack by the fused salt (LiF-BeF1-T1hF4-LUF4,
68-20-11.7-C.3 mole %) at 1100 deg C for 1011 hr,
Corrosion manifested itself as leaching cf titarium ard
possibly zirccnium frcm the alloy. The TZlM alloy exposed
to the salt partially recrystallized, wkile tlkat exposed
to the vapor did nct., This recrystallization was
attributed to the removal of titaniur ard zikeccrium, OD
the tasis cf this single test the magnitude and mechanism
of corrosion indicate no serious froklemgs fcr lcng-teIn
use ot TZM in the vacuum distillation processing schene
for the Folten Salt PBreeder Feactcr. Hcwever, the strength
properties c¢f the T2K allcy would approach thcse of
unalloyed molytkdenum as salt exposure 1ncred<ed- this is
nct censidered a chblem now.

compatibility + corrosion + molykdenum + prccessing +

distillaticn + equipmrent + capsules + PESBER + molten salts

GDX7 10025
Nicholscn EL
CCNCEETUAL CESIGN AND EEVELOPMENT PROGRAM F¥OR THE
MOLYBDENUM REDUCTIVE EXTRACIICN ECUIEFMENT TEST STPN[
Cak Ridge National Laboratory, Tenn. ,
03&L-CF-71~7-2 (July 1971), 45 p, 5 fig. 24 ref. :
- Reductive extraction regrocessing of wclten-calt breeder
reactcr fuel requires that the fuel salt be contacted with
‘molten bismuth containing lithiur and troriur metals (as
reducing agents) in crder to remove protactinium and rare
earths from the fuel salt., Bisruth is extremely corrcsive
te the usual matetrials cf construction for molten salt
systems, but molyhdenum appears to have adequate corrcsion
resistance. ToO date, difficulties in tatricaticr of
molybdenum have ruled against its use for vessels for
engineering-scale experiments tut develcpment wcrk in
progress indicates that equipment for reductive extraction

Accession Number GCX68003C tc GDX710C25

 
 

Page 98

Category G
Materials Other than Hastelloy N ard Grarhite

GDX710025 *Ccntinued*
reprocessing can now ke fabricated frem this natetxal. A

small packed cclumn, rerresentative of a typical equigment

unit in reductive extraction reprocessirqg, will be built
nolybdenum and operated for metallurgical and chemical

of

engineering evaluation in a versatile test stand in which
this and future rolybdenum components may be tested. This

report describes the conceptual designs of the test stand
and rclybdenum equipment and discusses the fakrication an
process development work that ¥ill Lte required before the
equirrent can be designed ard built. A brief summary of
state of the art of molyltdenup metallurgy is alsc include
*conceptual design + *reductive extraction process + '
*extraction columns + *molykdenum ¢ *fakrication +
*bismuth + molten salts + *tecst facilities + !SBE +
development + plans + materials \
OTHER CATEGORIES: LDB ' '

GFX660023

Tolson GM + Taboada A :

A STUDY OF LEAL ANLC LEAD- SALT CORROSION IN THERMAL-
CONVECTION LOOPS

Cak Ridge Mkational Laboratory, Tenn.

ORNL-TM- 1437 (Apr. 1966) 19 g, 10 fiqg, 5 ref.

d

the
d.

Thermal-convection loop tests cf several structural allcys

were cperated using circulating molten lead. Screening
tests included carbon steel letween SC0 and 11CC deg F,
type 410 stainless steel betueen $10 and 1Z1C deqg F,
Croloy 2-1,/4 under toth conditions, and nicbiuvr 1%

zirccrium between 1000 and 14CC deg F. 1Iwc lcofs ccntained

surge tanks in which flucride salts, M-1% Zr alloy, end
graphite were placed in contact with the lead tc determin
the ccmpatibility of these materials in a direct-cooled
lead system, 2ll of the steel lcops terded tc gluqg ir th
ccld regions because dentritic crystals of iron and
chromium formed. The hot-leg attack ccrsisted cf general
surface removal, with a few large pits extending to a
greater depth. The NLk-1% Zr alloy shcwed nc measurable
attack; tovever, nicbium crystals were found in the ccld
leg of a loop that operated SCCC hr.

compatibility + *thermal convection + #*corrosion + #*iron +

linings + *lead + liguid metals + *gass transfer + '

secondary systems + *stainless steels + coolants ’

GGX670034
Tolson GM + Tahoada A ”
MSRE CONTECL ELEMENTS: MANUFACTURE, INSPECTION, DRAWRINGS,
AND SPECIFICATIONS |
Cak Ridge ¥National Laboratory, Tenn.
ORNL- 6123 (July 1967) S3 ¢, 8 fig, 7 ref.
The control elements for the Molten Salt Reactcr are

Accessicn Number GDX7160Z5 to GGXe7CC3Yy

e

e
)

Page 99

Categdry G
Materials Other than Hastelloy N ard Graghite

GGX67C034 *Ccntinued*
GAd2C3-21203 bushings canned in Tnconel. The Lefcrt
includes material selection and development of fakrication
methods, The can was made frcm fully irspected Inccnel
closed by four TIG welds. The Gd203-2A12C3 bushkings were
made by conventional pressing and sintering sethcds after a
special prereacticn step was used. The bushings were
given thermal shock tests, weighed, dimensicnally
insgected, and given a final visual inspection for chips
or cracks. As-built drawings, specificaticns, and
manufacturing procedures are included. By methcds
described in this report, 160 MSRE ccntrol rcd e€lements
were ranufactured.

velding + specifications + rare earths + MSRE +

nickel alloys + incorels + *fabrication + *control rods ¢+

ceramics

GXX68CC39

Metzger GE

SURVEY OF STRUCTURAL MATERIALS FGR EHE FCITEN SALT

EXPERIMENTAL (MCSEL) REACTOR

wright-Patterscn Air Force Base, Chio.

Nucl. Eng. and Design, Vol. 7, No. 1, {(Jan. 1%€8).
Survey of metal-base structural materials for use in polten
lead and fluoride salts at temperatures tetween 5C0 ard
100C deg C., The mechanical fproperties, fakrication and
corrosion properties are considered with resgpect to tte
Molten Salt Experimental MOSFL Reactor Concegt.

*alloy corpesiticr + *ccnverters + lead cooling +

corrosion + fakrication + mechanical properties +

molten salts + niobiur + tantalum

Accession Numker GGX€7C034 to GXX6ECC3S

 
 

Page 100

Category H
keactor Component Develogpment

HAX70C050

Kedl KJ

FLUID DYNAMIC STUDIES OF THE MOITEMN-SALII EEACTCR EXEEEIHENT

CCKE

Oak Ridge National Laboratcry, Tenn.

ORNL~TM-3229 (Nov., 19, 1970), 33 p, 16 fig, 1C ref.
In the MSRE reactor vessel, fluid fuel was circulated at
1200 gpe down through an annular regicn and ugp through 1140
passages in the grarhite core. The core desiqn was based
on preliminary tests in a one-fifth scale wmcdel, follcwed
by detailed measurements with water scluticns ir a
full-scale mockufp cf the reactor vessel and internals.
This report descrites the models, the testing, and the
data from which velccity, pressure drop and flow patterns
are deduced, It also descriltes hcw the reasurerents sere
extrapolated tc molten salt at 1200 deg F in the actuel
reactor. The few olservations possitle in tke reactor were
consistent with the rredicted behavior. |

cores + desiqgn + development + flow measurement +

tluid flow + MSRE + reactcr vessel + models

HBX620006

Smith PG

WATER TEST LCEVELCEMENT OF THE FUEL PUMP FOR THE MSRE

Oak Ridge Natic¢nal Laboratcry, Tenn. :

CRNL-TM-79 (March 1962), 47 p, 19 fiqg, € reft. _
A vertical-shaft, sumg-type cemtrifugal pump with
overhung impeller, of conventicnal hydraulic design wvas
specified for circulating molten salts in the KSFE. This
report describes water tests cf a prctctype, including
hydraulics and the perfcrmance of a spray device for
stripping gas from the circulating liquid.

*development + *¥MSRE + *prototypes ¢+ *pumps + components +

design + hydraulics + testing

OTHER CATEGORIES: MAB

HBX670042
Smith PG
EXPERIENCE WITH HIGH-TEMPERATURE CENTRIFUGAL PUMPS 'IN
NUCLEAR REACTORS AND THEIR APFELICATICHN TC ECLTEN-SAILT
THEEMAL BEEELER REACTORS
Oak Ridge National Laboratory, Tenn.
CRNL-TM-1993 (Sept. 1967) 44 p, 12 fiqg, 8 tab, 24 ref.
Des:gn features, develofrment problems, and operating
experience were compiled for llqu1d-metal— ard
molten-salt-circulating pumps used in various nuclear
reactors and test facilities., The compilaticn sas rade to
determine probler areas and select combinations cf features
for the pumps required Ly eack of the tiree zcltensalt
- systems. The short-shaft pump is favored for the
coolant-salt system because of reliakility, tke lcng

Accessicn Number HAX700050 to HEX670042
»

»

Page 101

Category H.
Reactcr Component Development

HBX670042 *¥Continued* ‘
shaft for fuel and blanket salt systems because it prcvides
greater thermal and radiation protectics tc tke drive motor.

*design + *development + #*liquid metals + *molten salts +

*pumps + *reviews + MSER + two-fluid reacter

HBX69C0S8

trindell 2G o HcGlothlan K

CONCEPTUAL SYSTEM DESIGN DESCRIETICN OF THE SALT EUHP TEST

STAND FCR THE MOLTEN SALT EREFEDER EXPERIHENT

Oak Ridge Naticnal Laboratory, Tenn. j

ORNL-TM-2043 (R2ug. 1969) 53 p, 7 fig, S tak.
A stand is required toc test the salt punmps for the Molten
Salt Ereeder Experiment (MSBE). It will be designed tc
acccrmcdate pumps having capacities rancing fror 3000 tc
700C gpm and ofperating with salt of specific gravities to
3.5 at discharge pressures to 400 psig and temfperatures to
130C deg F ncrrally and to 1400 deg F for short times. Both
the drive-motor electrical supply and tke hteat resoval
system for the loccp will be designed for 1500 hg.
Preventive measures to protect personnel and eguipment from
the hazardcus effects of a salt leak will be taken.

*deqcrlptlon + *MSBF + *pumps + *test facilities + decsign +

develcpment + rolten salts + fplans + testing

HBX690059
Wilson LV + Grindell AG - ‘
PRELIFINAFY SYSTEMS LESIGN BESCRIPTION (TITLE I DESIGN) OF
THE SALT PUMF TEST STANI FOR THE MOLTEN SALT BREEDER
EXPERIMENT
Cak Ridge MNational Latboratory, Tenn.
ORNL-1TN-2780 (Dec. 1969), 100 p, figs, tabs.
The preliminary system design descripticn and tke Title I
design calculaticns of the test stand are presented.
Descriptions, functlons, and design requirements for
compcnents and subsystems are provided. The frincigles of
- operation of the test =tand the safety precautions, and the
maintenance phllosophy are dxscussed. The Quality-Assurance
- Program Plan is being fprepared. ' ,
*description + *MSEE + *pumps + *test facilities +
design + develcpment +. mclten calts + plans +
quallty dsqurance,. , -

PBX70C012

Saith EG =

DbVELOPHENT OF FUEL- AND COOIANT SAIT CENTEIFUGAL EUFES FO?

"THE MSRE - -

Oak Ridge Natiocnal Labo:atory, Ienn. _

CRNL-TM-2987 (Cct 1970), S0 p, 18 flg, 15 ref. _
The two salt purps in the MSRE are vertical-shaft sumr
pumps with overhung impeller and cil-lubricated kearings.

Accession Number HBXE7C04z ta HEX7CCC12

 
 

Péqe 102

Category H
Reactor Component Develcpment

HBX70C012 *Ccntinued*
The fuel pump delivers 1200 gpr and the coclant Fumg 800 qpm
‘of salt at 1000 - 1200 deg F. The fuel pump is designed for
renote replacement of the motcr cr entire rctary element and
includes in the pumfp tank a sgray device fcr rescving xenon
fror the circulating fuel. A replacement fuel fpump with
larger tank and longer shaft was develofped but rever
installed. This report describes the development, testlng
with molten salts and performance in the MSRE.

*developlent + *MSRE + *fpumps + components + design +

experience + hydraulics + maintenance + molten salts +

testing

CTHER CATEGCRIES: MAR

HCX68C037

Kedl FJ + PFKcGlothlan CK

TUBE VIBRATION IN MSRE PRIMARY HEAT EXCHANGEFR

Cak Ridge National Lakoratory, Tenn.

ORNL-TM~-2C98 (Jan. 1968) 43 p, 8 fig, 4 tab, 16 ref.
The primary heat exchanger for tke Mclten Salt Reactor
Experiment was ccmfpleted in 1963, Preoperational tests with
vater revealed excessive tube vibration and high fluié
pressure drop on the shell side ¢f the exchancer,.
Modifications were rade to correct these deficiencies, Fronm
January 1965 through November 1967 the heat exckanger has
operated fcr abcut 14,0C0 hrs in molten salt without
indications of leakage or change in perfcrroarce., -

*design + *develcpsent + *heat exchangers + ¥JFSRE ¢

experience + hydraulics + testing ¢ vikraticn

OTHER CATEGORIES: MAB

HCX710022
Bettis CE + Crcwley &K + Nelms HA + Eickel TW +
Siman-Tov ¥ + Stoddart WC

COMPUTER PROGRAMS FOR MSBR HEAT EXCfiANGERS

Cak Ridge Kational Laloratory, Tenn.

ORNL-TM-2€61% (April 1971), 158 p, 7 fig, 22 ref.
Three programs vere developed to rake design calculaticns
for the heat exchangers fcr molten-salt reactors. The
programs are: for the primary heat exclangers, PRIME);
for the reheaters, RETEX; and for the steam generator-
superheaters, SUPEX. Fach type ot exchanger is described,
the basic equations used in each analysis are given, and
the logic used in each program is discussed briefly ir this
report. The rrcgrags developed sere used in designing the
four 556 MW primary exchangers, €iqgqht 3€.6 MW¥ reaheaters,
and sixteen 121 MW stear genetrator superheaters, All are
basically baffled shell and tule exckancers; the stearx
generator superheater is a U-tube, U-shell exchanger.
Flow diagrams, lists ot input regquired and cutput received,
coprplete program listings, and the nomenclature for tkhe

Accession Numker HBX70001Zz tc HCX71CCZ2

{n
o)

Page 103

Category H
Reactor Component Develcpment

HCX71C022 *Ccntinued*
- programs as well as exacple computer input and cutput fcr
the exchangers described are appended.
*computer codes + *heat exchangers ¢ *MSBR + frirmary salt +
secondaty salts + steam generators + steam cycle + stress +
thermal properties ¢ analysis + fluid flow +
conceptual design + design data + expansion +
single-fluid reactors + fluoroktorates ¢ vitration +
Hastelloy N
OTHER CATEGCERIES: HLX

HFX62C007

Richardson M

DEVELOPMENT OF FREEZE VALVE FCR USE IN FESRE

Cak Ridge National Laboratory, Tenn.

ORNL-TM-128 (Feb. 1962), 24 f, B fig, 2 ref.
Early in the MSRE development prcgrar tkree typec of
devices were tested for blocking flow in small salt lines
by freezing a pluy in a restricted secticn. After 10C
test cycles, one design was caosen for further development
and testing,

development + freeze valves ¢ MSRE + prototypes +

testing

OTHER CATEGORYILS: MAB

HIX660026

Hitch BF + Ross RG + McDuffle HF

TESTS OF VARIOUS PARTICLE FILTERS ¥R REMOVAL OF OIL MISTS

AND HYDROCAREON VAPOR )

Cak Ridge National Labkoratory, Tenn,

ORNL-TM-1€23 (Sept. 1966) 27 ¢, 9 fig, 3 tab.
various filter and adsortent paterials were examined fcr
possible use in the remcval of o0il mists and hydrocarton
vapors. A controlled flow of 0il was irjected into a heated
nickel reactior vessel to cause vaporization and some
cracking of the o0il. Heliuwm flcwing thicugh the reacticn
vessel carried the cil mist aad hydrocarbon vapor thrcugh a
filter system, Filter effectiveness was deterrmined by the
use of a combinaticn of felted metal fiters and ceramic
fibers in a confiquration progosed fcr use in the MSRE.

_fGranulated charcoal removed hydrocarbon vapors {(C-6 and

‘above) in a manner consistent with the establzched
adscrpticn isctherms fcr this material. -

*development + *filters + *off-qas systems + adscr;tlcr +

charccal + compcnents ¢+ filtration + hydrocartons +

materials + mists + MSRE ¢ testing -

HXX64C019

Scott D

COMPONENT DEVELOPMENT IN SUPPCRT OF BESEKRE (EART CF MSKE

SEFIANN EFCG REPT 7/31/6E4)

Accession Number HCX7100ZzZ to HXIX6U4(CC1¢

 
 

 

Page 104

Category H
Reactcr Component Development

HXX640019 *Continued*
Oak Ridge Naticnal Labora tory, 1Ienn.
ORNL-3708 (Nov. 1964), pp 167-190, 2z4 fig.
Develorment and cperaticn of prototype units for evaluation
of performance and maintainakility are descriked in this
paper. Included are the core hydraulic mockup, the heat
~exchanger hydraulic tests, electric heaters, freeze f langes,
freeze valves, contrcl rods and drives, and the fuel
sampler-enricher,
components + develofprent + maintenance + NMSEE +
prototypes + testing
OTHER CATEGORIES: MAD

Accessicn Number HXXo640019 to HXX640019
L

.

Page 10%

Category I
Reactor Design

ITAA650024

Alexander LG ¢ Carter HL + C:aven CW + Janney EB +
Vvan Winkle R

MOLTEN-SALT CONVERTER REBCTGR -- DESIGK STULY AND CCST
ESTIMATES FOF A 1000-MWE STATION

Oak Ridqge Natic¢nal Laboratory, Tenn,

~ ORNL-TM-1060 (Sept. 1965), 348 p, U5 fig, 112 ref.

In 19€1-1962 a study was made of a molten-salt converter
reactor based on technology tp te demcnstrated ir the MSBR.
The ccnceptual design is a one-fluid reactcr «ith
cylindrical graghite mcderator elements in a 20-ft reactor
vessel, Ten cu ft of salt (of 25C0 cu ft) is remcved
daily for recovery and recycle of uranium in a central
plant serving many such reactors. With highly enriched
U-235 feed the equilibrium conversioan ratio is 0.90.
Estimated power costs encourage ccntlnued effcrt cn
mclten-salt reactors

¥*conceptual design + *converters + ¥costs ¢

treeding performance + design data + economics +

fluorination + MSEP + plans

IAA66C030
Kasten PR + Bettis ES + Fauman BF + Carter WL +
McDcnald WB + Rcbertscn RC + Kestsik JH
SUMMAEKY OF MOLTEN-SALT BREELER REACTOR DBSIGN STUDIES
Oak Ridge Naticnal Labcratory, Tenn.
CONF-66-524 (Proc. 2nd Int, Thorium Fuel Cycle Sympcczum,
Gatlinburg, May 3-6, 1966), pp. 41-63, 7 fig, U ref.
This paper discusses molten-salt rcactcr techrclcgy ard
presents a conceptual design, breeding performance, and
- cost estimates for a two-region, two-fluid MSBR with
qraph1te tubes in the core. (A more detailed presentetion
is in CRNL-3996, IAR660025.) '
*conceptual design + #HSBR + *two-fluid reactor +
breeding performance + capital costs + economics +
flowsheets + processing + protactinjum + thcrium

IABHATCOU3

Briggs KB : ' o

EFFECTS OF IRRADIATION ON THE bERVILE [IFE CF THE

MOLTEN-SALIT EEACTOR EXPERIHENT ' 7

Oak Ridge Naticnal Labecratory, Tenn.

ANS Trans. 10{1), (June 1967), pp. 166—167. ,
Therral reutron irradiation adversely afrects the
high- temperature. stress-rupture life ot the Hactellcy that
was used in the MSBRE. An allowvance was made in the design
for damaging eftects of irradjation, tut the ruch better
understanding cf the effects, obtained during the years the
reactor vas being tuilt, indicated thkat the allcwance might
not be sufficient and that the service life of the reactor
should be reevaluated. Concluded that the reactcr vescel

Accessicn Number IAA6L00z4 to TAB670C43

 
Page 106

Category 1
Reactor Design

IAB670043 *Continued*

would have a minimur service life of 20, 00C hr.
*analysis + *¥Hastelloy N + *MSRE + *operatiocn ¢+
*radiation damage + *stress xupture + design + llmlts +
plans + reactor vessel -
CTHER CATEGCRIES: MAA

IAC6€C024 |
Kasten FR + Bettis ES + Fauman HF + Carter WL +
BcDcnald WB + Rcbertson RC + Hestsik JH

SUMMAFY CF MCLTEN-SALT BREECER REACTOR DESIGN STUDIEc

Oak Ridge National Laboratory, Tenn. - -

CRNL-TM-1467 (March 1966), 31 p, 7 fiq, 11 taltles, 4 ref.
A preliminary report cn the conceptual design studies of a
twvo-fluid tvo-region molten-salt therral-breeder reac tor
power staticn cf 1000 Mk{e) capacity. 2 much nore detailed
report on the same studies was sulsequently published as
ORNL-3996, MSRIS accession IAC660025, which see,

*¥S3R + *conceptual design + %*performance + *[ovWwer ccsts +

reactcrs + containment + structures + molten salts +

processing + neutron physics + ¥twc-fluid reactor

IAC66CC25

Kasten PR + Bettis ES + Rokertson RC

DESIGN STUDIES OF 1000-MW{(e) FCITEN-SALT BEEEILER RERCTCES

Cak Eidge MNational Lakoratory, Tenn.

ORNL-3996 (Aug. 1966), 150 p, 43 fiq, £z tables, 3C ref.
Design and evaluaticn studies were made of a twc-regicn
molten-salt thermal-treeder reactcr which uses fuel ard
blanket salts separated by the wvwalls of graphite tubing,
which acts as the moderator. A ccolant salt trarspcrts the
heat frox the primary heat exchanyers to steam generators
and reheaters. The reference design fuel salt is
LiP-Bef2-UF4 (68.3-31.2-0.5 mole %), the blanket salt is
LiF-ThF4-BeF2 (71.0-27.C0-2.0 wcle %), ard the ccclant salt
is NaF-NaBF4 (61.1-38.9 mole %). On-site fuel recycle
processing was assumed, with flucride volatility and
vacuum distillation employed for the fuel salt and direct
protactinium-rescval fprccessing used for the tlanket s=alt.
Estimated power cost is akout 2.7 mills, k¥Wkr, tke specific
inverntcry about 0.7 kg/E¥{(e), the fuel doubling time about
13 yrs and the estimated fuel-cycle ccst is 0.35 mills/kWhr.
General flowsheets and conceptual designs ifor the reactor,
primary heat exchangers, salt circulatirg pumgs and steanm
generators are gresented. Cost and perfcrmance estimates
are also given. Several alternate designs are triefly
described: (a) a modified primary heat exchanger design;
(b) a syster using 580 dey F rather than 7CC deg F
feedwater; .{c) a modular concept using fcur snall reactors
rather than one large reactor; (d) an MSBR cperating sithout
Pa reroval in the chemical frocessing plant; (e) a concept

Accession Number IAB670043 to IAC660025
[ )]

u

“

Page 107

Category I
Reactor Design

JAC66C025 *Continued* o |
in which the fuel salt is cooled Lty direct ccntact with
circulating molten lead; (f) a single-stream core-breeder
with direct Pa removal; ({g) a lead-cooled reactcr withcut
grarhite moderator cperating in the intermediate~to-fast
range (10 to 20 kev), and {(h), a graphite-moderated single-
regicn, single-fluid ceonverter reactor.

*two-fluid reactor + *MSER + *conceptual desicn +

*performance + *power costs + *capital equlpment +

*neutron physics + reactors + pumps + -

heat exchangers + steam generators + steanm systens»+

off-gas systems + fuels + blanket + coolants +

physical properties + breeding perfcrmance +

tuel cycle costs ¢+ protactlnlum + lead + converters +

coollng ‘

IAC70C047
Robertson RC + Eriggs RB + Smith OL + Bettis ES
TWO-FLUID MOLTEN-SALT BREEDEB REACTCE DESIGL STULY (STATUS
AS CF JANUARY 1, 1968).
Oak Ridge Natioral lLaboratory, Tenn.
GBNL-QSZS (Aug. 1970), 80 p, 44 fig, 30 talkles, 45 ref.
The January 1, 1968 status of the concegptual desiqgn study of
a 1C0C MW{e) MSBR pcwer station employing separate fuel and
blanket salts in the reactor is reportec¢., The requirerments
fcr Hastelloy N, the graphite moderator and reflector, and
- for the fissile, fertile and heat-transgcrt salts arze
discussed and the frcperties of the available materials are
‘tabulated. The selected fuel salt is LiF-BeF2-UFU
(68.5-31.,3-0.2 mole %), the blanket salt is LiF-ThF4-EeF2
{11-27-2 mole %) and the coolant salt is NaBF4-NaF
' (92-8 mole %). (The lithium is separated Li-7.) Ccnceptual
designs are presented fcr the reactors, pumps, primary heat
exchangers, drain tanks and steam-generatirg eguifgment.
Flowsheets are given for the main systems, but the of f-gas,
afterheat removal, fuel-processing, and steam-fcwer systenms
~are described cnly in sufficien detail to indicate |
- feasibility and to estimate cg¢sts. The reference design,
vith a pcwer density of 20 kW/liter and an estimated
- 'graphite life of 8 years, has a treedinc ratic cf 1.0€,
" a specific pover of 1.77 MW (t),/kg, and a fuel yield of
4.07%/year. The dimensions of the fcur reactcr vescsels are
- about 14 ft D x 20 ft high. The estimated ccrstructicn cost
-~ of the power staticn is about $141/kW {1968 prices) and,
based on 14,7% fixed charges and €0% plant factcr, the'“
- estirated power prcduction. cost is about 4 mills/k¥h. The
- estimated fuel-cycle cost is 0.7 wrills/k¥h. .
*tvo-fluid reactor + *MSBR + #¥conceptual design +
sperformance + *pover costs + *capital equipment +
*neutron physics + reactors + contrcl rcds + c¢rain tarks +
heat exchangers + structures + stear generators + pumps +

Accession Numter IAC66C025 to IACT00047

 
 

Fage 108

Cateqory I
Reactor Lesign

IAC70C047 *Ccntinued*

steam systems + off-gas systems ¢ containment ¢+ mcltern salts +
coolants + physical froperties ¢+ graphite + neutron flux ¢
breeding performance + fission products + noble petals +

fuel cycle costs + afterheat :

IAC700051

Bettis ES + Robertscn RC

THE DESIGN AND PERFORMANCE FEATURES OF A SINGLE-FLUID

MOLTEN-SALT BREEDER REACTOR

Cak Ridge Kational Laboratory, Tenn.

Nucl. Appl. Tech. 8, 190 (Feb. 1970), 18 p, 9 fig, 5 ref.
A conceptual design has Fkeen made of a single-fluid 1600
M¥ (€¢) MSBR powver staticn. The reactor vessel is 22 ft
in diam x 20 ft high, of Hastelloy N; with graghite
mcderator and reflector. The fuel is U-233 carried it a
LiF-BeF2-ThFY4 rixture which is molten akove 930 deg F.
With continuous chemical processing tc isolate frctactinium
and rerove fission rrodicts, conversion of thorium to U-233
exceeds fissile burnup, The estirated fuel yield is 3. 3%
per year. The estirated construction cost of the station
is comparable to PWE total constructicn costs. The pcuwer
producticn cost, including fuel-cycle and grarphite
replacement costs, with private utility financirng, is
estigated to be less than that for present-day light-water
reactors, largely due to the low fuel-cycle ccst and bhigh
plant thermal efficiency. After some engineering
development, such a plant ApEears feasikle and ;ractlcal.
(Corpanicn papers in the same issue discuss the status of
material developrent, fuel prccessing, and pcterntial «<f
the MSBR ccnceft.)

*MSBR + *conceptual design ¢ *pecformance + X[CWEL cCsts ¢

*capital equiprent + *paterials ¢ *processing +

*fuel cycle costs + plant ¢ reactors +

heat exchangers + punps + cff-gas systenms +

steam systems ¢ maintenance

IAC71C013

Robertson FC (editor)

CONCEPTUAL DESIGN STUDY OF A SINGIE FI1UID FCLTEN-SAILT

BREEDEE EFEACTCE

Oak Ridge Naticnal Labcratory, Tenn. ,

OBNL 4541 (Feb. 1971), 192 p, 9Z fig, €z takles, 12¢ ref.
Concertual design cf a 1000-KW(e) molten-salt thermal
breeder reactor pover station indicates that such a plant
is technically feasitle and econcmically attractive., The
plant operates on the Th - U-233 cycle, using a fuel =salt
of the composition -LiF-BeF2-ThF4-UF4 (71.7-16.C-12,.0-(.3
mcle %). The salt is pumped through a 2z2-ft diam x 2C ft
high graphite-moderated and reflected reactor vessel and
ther thrcugh primary heat exchangers vwhere it is cooled

Accession Number IAC7C0047 to IAC7 10013
»)

Page 109

Category I
Reactor Lesign

IAC71C013 *Continued*
from 1300 deqg F to 1050 deqgq F. The ccre graghite is
replaced by rerote maintenance procedures at 4-yr
intervals. The chief material of constructicr fcr the salt
systems is Hastellcy N improved by additives to increase
the resistance to irradiation damage, 1Tritium, Xe and Kr
are srarged from the circulating tuel salt by helium
bubbles, &An off-gas system removes the fissicn-products
for storage and decay and recycles the helium. 2 1-qm
side stream of fuel salt is ccntinucusly prccessed to

- remcve Pa-233, reccver the bred 0U-233, and to adjust the
fissile content, Heat is transported frcm the fcur primary
heat exchangers by a circulating coolant salt, NaEF4-KaF
(92-8 mole ¥), to steam generators and reheaters sugplying
a 3500 psia 1000 deg F/ 1000 deg F steam turbine. The
specific inventory of the plapt is 1.5 kg fissile/N¥(€),
the breeding ratio is 1.06 and the annual yield is ab(ut .
3.3%., The net therrmal eff1c1mncy is 441%, and the
~estimated capital cost is akouat thke same as fcr a

light-wvater nuclear power station. The fuel-cycle cost is
about 0.8 mills/kWhr, Flowsheets and ccnceptual decsigne of
the rmajcr compcnents are presented. Cost and performance
estimates are tabulated. The princirpal desigr
uncertainties are inr areas of tritium confinement,
fuel-salt processing, graphite and Hastellcy N Lehavicr
under irradiation, suitability of coolant salt, maintenance
procedures, and the bebav1or ct fJSSJCn-prcduct
particulates.

*M3BR + *conceptual design ¢ *performance + ¥[OWEL Cccsts +

*capital equiprment + coolants ¢ physical properties ¢+

graphite + Hastelloy N + reactors + cores + ccntrcl rcds +

drain tanks + heat exchangers + structures + pumps +

steam generators + steam systems + cff-gas systems +

containmrert + peutrcr flux + frccessing + _

breedinyg performance + fission'productc + nckle metals +

fuel cycle costs + afterheat ¢ bubbles ¢+ gas separation +

heliub + maintenance + control + instrumentation +

freeze valves + =1tec + heat generatxon

IAC710014

Rotertson RC -

ESTIMATED CCST OF ADDING A THIRD SALT-CIRCULA!ING SYSIEM
FOR CONTROLLING TRITIUNM- HIGEATICN IN THE 1000 - Fw(e) ESBB

Cak Rldqe National Laboratory, ‘Tenn.,

_ORHL TM-342€ (July 1971), 26 ¢, 2 £ig, 1 ref.

Controlling tritium migraticn to the steam systeg of the
100C-MW{e) reference design MSBE power station ty ‘
interposing a KNO3-NaNOZz-NaNO2 'salt-circulatirg systen to
cherically trap the tritium would add akout $13 millicn to
the total of $206 million nov estimated as the ccst of the
reference plant if Hastelloy N is used to contairt the

Accessicn Number IAC710013 to 'T2C710014

 
 

 

Page 110

Category 1
Reactor Design

TAC710014 | *Continuedx
LiF-BeF2 salt ewmployed to transport heat from the fuel
salt to the nitrate-nitrite salt, and atout $1C rillicn if
Inceclcy could be used., The major expenses associated vith
the modification are the costs of the additicral heat
exchangers ($9 millicn), the additional pumps ($5 million),
and the LiF-BeF2 inventory ($4.8 rgilliorn). Scre cf the
expense is cffset by elimination of some equipment frcm the
feedvater systea ($2 wmillion), through use of less
expensive materials in the steam generatcrs ard reheaters
(abcut $2 rmillion), and through an improved therwmal
efficiency of the plant (worth akcut %1 millicnm). In
additicn to acting as an effective tritium trap the tthird
circulating system would simplify startup and cperaticn of
the MSBR. A simplified flowsheet for the modified plant,
a cell layout showing location of the new equifpternt
physical prcperties of the fluids, design data and cost
estimates for the new and modified equipment are fpresented.

*MSBR + *tritiue + *cagpital ccsts + conceptual design +

loop + coolants + heat exchangers + fumps + pcwer ccsts +

fuel cycle costs ¢+ steam systems

IAD700052

Bettis ES + Bauman HF

EOLTEKE-SAIT CCNVEFTER REACTORS

Oak Ridqge Naticnal Laboratory, 1Tenn. :

Power Engrg. Vol. 74 No. 8, 4z (Augq 197C) 3 p, 2 fiqg.
Develcpment of rapid fuel-salt processing and lcnger-lived
graphite is needed trefore molten-salt Lreeder 1eactcr fpcwer
staticns are built, The molten~salt converter reactor,
however, is generally within present technclcgy; the
graphite would last the lifetime of the plant and the
occasional fuel processing would invclve cnly the
vell-froven fluoride vclatility process. The breeding
ratio would be about 0.84, tut the fuel-cycle ccst wotld be
only about C.8 mrillss/kwh and the construction costs are
expected to be attractively lcw. The reactor design
receiving the most study is the type used successtfully in
the MSKE: a vessel filled with reflectcr and ecderatcr
grarhite having salt flcw passages formed by grooves in
the faces of the pieces. A pelttle-Led type cf ccnvertet
reactor has a structure that easily accommodates
dimensional changes in the graghite and the shkapes are
econcrical to ranufacture, but the salt-to-graphite ratio
in the core is higher than desired. Thermal reactors
operating on the Th - U-233 cycle are more efficient than
those using the U-228 - Pu-~-239 cycle, but Pu car be used
efficiently fcr startur and makeup fuel. Construction of
a molten-salt converter reactcr wculd lead tc lcw-gest
power in the neatr term, provide a market for the Fu
produced in light-water reactors, and give ingetus to

Accessicn Number IAC710014 to IAD7000¢%2

)
»

o

»)

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Category 1I
Reactor [esign

IAD700052 *Continued*
- development of the molten-salt breeder reactor which will
be needed to assure low cost fpover in tie future,
*conceptual design + *performance + MSEF ¢ FSFE +
*fuel cycle costs + *materials + plutorium + *converters

IAE70C059

McWherter JF

MOLTEN SALT BREEDER EXPERIMENT DESIGN EASES

Cak Ridge National Laboratory, Tenn.

ORNL-TM-3177 (Nov. 1970), S4 g, 15 fig, 12 ref.
The design bases for the MSBE are based cn infcrsaticr from
the MSRE and the reference plant design of a 1CCC MR {e)
single~fluid MSBR. Calculations indicate that a 150 K«
(thermal) reactor is a reasonaltle size that meets the
project chbijectives for the MSBE. The primary salt for the
MSBE contains both the fissile (U-233) and the fertile (Th)
material. The heat generated in the primary system is
transferred by a secondary salt 1lcop to the stear
generators. Provisicns are made in the MSEE core to permit
exposure of removaktle graphite samples at conditicns

~similar tc those expected in the PFSBE. The pumps and heat

exchangers in the MSBE are sipilar tc tkose fprcpcsed for
the MSBR.

conceptual design + 6951gn]f'69519n criteria +

design data + grarhite ¢+ irradiation ¢+ materials testing +

MSBE + MSEF + reactors + test facilities

IAF57C047 ' - -
Taube M + Fielcarski M + Poturaj-Gutniak S ¢+ Kowalew A
NEW BOILING SALT FAST BREEDER REACTCE CCNCEETS
Inst. of Nuclear Research, Warsaw, Pcland
Nucl. Engrg. and Design 5 (1967), pp. 109-112, 1 £ig, 30 ref.
Use of molten chlorides in homcgenecus-ccre fast breeder
reacters 1s env151cned.a In the SAWA reactor concept the
core is filled with a molten mixture of :
- NaC1-a1C13-UC13-PuCl3. Heat is removed by boiling in the
core, producing AlCl3 vapor. The WARS ccnceft uses UC13
~and PuCl3 in a mlxture cf NacCl and Kcl, with t0111ng -
‘mercury removing the- heat.

_*b0111ng + *cores + *conceptual declqn + *reactors +

breeding performance + chlorides + tast neutrcn* +

-forelcn + ‘mercury + molten ¢a1t¢

IAF670048 | o |
Taube M + Kcwalew A + Potura]—Gutnlak S + rlelcarbxl v
KCNZEETION DER aALZSIEDEREAKTOREN SAWA UND WARS
Inst. of Nuclear Research, Warsaw, Foland
Kernenergle 10 (1967), pp. 184-186, 1Z ref.
Fast breeder reactcrs with horogeneous cores cf molten
chlorides are not impossilkle: twc ccncepts have been

Accessicn Number IAD700052 to I2F670048

 
Page 112

Category I
Reactor Design

IAF670048 *Continued¥*
envisioned., In the SARA reactor concept the core is filled
vith a molten mixture of NaCl-Al1C13-UC13-PuC1l3., Heat is
remcved by boiling in the core, producing 31C13 vapor.
'The WARS concept uses UCl3 and PuCl3 in a mixture of
NaCl and KC1l, with boiling mercury removing the heat.
(This article, in German, 1= very similar to one in
English: IAF670047.)

*toiling + *cores + *conceptual design + *reactors +

breeding performance + chlorides + fast neutrcns +

foreic¢cn + mercury + mEclten salts

IAF690014

Perry AM

A HIGH-YIFLIC MCLTEN-SRALT EURST REACTOR

Oak Ridge Naticnal Laboratory, Tenn.

Proceedings of the National Topical Meeting on Fast Burst
Reactors, Albtuquerque, N. Mex., Jan. 28-30, 1969,
Conf-690102, 387, 15 p, 11 tiqg, 3 ref.

A pulsed molten-salt reactor appears capable of producing
neutrcn fluences of 10(16th) nvt in neutron-irradiaticn
specimens in single Ltursts with widths cf less than 1 msec.
A reactor design is fpresented shich achieves these goels,
using as fuel lithium-uranium fluocride (73-27 ecle %)
eutectic salt., Neutronic, mechanical, and hydraullc
analyses of the reactor are discussed.,

*polten salts + *reactors + neutron spectra +

nuclear analysis + neutron fluence ¢+ *paterials testirg +

*irradiation + fuels + excursions + neutron sources +

neutron physics + description ¢ *conceptual design +

experiment + fast neutrcns

IBA710005

Tallackson JR

THERMAL RALDIAT ICN TRANSFER OF AFTER HEAT IN MSBR HEAT

EXCHANGERS

Cak Ridge Kational Laboratory, Tenn.

ORNL-TM-314S {(March 1971), 108 p,_u3 fig, 28 ref.
About 40 percent of the nolkle-metal tissior picducts are
expected tc deposit cn metal surfaces in the fuel 1loof of an
MSR, predominantly in the heat exchangers, The rcrmal means
of afterheat removal is continuved circulation of the =alts,
but the design must permit afterkeat resoval frcs the heat
exchangers entirely by radiative heat transfer without
coapromising containment. Steady-state terperature profiles
in £ *'reference design' heat exchangers ranging in size fron
94 PFW(t) to 565 MW (t) were computed. The trarsients
folloving a drain vere estimated for the largest and
smallest heat exchangers. The raximuf temferatures,
occurring 3 to 4 hrs after shutdoun, were estimated tc ke
about 2100 deg F and 1800 deg F in the £65-MW and 94-P%

Accessicn Number IAF67C048 to IEAT1CCCS

i
-

Page 113

Category 1
Reactor Design

IBA710005 *Continued¥ :
units respectively. The calculated temperatures are
believed to be conservatively high. Elasinaticn cf cne of
the two cuter shells frcm the 'reference design'® exchéengers
" would reduce steady-state temperatures ty 2CC deg F tc 300
deg F. It is ccncluded that MSER heat exchangers with
ratings of 500 - 60C MK (t) can Le designed tc acccmrodate
safely this worst case afterheat situaticn,
accidents + afterheat + ccoling + design +
radiation heating + heat exchangers + Leat transfer + MSBE +
MSBR + nokle metals
GTHER CATEGCEIES: HCX

IBB67C039

{Staff Report)

DESIGN STUDY OF a HEAT-EXCHANGE SYSIEP FOE CNE HSEE

CCKCEET

0Oak Ridge Naticnal Laboratory, Tenn,

CRNL-TM-1545 (Sept. 1967), 201 p, 9 fig, 1z tab, 4( ret,
A system is described which uses five types of heat
exchangers -to transfer the heat generated in the reac tcr
ccre cf cne ccrcegpt of a 1000 MW (e) KSBF to the
supercritical steam needed to drive a turbine fcr the
generaticn cf electrical powetr. The two major design
approaches reported here are for flow circuits in which heat
is transferred from the mclten core fuel and fertile
blanket salts to the molten cooclant salt and then tc the
supercritical fluid. 1he Case-3 system involves relatively
high fuel- and blanket-salt fressures in the reactctr core.
These fpressures are reduced in the Case-P systenm Ly
reversal of the flows of the fuel and tlanket salts thrcugh
the reactor core and the respective pumps and exchangers,
vhile the operating pressures of the coclant-salt systenm
are raised above thc¢se in theé Case-A system, The cri teria
used, assumptions made,_relatlcncbxps erplcyed, and the
‘results cbtained in the desiyn for each of the five types of
exchangers used in these cases are repcrted. The restlting
design for the Case-B heat-exchange syster and the
exchangers appears to te the ercst workakle cne.

*deulgn + *heat exchangers + *steam generators +

*two-fluid reactor + blanket + compcnents + ccmputer ccdec +'

wolten salts + MSBR + ‘steam cycle

CTHLB CATEGCBIES' IEC o

IBB71C015
Fraas AE

A NEW APPROACH TO THE DESIGK CF STLA& GENEFATCES FCE

EOLTEN SAIT BEACTOF POWER PLAHTb
Oak Ridge Naticnal Laboratory, Tenn.
CRNL-TM-2953 ({(June 1971), 69 p, 25 fig, 20 ref.
A new tyre of steam generator has been devised to meet the
special requirements of high-tersferature liquid-setal and

Accessicn Number IBA710005 to TEB710C15

 
Page 114
Q Category I
Reactor Design

IBB710015 *Continued* , : ,
molten-salt reactor systems. The basic design concept is
such that boiling heat transfer instakbtilities and their
attendant severe thermal stresses are avoided even for a
temperature difference of as much as 1000 deg F between the
feedvater and the high-temperature liguid, thus givinc good
control characteristics even under startup ccrditiors.

This is accomplished by employing a vertical reentry tutke
geometry with the feedwater entering the becttcoe cf the-
inner swall diameter tube (approximately 1/4 in. diam)
through vhich it flovs upward until evafpcrated tc dryress.
The slightly superheated steam emerging from the top cf the
small central tube then flows tack dcwnward thrcugh the
annulus between the central tube and the outer tuke. A
portion of the heat transferred from the high- tenppratu:e
liquid tc the superheated steam in the annulus is in turn
transferred to the water toiling in the central tube.
Design studies indicate that this type of toiler not only
avoids thermal stress and salt treez:ng Frcblers but it
alsc gives a relatively compact and inexpensive
construction., Further, it apgears tc rake pcssible a
simple plant ccntrcl system sith exceptionally good plant
response to changes in load dewand.

*stear generators + *design + steam systems + control

TED6800 36 ,

Peektles FN !

REMOVAL OF XENCN-135 FROM CIRCULATING FUEL SALT OF THE MSBR

BY MASS TRANSFER TO HELIUM BUBBLES

Cak Ridge National Laboratory, Tenn.

ORNL—-TM~-224% (July 1968), 33 g, 8 fig, 2 tab, 21 ref,

Bemoval of dissolved xenon-135 Lty mass transfer tc helium
bubtles cffers an attractive means of centrolling the
xenon-135 poison level in molten salt breeder reactcrs. To
provide necessary engineering information for evaluation of
the proposed method, the existing data cn rates cf rass
transfer tc gas bubbles were reviewed. Father extensive

- literature references pecint tc relialle equaticrs fcr
prediction of mass transfer rates to single tuklles rising
in stationary liquids under thke two extreme cases of a rigid
bubble interface and ¢f a perfectly mobile butble interface.
In general, experimental data are available wtich supjcrt
these predicticns, Nc reliable critericn for predicting the
transition from one type tehavior tc ancther is available.

*tubbles ¢+ *mass transfer + #%*xenon + analysis + data +

design + models + MSER + reviews

OTHER CATEGORIES: HEX ~

Accession Number IBB710015 to IRD6ECC36
 

 

o

-

Page 115

| ~Category J
Inst:umentation and Controls

JAAT710009

Chang SI
A SYSTEMATIC PROCEDURE FOR DETEREINING SYSTEM FAFAMETEFFS EY
PERFCEMANCE INDEX MINIMIZATION (THBESIS) '

Oak Ridge Naticnal Labcratory, Tenn,

CRNL-TM-3311 (NMay 1971, 97 p, 8 fig, 19 ret.
‘A method was developed for determining optimum control
system parameters by performance index minimizaticn. The
centrcller parareters appearing in a mathematical model of
the system are optimized bty adjustinmg tlem ip such a sanmer
as tc mipnisize certain integral measures of the difference
between the actual and desired output resgcnse. The
minigization is performed according to an optimization
scheme known as the steepest descent prccedure Lty a
coomputer code written fcr the IBE-360. The technigues were
used to solve several protlems in order to demcrstrate the
validity and practicality of the methods, The probless
included control parameter optimizaticn for a nineteerth
order wmodel of the Molten Salt Breeder Feactor. The
method easily found the optimum ccntrcller paraseters for
this system in only 5 minutes of computer tinme.

control + design + MSER + computer codes

JAB69C018
Sides WH
MSBR CONTROL SIUDIFS
Cak Ridge National Latoratory, Tenn.
ORNL~TM~-2489 (June, 1969) 43 , 16 fig, 7 ref. :
A preliminary study was made of thke dynamics and conticl
of a 100C Mw{e), sirgle~fluid ¥SEF by an analog computer
simulation. An abkreviated, lumped-parameter mcdel
vas used. The contrcl system included a steam temperature
controller and a simplified versicn of the MSRE reactcr
temperature control system. The results of the study
"indicate a need for a variatle- speed, secondary-salt
' pump for clcse centrol of the steam temperature. During
~ severe transients, consideralkle care must be taker
in designing the ccntrol system if freezing or overheatlng
of the salts is to Le avoided.
*MSBR + *ccntrcl + *dynaric characteristics + *sxmulatxon +
analog systems + analysis + tehavior + computers +

~efcursions + instrumentaticn + plant + btabilxty + systems
CTHFB CATEGCEIES' BCX

gJAB?OCOl?
Sides WH

CONTROL STUDIES OF A 1000-Hw(e) fiSBE
cak Ridge VNational Laboratory, Tenn.
ORNL-TM-2927 (May, 1970) 46 p, 16 fig, 7 ref.
Preliminary studies of the dynamics and control of
a 1000-Mw(e), single~fluid MSBR were cortinued., Previcus

Accessicn Number JAA710009 to JRAH700017

 
 

Page 11¢

Category J
Instrurentation and Controls

JAB700017 ¥Continued*
studies vere refported in CRNL-T¥-2489, FSBF Control
Studies, W. H. Sides, Jr. An analog sigulaticn cf
an exranded luzped-pararmeter mpodel was used, Steam
temperature contrel was accomplished by applying the
load demand signal directly tc the reactor cutlet
temrerature contrcller as vell as to the steam generators,
*MSBR ¢+ *control + *dynamic characteristics + *sipulaticn +
analog systems 4+ analysis + behavior + computers +
excursions + instrumentation ¢ plant ¢+ stakility + <y¢temc
OTHER CATEGORIES: BCX

JAB710008

Sides WH |

BMSBR CONTEFCL STUDIES: ANALOG SIMULATION PROGRAM

Oak Ridge Naticnal Laboratory, 1Tenn.

ORNL-TM-3102 (May 1971), 29 p, 8 fiq, € ref.
Ar analog computer simulation of the prcoposed 1CCC-MW MSER
vas devised and preliminary studies were rade cf dytawics
and ccntrol. This report descriles thke reactcr glant
model and the computer simulation. The analog simulation
of the plant consisted of a lumped-paraneter reat trarcsfer
model for the ccre, primary heat exchanger, and steam
generator; a two-delayed-neutrcn-qroup ncodel c¢f the
circulating-fuel nuclear kinetics with temperature |
reactivity feedbacks; and the external contrcl system., . So
that the model would have the maximum dynamic range, the
system differential equations were nct linearized, and as
a result the mcdel was severtely limited spatially to
minimize the number of equations. In acditicr, the
Fressure in the water side of the steam generator, as well
as in the rest of the plant, and the physical ficperties of
the salts and vater were takem to be time invariant. The
temperature of the feedwater to the steam cerneratcrs was
alsc held constant. Results and conclusions are given in
CRENI-T¥-2927 (MSRIS accession JAB700C17).

MSBR ¢+ contrcl + simulaticn + dynamic characteristics +

analog systems + computers

OTHER CATEGORIES: JBX

JCX690019

Clark FH + Burke OW

DYNAKIC ANALYSIS CF A SALT SUPERCRITICAL WATEERE HEA1

EXCHANGER AND THROTITLE USED WITH ESBF

Cak Ridge National Laboratory, Tenn.

ORNL-TM-240% (Jan. 1969), 42 f, 24 fiqgq, 6 tables, 3 ref.
A linearized, coarse space mesh rcdel cf a
salt-supercrltlcal sater heat exchanger and the downatream
throttle was simulated on analcg computers. 1he effects on
certain cutput quantities of changes in certain input
quantities were noted. The output quantities were

Accessicn Number JAB700017 to JCX&ESCC1S
=)

.

.}

-~

Page 117

Category J |
Instrugentation and Controls

JCX690019 *Continuedx* '
heat-exchanger vater outlet temperature and pressure, salt
outlet temperature, and enthalpy cutput. 7The input ‘
quantities were heat-exchanger vater inlet temperature and
pressure, salt inlet temperature, salt velccity, and
threttle setting. Changes vwere studied only arcund design
steady state, | ‘ |

*heat exchangers + analysis + dynamic characteristics +

computers + steam systems + feedktack + MSBR + analcg systems +

mathematics + stability + wodels + coolants + heat transfer +

secondary salts + secondary systems + simulation-

CTHER CATEGCEIES: JRAE

JDX67C037

‘Russell JA + Knowles [G

DESCRIPTIOH OF FACILITY RADIATICN AND CCNTAMIMNATICN SYSTEMS

INSTALILEL IN THF MOLTEN-SALT REACTOR EXPERIMENT BLDG. 7503

Oak Ridge Naticnal Labcratory, Tenn.
ORNL-TM-1127 (Eev. 1) ({(August 1967), z3 p. 7 fig.

' A radiaticn mopitoring system continuously and
automatically determines the conditions in the entire
facility and records this information at a central control
panel. When preset values are exceeded, audikle and
visual alarms inside and cutside of the building area are
actuated. Beta-Gamma constant air monitcrs scurd caution
alacrms at 100 ccunts/min and high level alarms at 400C

. counts/min., The gamma monitrons sound tells at 7.5 wE/hr.
Building evacuation alarms are actuated from coinciderce
modules. This report describes the system and triefly
discusses operating experience. An eatclier vercsicn of
‘this document was issuved in May 1965,

*anLronment + *instrumentation + *MSRE +
*radiation measurement + contanmnination + monltorb +
sampling + stack + off-gas- systcmq

OTHER CATEGORIES: MAC

JDX690060
Bauman, CD
FISSICN-PRCEUCT MCNITORIKG IN HIGH-TEHPERATURE GAS-LOCLED
- REACTORS
Cak FRidge National Laboratory, Tenn. . -
ORNL-TM-2791 (Dec. 1969), 33 r, 5 fig, 1 tab, 41 ref.
The report proposes the developrent cf an instrivmenta ticn
_jsysteu capable of 1dent1fy1ng and measuring the accumulation
of fission products in. hlgh-temperature gas-cccled reactor
(HTGQ) ccolant loofp circuits and loop components. [iscussed
‘is the applicability of iomizaticn chamters, Feta and gamma
Spectrometers, charged-wsire prec1p1tators, Cerenkov
‘detectors, filters, diffusion tukes, theérmal gradient tubes,
depcsiticn tubes, and impactors as plateout monitors. It
is recommended that the depositicrn-tuke and
gamua-spectromete: systems be further developed as component

Accessxoa Numher JCX69001° tc JDX6§CCGC

 
 

 

Fage 118

Category J
Instrumentatior and Controls

JDX69C060 *Ccontinued*
plateout monitors and tested first in a high- etperattre gas
loog and then in reactor service.

*development + *fission products + *instrumentaticn '+

*off-gas systems + components + coolant loops + HTGE +

plans t+ testing + monitors

OTHER CATEGORIES: IBD

JEX650020

Engel JR ,

APPLICATICN OF AN ON-LINE LIGITAL COHPUTER T0 A REACTIOR

EXPERTMENT

Cak Ridye National Laboratory, Tenn.

ANS Trans. 8(2), (196%5), rp. 985-586.
The MSRE uses an on-line digital computer fcr acquirirg and
analyzing data. 273 analcg process signals are scanned
every 5 sec., These are stored on magnetic taje every
10 min. Alarms are taken from 171 signals. 2 number of
calculations using current reactcr data are rerfcrmed
pericdically and on demand. Summaries ot the calculations
are recorded by typewriters. The Bunker-Ramc Mcdel-340
computer can supply data for processing by other computers.

*data acqulbxtlon systems + * MSRE + ccmputers ¢

data fprocessing + instrumentation ¢+ operation + plans

OTHER CATEGORIES: HMaC

JFX660027

Moore RL

CLOSED-CIRCUIT TELEVISION VIEHING IN MAINTENANCE OF

RADIOACTIVE SYSTEMS AT CRNIL

Gak Ridge NKational Laboratory, Tenn.

ANS Trans. 9(2), (1966), fpp. 530-531.
Discusses the development of televisicn carera systems for
vieving remote raintenance operations at reactors. {2 '
later, more detailed reference cn this subject is
JACETCO3E.)

description ¢+ equipment + MSRE + remote maintenance +

viewirg devices

CTHER CATEGCRTES: KEBA + MEE

JFX67C036

Moore RL

CLOSED-CIRCUIT TELEVISION VIEKING IN EAIhTELAlCE CF

RADIOACTIVE SYSTEMS AT ORNL :

Oak Ridge Naticnal Laboratory, Tenn.

CRNL-T#-2032 (Nov. 1967), 13 p, 7 fig.
This repcrt discusses factors affecting the use of
closed-circuit television in radicactive systers, ther
describes equirrent used for closed-circuit television
viewing at the Homogeneous Reactor Test and at the MSEKE.
" The results ¢f a radiation test of a miniature,

Accession Number JDX69C0&6( to JFX67CC36
)

a)

»)

Page 119

‘Cateqory-J
Instrumentation and Controls

JFX67C036 *Continuved#*

radiation-resistant television camera are alsc fresented.
*instrumentaticn ¢+ *MSRE + *remote maintenance +
viewing devices + equipment ¢ glass + maintenance +
manipulatcrs + optics + in-pile tests
CTHER CATEGCEIES: KEA

Accessicn Number JFX670036 to JFX670036

 
 

 

Page 120

Category K
Operatiocn and ¥aintenance ' -

KBB630006

Holz PP ‘ S

FEASIRILITY STUDY OF BEHOTE CUTTING ANLC WELDING FOR NUCLEAR

~ PLANT MAINTENANCE : _

Cak Ridge National Laboratory, Tenn.

ORNL-TM-2712 (Nov. 1969), S3 p, 13 fig, 21 ref.
Remote cutting and welding are pctentially valuakle
in the maintenance c¢f radioactive portions of reactors,
particularly if the system is designed to explcit
these technigues. CRNI has started to adapt an orbitel
cut-and-wveld system (developed fcr the Air Fcice) tc
perkxit ccmpletely remote work applications. This report
describes factors involved in radioactive systern
maintenance, sumrarizes some previous work on remote
maintenance development, and explains hcw the autcmated
orbital cutting and welding machinery system may overcome
problems that have Lkeen encountered in nuclear repair wcocrk.
Progress of the ORNL study is summarized, including
descriptions of the prototype equipment and tke results of
machining and welding tests., The report descrites
additional requirements for development to prcvide fully
remcte oreraticns and centrols and proposes a lcng range
program for development of a ccmfplete system fcr radicactive
Ssyster equifpment regplacement by cutting and welding.

*maintenance + *reactors + *remote ueldlng + cuttirg tccls +

- development -

Accessicn Number KBB690006 to KEPF690006

e
*)

Page 121

Category L
Fuel Preraration and FErocessing

LAX690010

Chandler JM + Bolt SE

PREPAFATICN CF ENRICHING SALT {LITHION~-T7 UBANIUH 233
TETRAFLUORIDE) FOR REFUELING THE FECLTEN SALT1 REACICE

Ccak EBidge National Laboratory, Tenn

ORNL-4371 (karch 1969y, 73 p, 24 fig, € ref.

*fuel

The Molten Salt Reactor Experiment has been refueled

with an enriching salt concentrate, Its pregaration

in a shielded call c¢f the Thorium-Uranium Fecycle Facility
at CEXL was required because of the higk uraniuvmg-232
content (222 prm) cf the uranium-2z33. & two-step process
was used in which the uraniur oxide was reduced tc

uranium dicxide by treatment with hydrocen and ccroverted
to vraniur tetraflucride by hydrofluorination., Lithiun
fluoride was added and the eutectic was formed by fusing
the ccroponents, The eutectic %as purified by treatment
with hydrogen, which reduced the corrcsicn grcducts

tc retal and subsequently allowed their removal Ly
filtration. The quality of the product was well within
the requirements established for the ¥SEE. The fuel
concentrate, containing 39 kg cf uraniur (S1.4%
uraniup—233), was rackaged in nine containers of various
sizes (0.5 to 7 kg of uranium) fcr additior tc¢ the reactor
fuel drain tank and in 45 enrichment cagpsules, each
containing 96 g of vranium, for addition to the fkowl of the
fuel circulating pump, The tuel was shipped in shielded
carriers tc the MSRE to accomaoddte the reactor enrichment
schedule,

pregaration + *uranium~233 +¢:corrosion products +

filtration + fluorides + hydrofluorination + kydrcgen +
lithium flucride + MSRE + uranium~232

CTHEER

CATEGCEIES: MCE

LAX70C013
Chandler JM 4 Polt SE

‘URANIUM-233-BEARING SALT PBEPARATICN FCB THE PSRE

Cak Ridge National Laboratory, Tenh. . _ ,
Nuclear Appllcatlons and- Technology, Dec. 1970, 16 p, 5 fig,
2 ref. | o

The MSRE has been refueled ¥ith an enriching salt

,fl'concentrate, lithium-7 fluoride - uranium-233 tetraflucride

(73-27 mcle %) . Slxty—thtee kilograms cf this was prepared

in a shielded cell in the Thorius-Uraniun Recycle Facility
~at Oak Ridge National Laboratory. The preparation process

involved reducing uranium tricxide tc uranius dicxide by

treatment with hydrcgen, converting the uranium dioxide to

fiouranlum tetrafluoride by hydroflucriraticn, ard fusing the
‘urarium tetraflucride with lithium' fluoride., 1Its
‘preparation in a shielded cell vas required because of the

high uranium-232 ccntent (222 ppm) of the uranium. The
product salt, containing 39 kg cf uraniur (91.4%
aranium-233) vas lcw in oxide content (50 ppm) and the

Accession Numher 1A1690010 tc IAX700013

 
 

Page 122

Cateqgory L B
Fuel Preparation and Frocessing rs

LAX700013 *Continued* -
concentration cf the coxz051cn products, chromium, ircn, and .
nickel, was winimal at less than C.5% tctal. (Abstractcr's
ncte: This work was regorted in CRNL-4371., See
LGX6900100.) - : , - .

*fuel preparation + *MSRE + *uranlun-23’ +

corrosion products + hydrcfluorination + lithium fluoride +

reduction + uranium-232

OTHER CATEGORIES: MCD

I1AX710019

Shaffer JH

PREPAEATICN ANT HANDLING OF SALT MIXTURES FOR THE

MOLTEN-SALT REACTIOR EXPERIMENT

Cak Ridge RKational Laktoratory, Tenn,

OBRNL- 4616 (Jan. 1971), 41 g, 30 fig, 36 ref. :
A molten mixtuure of LiF, BeF2, ZrF4, ard UF4 served as the
circulating fuel fcr the ¥olten-Salt Reactor Experimert.
Its secondary coolant fer transferring heat tc an
air-cooled radiator was a molten mixture of LiF ard BeF2.
A third mixture that was chemically identical toc the
coolant mixture was used in place of the fuel fcr
prenuclear cperatices and subsequently to flush the
reactor core after a fuel drain. Apfproximately 26,00C 1b
of these fused flucride mixtures vwere prepared from
component fluoride salts aud loaded intc the reactor
facility by ORNL's Reactor Ch2a2mistry Livision. Techniques
for handling molten fluorides and tkeir prcducticrn fprccess
fcr attaining high chemical purity were developed and
applied simultaneously with the develcpuent ct the
molten-salt nuclear reactor concept. The plans and
operations which were part of the fuelirg ¢f the MSRE are
described.

*polten salts + *procurement + *production +

hydrofluorinaticn + MSRE + locading + beryllxum tluor1de +

lithium fluoride + uranium fluorides + zirccnium ftlucride

L8X68C027

Failen JC ¢+ Cathers GI

FLUOORINATION OF FALLING DBOPLEIS CF FECITEN FIUCRILCE SZLT AS

A MEANS -C¥F RECCVERING URANIUM AND PLUTONIUM

Oak Ridge Naticnal Labcratory, Tenn,

CRNL-4224 (November 1968) 21 p, 7 fig, 10 ref.
A flucriration methcd in whica molten-fluoride droplets fall
countercurrently through fluorine was devised fcr the
reccvery of uranium and plutopium from molten fluoride .
salts., Advantages over methods in whick flucrice is
bubkled thrcugh the salt are: (1) higher tremoval rates for

. both uranium and plutonium, (2) lcver ccrrcsicn rates since .
the molten salt dces nct contict the vessel wall, [
(3) possibility of continuous ocperaticn, and (4) esinirpal \sa)

corrcsicn-groduct ccntarxinatisn of the flucrinated =alt.

Accessicn Number LAX7)0013 to LEX680027
-

»i

Page 123

Category L
Fuel Preraration and Erocessing

181680027 *Continued* '
Experimental equirrent was developed, and small-scale
fluorinations were made using several salt scluticns From
thic data it was calculated that, by using a E-ft-long
fluorination column at 650 deg C, 99.9% cf the vuranium can
be removed from 1l00-micron-diameter droplets cf MSBR fuel.
With an 11-ft-1c¢ng tower at 640 deg C, 99% of the plutonium
can be removed from 100-micron-diameter drcplets cf salt. In
similar exfperiments using salt droplets containing
protactinium-231, no protactinium was fluorinated, even at
temperatures as high as 613 dag C.

*columns + *fluorination + corrosion + corrcsicn prcducts +

polten salts + MSBR + plutonium + protactinium + uranium

ICA67001%4

McNeese LE

CONSTLCERATICNS CF LOW PRESSURE CISTILLATION AND I1E
APPLICATION TO PROCESSING CF FMOLTEN-SALT EEFEDER
REACTCR FUELS

‘0ak Ridqe Naticnal Laboratory, Tenn.

CRNL-TM~-1730 (March 1967), 46 p, 1z fig, 1C ref.
Distillaticn at lcow pressure was examined as a method
for removing rare earth fluorides frcm the fuel strearn
of an MSBR. It was concluded that distillation allows
adequate rare earth fluoride removal with the simultarecus
reccvery ¢f mcre thanm 99.5% of the fuel salt. '
Characteristics of eguilitrium and mclecular distilla ticn
vere noted and expressicns for the relative volatility of
rare earth fluorides were derived for thkese types of
distillaticn. Expressions for the separation potential of
several modes of distillation were derived and reforted rare
earth fluoride relative vclatilities were shown to allow a
qgqreat deal of latitude in still design and cperatioral mode.
It was ccncluded that a single contact stage such as a well

. mixed liquid pool provides adequate rare eartlt flvoride

- removal and that rectification is not reguired. The fkuildup

- of rare earth fluorides at the vaporization surface vas
shown tc seriously reduce the effectiveness of a
distillation system., Liquid circulaticn was shcwr tc te an
effective means fcr preventing this buildup.

*dlstlllatlon + *polten salts + fuels + MSBR + rare earths +

volatxlxty

1CA680008

Hightcwer JR + McNeese LE . '

MEASUBREMENT OF THF RELATIVE VOLATILITIES OF FLUORIDEc OF
CERIUM, LANTHANUM, PRAESODYFKIUNM, NECLCYMIUEK, SAMAFIUFK,
"EURCPIUM, EARIUM, STRONTIUM, YTRIUM, AND ZTIRCONIUNM IN
MIXTURES OF LITHTUM-FLUCRIDE AND EERYILIUM-FIUCFILE

Oak Ridge National Laboratory, Tenn.

ORNL-TM-2058 (Jan. 1968, 43 p, 8 fig, 9 ref.

Cne step in processing the fuel streas ct a gclten

Accessicn Number LBX680027 to LCA680008

 
 

 

Page 124

Category L
Fuel Preparation and EroceSSLBg

ICA680008 *Continued#* '
salt breeder reactor is removial of rare. earth flSSlOD
product fluorides from. the lithiur flucride-beryllium
flucride carrier salt by low pressure distillation.
For designing the distillation system we have Ekeasured
relative vclatilities of the fluorides of cerium, lanthanum,
praesodymium, neodymium, samarium, eurcpiunm, karium,
strentiur, ytriue, and zirconium with respect to lithiunm
fluoride, the major component. The measuremerts were
made using a recirculating equilibrium still operated
at 1000 deg C and at pressure from 0.5 to 1.5 mE mercery.
Errcrs from several scurces uere eatlmated and shown
to be small.

*distillaticn ¢+ *rare earths + beryllium fluoride +

equilibrium + fission products + lithiurm flucride +

mcasurement + MSBR + processing + volatility

ICA690037

Smith FJ + Ferris LM + Thcmpscn CIT

IIQUTC-VAECF ECUILIBRIA IN LITHIUM FLUORIDE-BERYLLICM

FLUORIDE AND LITHIUM FLUCRIDE-BERYIIIUM FIUCHICE-THCHRIUMN
FLUCRIDE SYSTEMS

Oak Ridge Naticnal Labo:atory, Tenn.

ORNL-HQID (June, 1969) 18 p, 4 fiq, 21 ref.
Liquid-vapcr equilibrium data for several lithium tlucr1de—
beryllium fluoride and lithium fluoride-beryllium tflucride-
thoriur flucride systems were oktained ty the trarspiration
methcd over the temperature range of 900 to 1050 deq (.
Relative volatilities, effective activity ccefficients, and
arparent partial pressures are tabulated for the major
components, as vell as for sclutes such as uraniuve
tetraflucride, zirccnium tetrafluoride, caesium fluoride,
rubidium fluoride, and scme rare-earth flucrides. The
values are in reascnable agresment with those reported in
the literature, Results of this study show tkat
distillaticn may nct be feasibple as a primary separations
method in the processing ot single-fluié MSBR fuels.

*equilibriue + *vclatility + beryllium fluoride +

distillation + lithium fluoride + MSBR + rare earthc +.

separations + thoriur flucrides

1CB680007T

Carter WL + Lindauer RB + McNeese 1F

DESIGN CP AN ENGINEERKING-SCALE, VACUUM DISTILLATION

EXPERIMENT FOR MOLTEN-SALT REACTCE FUEL .

Cak Ridge National Laboratory, Tenn,

ORNL-TM-2213 (Nov. 1968), 133 p, 43 fig, 14 ref, :
Experimental equipment has teen c¢esigned tcr an engxneer1ng-
scale deronstration cf vacuum distillation of molten-
salt reactor fuel. The distillation is carried out
at about 1000 deqg C and 1-mm mercury tc sefarate carrier

Accessicn Number ICA680008 to LCBG6EQCC?
Page 12€%

Category L

i . - Fuel Preparation and Erocessing

1CB6580007 *Continuedx
salt frorm less volatile flsc1on products, prlmarlly
the rare earths, The experiment is designed fcr either
ccentinuous salt feeding or for batchwise operatioa.
Sampling of the distillate furnishkes data cn the separation
between salt apnd fissicn products as a functicn of
still bottoms concentration, The equirrent ccnsists
of a 48-1iter feed tank, a 12-liter still, a 10-in, diam x
- 51-in., condenser, a 48-liter receiver, flus asscciated
temperature, frressure and level control instrumentaticn.
All vessels and parts contacted by molten salt are
‘made of Hastelloy N. The unit 'is heated by shtell-
type, ceramic heaters. About 90% of the experimental
program will be devoted to nonradiocactive cperaticn
using mixtures of lithium, beryllium, zirconium, and
selected rare earth fluorides., The experiment will
be ccncluded by distilling a 48-liter batch of uraniup-
free spent fuel from the MSRE.
*distillaticn + *molten salts ¢+ beryllium fluorlde +
Hastelloy N + lithium fluoride + MSRE + rare earths +
zircoriur fluoride

: ICB710007

Hightower JR + McNeese LE '
LC¥-FEESSURk DISTILLATION OF MOLTEN FLUORIDE MIXTURES:
NONRADIOACTIVE TESTS FCR THE MSRE DISTIIIATICN EXEERFIMENT

Cak Ridge National Laboratory, Tenn., '

ORNL- 4434 (Janvary 1971), 52 g, 16 fig, 10 ref.
Equipment built to demonstrate the lcw-jpressure
distillaticn of a 48-liter batch of irradiated fuel salt
from the MSRE consisted of a feed tank, a 12-1liter, ore-
stage still reservcir, a condenser, and a condensate
receiver, The equipment was tested in 1568 by frccessing
six 46-liter batches of nonradioactive [iF-EeF2-ZrF4—-NdE3
(65-30~5~-0,.3 mole %) at 1000 deq C. A distillaticn rate
of 1.5 cu-£ft of salt per day per square root of

‘vaporizatior surtace was achieved in the nonradioactive
tests., Evidences of concentration polarizaticn and/ o1

entrainment were ncted in some ruyns but not in others.

- Automatic operation vas easily maintained in each run,

- although certain deficiencies in the liguid-level measuring
‘devices vwere noted, Condensaticn cf vclatile salt
compcnents in the vacuur lines and metal deposition in the

- feed line to the still fpot are proklers needirg further
~attention. These results showed that use of distillation
‘in MSER fuel salt processing is feasitle and that the test
equipment was satisfactory for use Hlth rddloactlve
material from the MSRE. :

»)

. d1stlllat1cn + entrainment + fluorides + processing +
ki) molten salts + rare earths ¢ liquid level measuremert +
teryllium fluoride + lithium fluoride + zirconium flucride +

Accession Number LCB6B0007 tc LCB710007

 
 

Fage 126

Cateqory L
Fuel Preparation and Erocessing

ICB710007 *Continued*
MSRE ¢ segarations + testing

L1CC710024

Hightowver JR + McNeese LE + Hannaford EA + Cochram FL

LC¥ PEESSUEE DISTILLATION OF A PORTION OF THE FUEL CARRIER

SALT FROM THE MOLTEN SALIT BEACTOF EXFERIFENT

Cak Ekidge National Latoratory, Tenn.

ORNL-4577 (August 1971), 56 p, 17 fig, 10 ref.
High- temperature lov-pressure distillaticn cf 111ad1ated
MSRE fuel carrier salt was demonstrated., Twelve liters of
this salt was distilled in 23 hr with still fct temperatures
of ¢CC-980 deg C and ccndenser pressures of 0.1-0.8 terr.
Eleven condensate samples taken during the rur were aralyzed
for Li, Be, Zr, CS-137, ZR-95, CE-144, EM-WT7, FU-155,
Y-91, SR-90, and SR-89. Effective relative vclatilities,
with resgect tc LiF, fcr Be and Zr aqgreed with values
measured previcusly. Effective relative volatilities for
the slightly volatile materials Ce, Y, and Sr vetre wmuch
higher than previously measured values. The high values
are believed to Le the result of sample contamiraticn,
althcugh concentraticn polarization may have also been a
contributor. The effective relative vclatility fcr C5~-137
was only 20%, cr less, cr previously measured value; no
explanation of this discrepancy is available., 2lthcuch the
cffective relative volatilities for the lanthkanides were
higker than anticirated, the values observed would still
allov adequate recovery of LiF-7 from waste salt streams by
distillaticn.

*jistillation + *experience + *molten salts + *MSRE +

teryllium fluoride + cesiuvm + entrainment + fluorides +

processing + rare earths + liquid level measurement +

lithium flucoride + zirccnium fluoride ¢+ separations +

operation + volatility :

OTHER CATEGORIES: LIX

IDA690012
Ferris LM
SCME ASPECTS OF THE THERMOLYNAMICS OF EBE EXTRACTIION CF
URANIOM, THORIUM, AND RARE EARTHS FhCE MCITEN llTkIUM
FLUCRIDE-EERYLLIUM FLUOEKILCE INTC LIQUID LIIHIUH-BISECIH
SOLUTIONS
Cak EKidge National Laboratory, Tenn.
ORNL-TM-2486 (March 1969), 17 p, 18 ref. -
Expressions for the equilikrium distrikbuticn cf uranlcm.
thoriurx, lanthanum, and other solutes betveen lithium
fluoride-beryllium fluoride sclutions and lithium-bismuth
solutions at 600 to 700 deg C were calculated, using
thermcdyrarmic data frcm the literature. The results
obtained experimentally for uranium were in reascnably
good agreement with the calculated values. However,

Accession Numker LCB7100C7 to LDAG6SCC12
.}

ay

»

\i)

Page 127

Category 1
Fuel Preparation and Procesclng

LDA69C012 *Ccntinuedx
the results for thorium and lanthanur reflect the high
degree of uncertainty that exists in the availatle
thermodynamic data for these solutes. It is ccrccluded,
therefore, that an accurate measure of the relative
extractability of the various solutes can be cbtained
only by experikental means.

beryllium fluoride + bismuth + distritution + equilibrium +

lithium + lithiup flucride + rare earths + thorium +

uranium + *reductive extraction process + *design data

OTHER CATEGORIES: CBX

IDAK90013 ,

Mailen JC + Ferris LN + Ncgueira ED

ESTIMATE CF THE SCLUBILITY OF PROTACTINIUM IN LIQUID BISMUTE

Oak Ridge Naticnal Laboratory, Tenn, |

Inorg. Nucl. Chem, Letters, 5 (1969) pp 8€9-872, € ref.
The sclubility cf fprctactiniuz in bismuth was determined to
be about 1200 ppm at 500 deg <, The method used was the
extracticn cf protactinium from lithium-beryllium fluoride
into bismuth at 600 deg C under ccnditicns such that the
concentraticns of cther metals in bismuth was very low and
then cooling to 500 deg C and sarpling.

solubility + rrotactinium + bismuth '+ molten salts

1DA6900138 _
Dahlke O + Gans W + Knacke O + Mull=r F.
DISSCCIATICN PFESSURE OF EISMUTE IN THF SYSTEM:
BISMUTH-~-THORIUM '
Technischen Hochschule, Aachen and {ernforschungsanlage,
Julich, Germany
Z. Metallk., 60(5) {1969) pp. uo6u~-uc8, 5 fig, 8 ref.
ORNL-tr-2218 _
The dissociation pressure of tismuth in the'cystel
bismuth-thorium was determined using the effusicn methcd due
tc Knudsen betueen 600 and 14)0 deg C. Isothermal
~ dissociation curves and x-ray diacrams ccrrcbcrated the
compcunds thoriur dibismuthide and trithorium
tetrabismuthide which dc¢ not form solid soluticrs The
. supfposed compound dithorium bismuthide was not found using
' measurements of the partial fpressure.
*bismuth + thorium + reductlve extrxctlon process
CTHER CATEGCRIES' CEX

LDA70C01H
Schilling CE + Ferris LM

THE SOLUBILITY OF THORIUM IN IIQUID EISFUTH

Cak Ridge Kational Laboratory, Tenn,. S
Journal of the Less—Common Hetals, 20 (1970) rp. 150-159,
1 fig, 12 ref.
The solubility of thorium in bismuth was determined over the

Accession Number LDA690012 to LDATCCC 14

 
 

FPage 128

Cateqgory L
Fuel Preparation and Processing

LDA70CO014 *Ccntinued® . _
temperature range 450 - 900 degrees C, using a filtraticn
tecrnlque. Data cbtained using mild steel equipment can be
represented by the expression: 1lcqg S (pgp Th) = 7.,7C8EF -
3852/T(deqg K). Several data points were obtained using
molybdenum apparatus. These values are in gccd agreement
with those determined in mild steel.

*solukility + thorium + bismuth + filtration + *data

LDA700015

Smith FJ + Ferris LM ,

MUTUAL INTEFACTICNS OF TEORIUM, NICKEL AND BISMUTH IN

Th-Ni-Bi SOLUTIONS

Cak FRidge National Laboratory, Tenn. :

J. Incrg. Nucl. Chem. 32 (1970) pp 2863~ 2868, 2 119, 8 ref,
Thorium and nickel, dissolved in liquid kismuth, were fcund
tc interact with each cther and with the solvent to fcrm a
solid ternary compound cf the apparent composition thcriunm-
nickel-2 bismuth. This interacticn was studied at 50 deg
intervals, cver the temperature range of 550-700 deg (¢, At
each temperature studied, the joint sclubility cf thorium
and nickel could be exrressed as a mole fraction product.
S = mole fraction thorium x fecle fractichn rnickel. 1The
variation cf the sclubility with temperature can tLe
expressed as log S = 1,115 - €397,T (deg K). 1ke standard
free energyy cf foreaticn cf ThRizPi was estimated to le
-1 kcal/mole at 65C deg C.

¥*solukility ¢ thoriur + pnickel + Ltismuth + #*data

OTHER CATEGORIES: CBX

1DAT700046
Ferris L + Mailen JC ¢ lawrence JJ ¢+ Smith Fg +
Nogqueira EC
EQUILIBRIUM DISTRIBUTION CF ACTINIDE AND LAMEANICE ELEHENTS
BETWEEN MOLTEN FLUCRIDE SALTS ANLC LIQUID 3TSMUTH SOIUTIONS
Oak Ridge Naticnmal Laboratcry, Tenn. ,
J. Inorganic Nucl. Chem, 32, (1970) pp. 201S-.035, ¢ fig,r
21 ref.
The equilibkrium distritutions of several actinide and
lanthanide elerenrts betuween ligquid bismyth solutiomns &nd a
variety of lithium fluoride-ccntaining mclter flucride salts
were Jetermined between 500 and 700 deg C. At each
temperature, the distritution coefficierts (rcle fracticn in
bisruth phase divided by mole fraction in salt) okeyec the
relationship L = (CnLi)K (1st) in whick t is the valence of
the element in the salt phase. Cver the range of coaéitions
investigated, thorium, protactinium and zirccriuvnm existed as
tetravalent srecies in the salt; uranive, tepturium,
plutcniur, awericiue, curium, californium, lanthanum, and
neodymium were trivalent; and eurcpiur was divalent. The
distributicn behavicr c¢f each elenment was affected Ly salt

Accession Numbter LDA7CCO14 tc LLA7CCCYE

N\
”

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®}

Page 129

Category L
Fuel Preparation and Processing

LDAT70C0Ou46 *Continuyedx* '
composition and temperature., Values cf log K(1st) increased
reqularly with decreasing temperature. Wwith lithium
fluoride-~beryllium fluoride-thorium flucride (72<-16-12
mole %) as the salt phase, the values of log KSlst) at
600 deq C fcr the trivalent actinide elements varied
systematically with atomic numker, and passed tkrcugh a
minigum near Z = 96 {curium).

*distribution + actinides + teryllium fluoride + bisruth +

lithium fluoride + prctactinium fluorides + rare earths +

thorium fluorides + zirconium fluoride

LGX650002

McKeese LE + Scott CEL

RECONSTITUTION OF MSR FUEL BY REDUCING URANIUX

HEXAFLUCRIDE GAS TO URANIUM TETRAFLUORIDE IN A MOLTEN SALI1

Oak Ridge Naticnal Laboratory, Tenn

ORNL-TM-1051 (March 11, 196S5), 15 p, 5 tiqg, 7 ref
The direct reducticn c¢f uranium hexaflucoride to uranium
tetrafluoride in a molten salt is propcsed as a steg
in the purification of fuel salt from a molten salt
reactor, This step would replace the ccnventicral
metkcd of reduction in which uranium hexofluoride is
reduced to uranium tetrafluoride power in a hydrcgen-
flucrine flame. Reduction of the uranium hexafluoride
in a molten salt will result in a shcrter and mcre
direct prccess for fuel salt purification. The reduction
is to be effected in two steps which consist cf abscrytion
of uranium hexafluoride into a molten salt conrtaining
uranium tetrafluoride and of reducticn c¢f the resultirg
intermediate fluorides to uvranium tetrafluoride with
hydrogen. EFExperimental data on the atscrpticr sterp
are rresented and irnformaticn concerning the reduct1on
of intermediate fluorides :is ccnsidered.

*hydrcqen + *reducticn + abscrption + chemical reactions +

tluorlne + molten salts + hydrogen_* uranium iluc:zdec

LHX69C011
lindauer EFP + FcGlothlan CK :
DEbIGH,_LONSTRUCTION, AND TESIING CF A IARGE FCLTER*SELT
FIITEF
Oak - Ridge National Laboratory, Ienn.
CRNL-TH~ 2478 (March 1969), 35 Pe 9 fig, 12 ref.
The Mcltenm Salt RPeactcr Experiment uses mixtures of fluoride
salts as fuel. Routine on-site processing cf ttese
- molten salts resultc in formation of corrosion groducts.
" This report describes development, desicn, ccrstructicn,
'1nsta11at1cn, and tectxng of a large salt filter to remove
these corrosion products. The filter is designed tc
remcve approxirately 15 kilograms of corrosion products
from 9000 kilograms of flush and fuel salt at a tempettature

Accessicn Number IDA700046 to L1HX6900 11

 
Page 130

Category L
Fuel Preparation and Erocessing

LRX690011 *Continuedx*

of 1200 deg F. |
*filters + %*MSFE + constructlon + ccrresicn ptcductc +
design + develcprment + fluorides + testing

1TX650023
Lindauer RB o S
MSRE LCESIGN ANLC OFERATICNS REPORT, PART VII, FUEL HANLCLING
AND PROCESSING PLANT
Oak Ridge National Laboratory, Tenn.
ORNL-TN-9C7 (May 1965), 96 p, 20 fig, 8 ref.
The on-site plant is designed to remove cxides frcm tke
salts by treatment with H2-HF mixtures and uraniuam as UFé6
by treatment with fluorine. The repcrt includes Flant
descrlpt1on, safety analyses, and procedures. (A revision
-coverlnq changes in equipment and plans, issued in 19¢€7,
is filed as LI2670013.)
¥*MSRE + *frocessing + equigment + flowsheets +
fluorination + hydrogen + hydrogen ccmpcundc + oferaticn +
oxides + plans + rlant + safety
CTHER CATEGCRIES: MAA

LIX67C013

Lindauver FE

MSRE DESIGN AND OPERATIONS REECET, EART VII, FUEL HANILING

AND ERCCESSING PLANT

Oak Ridge Naticnal Laboratory, Tenn

CRNL-TM~907 Revised (Lec. 1967), 6% g, 19 fiq, 7 ref.
Flowsheets and equigrment for the FSBE Fuel Frocessing
Plant are descrited. The plant is desicned tc remcve
oxides fror the flush and fuel salts ty treatrmert
with bydrcyen-hydrcgen fluoride gas mixtures and to
recover uranium by fluorination. cConsecuences cf the
maximum credible accident are described in addition
to the expected radiation levels during prccessing.

¥*MSRE + *processing + accidents ¢+ equipment + flowsheets +

fluorination + hydrogen + hydrogen ccmpcunds + ofperaticn +

Flant ¢ safety + urariur

CTHEK CATEGCRIES: MAR

LIX69C008

Iindauer FB

PROCESSING OF THE MSRE FLUSH AND FUEI SALTS

Cak Ridge Fational Laboratory, Tenn.

OUNL-TK-2%78 (Aug. 1969), 75 g, 25 fig, 12 ref.
The BSRE Fuel Processing Plant, shakedcun tests cf
equipsent and procedures, and the uranium recovery
operation are descriked. The MSRE fluslt and fuel salt
batches were flucrinated to recover 6.5 and 216 kg
of uranium, respectively. Known losses during prccessing
vere less than 0.1%. Gross beta and gamma decontamination

Accession Numker LHX69C011 tc LIX690008
al

o)

Page 131

| Category 1
Fuel Preparation and Processing

'LIX69C008 *Ccntinued#

factors of 1.2 x 10 (9th) and €.6 x 1G¢8th) were cbtai red.
Corrosion averaged about C.1 mil/hr. The corrosion
product fluorides were reduced and filtered tc¢ grcvide
a carrier salt having a lower concentration cf retallic
contaginants than the orlglnal carrier salt.

*HSRE + *¥processing + corrosion + decontamination +

filtration + f1u0r1nat10n + 10 ses + operation + plant +

uraniuoem \ n ‘

OTHER CATEGORIES: NMCD

IIX690009 :
Chandler JM + Lindauer RB -
PHEEPARATICK ANL PKOCESSINCG OF MSRE FUEL
Oak Ridge Naticnal Labcratory, Tenn., |
CONF-69081 (August 1969), Symposiuk on Reprccessing cf
Nuclear Fuels, Nuclear Metallurgy Volume 15, Ames, Iowa,
August 2%5-27, 1969, pp. 97-120, 8 fig, 4 ret.
The MSRE has been refueled with an enriching salt
concentrate, lithium-7 fluoride-uranium-233 tetrafluoride
(73-27 mole %), which was prepared in a shielded cell in the
Thorium- Uraniur Recycle Facility at CEML. The preparation
process involved reducing uranius tricxide (uranium-232
content, 222 prm) tc uranium dioxide by treatment with
- hydrogen, converting the dioxide to tetraflucride by
hydrcflucrination, and fusing the tetraflucride with
lithiur-7 fluoride. The original KSRE fuel salt, which
contained 220 kg of uranium (3%5% uranium-235), was
flucrinated to volatilize the uranium as the hexafluoride
which was absorbed ‘on sodiur flucride. The urarium was
deccntaminated frow fissicn products by a factor of almost
ten to the ninth, Fluorine utilizaticn averaged 39%.
Corrcsion products were removed from the barren carrier salt
by reduction and filtraticn. Corrosicn rates fcr surfaces
- expcsed toc flucrine dnrlng fluorlnatloa averaged 0.1 wmil/hr
~for 47 hours.

_*fuel pzeparation + *MSRE + *procescan '+ decontaminaticn +
filtraticn + hydroflucrinaticn + hydrogen + lithiunm fluorzde +
- reduction + sodium fluoride +  uranium fluorices

OTHER CATEGORIES' LAX + ¥FCD

LJX660006

Scott CD e Carter WL - ' o

PRELI I NAFY DESIGN STUDY OF A CONTINUOUS FLUORIN&IION-"
-VACUUM-DISTILLATION SYSTEM FOE BEGENEEATILG FUEL ENI
FERTILIE STRERAMS -IN B MSER

vak Ridge National. Laboratory, Tenn.

~ CRNL-3791 (Jan. 1966), 123 p, 31 flg, 39 ref.

A preliminary desiqgn and engineering evaluatlcn is macde
of a conceptual plant for treating the fuel and fertile
streams cf a 1000 MWE MSBR. The reguirements are to
recover unburned fuel and fused salts frcr the fuel

Accessicn thbe:*iIXbQOOOB to LJX660CCe

 
 

 

Page 132

Cateqgory L
Fuel Preraration and Erocessing

LJX660006 *Continueds
- stream and the bred urarium and fused salts from the
fertile stream. Decontaminaticn rust ke sufficiest
for attractive breeding. Plant capacity is 15 cu ft/day
of fuel salt and 105 cu ft/day of fertile salt. The
fuel stream is purified by fluorination and distillat icn.
Volatile fissicn prcduct fluorides are removed Lty sorgption
on sodium fluoride. The fertile stream is flucrina ted
tc rerove bred uranium sufficiently fast to keep a
lovw concentration in the klanket. Tke chief ccrclusicns
of this study are that this process can be engineered
with a normal amount of development work and that :
integration of the p:ocessinq and reactor facilities is of
primary importance in lowering the prccessing ccst. 1lhe
ccst of the prccessing plant contributes about 0.2 m11/kwh
to the fuel cycle cost. :

*design + *MSBR + *processing + blanket + costs +

decontamination + distillaticn + fluorination +

fuel cycle costs + sodium fluoride + uraniue +

- *two-fluid reactcr

1JX660032
Whatley ME ¢+ Carter WL + lindauer RB + FcNeese LE +
Scott CL + Hightower JR
ENGINEERING DEVELOPMENT CF ON-SITE FRCCESSING FCF
MOLTEX-SRLT EBFELFE REACTORS
Oak Ridge Naticnal Laboratory, Tenn.
CONF-66-524 (Proc. 2nd Int. Thorium Fuel Cycle Symgpcsiur,
Gatlinburg, May 3-6, 1966), pp. 653-669, 9 figq, 4 ref.
R processing schemeé is descrilted for a two-fluid MSBR.
The fuel strear is fluorinated to recover uranium and
distilled to recover lithium., The ktlanket salt {LiF- 1hFUY)
is flucrinated tc recover bred U-2z33.
*conceptual design + *distillation + *flucrinaticn +
*processing + costs + flcucsheets + molten salts + MSBR +
thoriunm

1JX670032
Carter RL ¢+ Whatley ME
FUEL AND FLANKET EROCESSING CEVELOPMENT FOR MOLTEN SAL1
BREEDER REACTIORS , ‘
Cak Ridge National Laboratory, Tenn.
OBNL-TM-1€£2 (June 1967) 52 p, 10 £fig, 13 ref.
This document describes the fuel and Lklanket ;rcces=e< for
" the two-fluid MSBR, giving the 1967 status of the
technology and outlining the needed developwert. Tt is
concluded that the rrincipal needs are to develop the
vacuum distillation and protactinium reascval cgeratiors,
vhich have been demcnstrated in the labcratory tut not on
an engineering scale. A program to develof ccrtirucus
flucride volatility, ligquid-phase reduction-reconstitution,
iaproved xenon control, and special instrurentaticr should

Accessicn Number 1JX660006 to LJX6700232
»

-

.}

Page 133

, Cateqory L e
Fuel Preparatlon 1nd frocessing

1JX670032 *Continued*
also be a major develcgmental effort., B2n estimate of
manpower and cost for developing MSBR fuel and fertile
processes indicates that it will require 288 manyears of
effort over a 6-year period at a total cost cf abcut
$18,0C0, COC,

*development + *¥MSBR + *processing + Llanket + ccsts +

distillaticn + fuels + prctactinium + two-tluid reactor

LKX620003 |

Carter WL + Milford RP + Stockdale &G

DESIGN STUDTES AND COST ESTIMATES OF T¥O FLUORIDE

VOLATILITY PLANIS

Cak Ridge National Laboratory, Tenn.,

ORNL-TM-522 (Oct. 10, 1962), 81 p, 25 fiqg, 16 ref
Studies are made for 1,2 cu ftysday ané 12 cu ft/day plants
processing fuel frcr a 1000-MWE one-region fused salt
converter reactor. Two conditions were considered
for the smaller plant: (1) retenticn cf the nacte
‘salt fcr protactinium-233 decay and recovery ty a second
fluorination and (2) discard cf all protactinjun-233
as vaste after the first fluorination. The larger pleant
wvas considered only for the case cf prctactiniur decay
and recovery. With protactinium recovery, the capital
cost is $25,750,00C for the larger plant and 12,556, (00
for the smaller plant. Operating cost is £2,241,00C fcr
the larger plact and $1,103,00C for the smaller plant.
With protactinium discard the cagital ccst is $10,183,000
‘for the sraller plant. Processing ccnsists of fluoriration,
abscrpticn on scdiurp fluoride, condensation in cold
traps, reduction in a hydrogen-fluorine flagme, disscltticn
in makeur salt and recycle to the reactor,

*capital costs + *converters + *operating ccsts + ¥frccessirg +

aksorgption + design + mpclten salts ¢ protactinium + reactors +

sodiur fluoride + volatlllty '

OTHFR LATEGORIES' LJX

[KX700030

McNeese LE

ENGINEERING DEVELCPMENT STUDIEb EOR HOLIEN-CALT BREEDER
'REACTOR PROCESSING NO, 1 =
Cak Ridge Kational Laboratory, Tenn.

ORNL~TM 3C53 (Nov., 1970), 85 g, 9 fig. 24 ref.

Several operations assoc13ted a1th MSBR prccessing are under

~.study. This refcrt describes {1) a recently completed

- facility for semi-continuous engxneerlng experiments oh.
reductive extraction,: (2) experiments related tc the
develcgment cf electrolytic cells for use with molten salt
and bismuth, (3) consideration of selective crystallization
of thoriur bissuthide from bismuth~-thorium-rare earth
solutions as a means for separating thcrium frcs the 1are

Accessicn Number LJX670032z to LKX700C3C

 
 

Page 134

Category 1L
Fuel Preparation and Processing

LKX70C030 *Ccntinued#
earths, and (4) a computer ccde that calculates the nuclear,
chemical, and physical procesises occurring in the fuel
stream of an MSBER., This work was carried cut in the
Chemical Technclogy Division during the period Cctober
through Cecember 1968. _
*development ¢+ *processing + *MSBR + bismuth +
computer codes + contactors + electrolysis ¢ rare earths +
reductive extracticn prccess + thorium :

IKX710001

McNeese LE

FNGINEERING DEVELOFMENT STULTES EOB MOLTEN-SALT BREEDER

REACTOR PROCESSING No. 2

Cak Ridge National Laboratory, Tenr.

ORNL-TM-3137 {Jan. 1971) 102 g, 50 fig, 5 ref.
Several operations associatec with MSBR prcces s1cq are under
study. This rercrt describes (1) a proposed reductive-
extraction flowsheet for a single tluid MSEF, (2) material-
balance calculations that shcw the effects ¢f the remcval
times for zirccnium, alkali retals and alkaline earths,
europium, and protactinium on reactor perfcrrance and that
indicate the magnitudes of the heat generation and mass
flows associated with the reactor off-gas, (3) calculated
results shoving the stead y-state performance of a
protactinium isolation systena, {(4) an evaluaticr cf the use
of the protactipnium isclation system to limit the uranium
concentration in the klanket cf a single-fluid MSBR, (5)
calculaticns tc predict the steady-state performance of a
rare-earth removal system Lased on reductive extracticn, (6)
prelirinary testing of the semicontinuous reductive-
extraction facility, (7) experiments related to the
develcrment cf electrolytic cells for use with molten salt
and bismuth, and (8) installation ot equipsert at the MSRE
fcr demonstrating lcw-fpressure distillation of molten salt
using irradiated MSRE fuel carrier salt. This scrk was
carried cut by the Chemical Technology I[ivisicn durinc the
period January - Xarch 1969.

*development + *processing + *MSBR + bismuth + electrolysis +

heat generation + MSEE + protactinium + rare earths +

reductive extractic¢n frccess + zirconium

L1XX660031

Blankenship FF

CHEMICAL SEFARATIONS IN MOLTEN FLUORIDES

Oak Ridge Naticnal Labcratory, Tenn.

CGNF -660524 (Proc. 2nd Int. Thorium Fuel Cycle Sympcsiuve,.

atlxnburg, May 3-6, 1966), pp. 647-652, 8 ref. '

There are several favoratle reactions ir mclten flucr1de
systems that wculd permit rapid removal of fission prcducts,
Xenon and krypton are only slightly scluble ard transfer
intc any gas phase. Iodine can be removed Ly sparging¢ with

Accession Number LEKX7CC03C tc LXX660031
"

¥

w}

Page 135

- Category 1
Fuel Preparation and Processing

LXX66C031 *Ccntinued* , , i
a mixture of H2 and HF, Rare earths can be separated by
distillaticn, by reducticn to insoluble intermetallic
beryllides in the salt, or bty reduction and extractior into
molten Li-Bi alloys. Protactlnlum can te removed by oxide
precipitation. '

*chemistry + *fission prcdactc + *molten salts +

*separations + beryllium + chemical reactions +

distillaticp + experiment + icdine + oxides + processing +

protactinium + rare earths + reductive extraction prccess

LXX70C029

whatley ME + McNeese LE + Carter HL + Ferrls 1M +
Nicholscn EL -

ENGINEEKIKG DEVELCEMENT OF THEF MSBR FUEL RECYCLE

Oak Ridge National Labcratory, Tenn.

Nucli. Appl. Tech. 8, 170 (Feb. 1970), 9 p, £ fig, 1€ ref,

" The MSBR being developed at CFKL requires continuous
processing of the fuel salt, lithiur flucride-beryllicm
flucride-thoriur fluoride (72-16-12 mole %) containing
approximately 0.3 mole % uranium=233 tetraflucride. The
reactcr and procescinq plant are planned as an integrel
system, The main functions cf the prccessing plart will be
tc isclate protactirium-233 trom the neutron flux and to
remove the rare-earth fission prcducts. The nethcd being
develcped involves the selective chemical reduction of the
various components into liquid kismuth s=cluticns at
approximately 600 deg C, utilizing multistage counter-
current extraction. Protactinium, which is easily
separated frcm uranium, thorium, and the rare earths, would
be trapped in the salt prhase in a storacge "tank 1lccated
between two extracticn contactors and allowed to decay to
uranium-233, FEare earths would le sirilarly sefarated from
thorium; hcwever this cperation will be more difficult

~because the separation factors are lower. Electrclytic
~cells would be used to introduce reductant into the bismuth
' phase at the cathode and to return extracted uater1a1= to

the salt phase at the anode.

*develoPment + *9ng1neer1ng + ¥MSBR + *¥processing +

tismuth + chemical reacticns + extraction columns +

fission groducts + flowsheets -+ fuel cycle + molten ¢a1ts +

MSBR + prccessing + serarations + '

reductive extraction process

OTHER CATEGORIES' LDX

1xx710021

McNeese LE -

ENGINEERING DEVELCPMENT STUDIES FOR MSBER PROCESSING Nc. 3

Oak Ridge Naticnal Labcratory, Tenn, , -

CHENL-TM-3138 (May 1971), 97 p, 35 fig, 13 ref. -
This repcrt describes (1) calculated steady—state
performance of a protactlnlum 1sc1at1cn sthen fcr

Acce551cn Numbet LXXG60031 to Lxx7100¢1

 
 

 

Page 136

Category L
Fuel Preparation and Erocessing

1XX710021 *Continued#* '
reactors fueled with uranium or plutonlum, (2) material-
balance calculations showving the effect of fissicn prcduct
remcval times cn reactcr performance; (3) experiments on
reductive extraction in a mild-steel flcw-thrcugh
facility; (4) a siwmulation of a flow control system for the
semicontinuous reductive extracticn system; (5) develcgment
of electrolytic cells for use sith molten salt and bismuthjg
(6) an analysis of the transfer cf materials in
electrolytic cells; (7) measurement of axial dispersicn in
packed columns using immiscitle liquids baving large
density differences; (8) calculated heat generation reates
and temperatures in a protactiniur extracticn cclumn; and
(9) 1lcw-fressute distillaticn of irradiated MSEEF fuel
carrier salt.,

“*¥distillaticn + *procesqlnq + *protactinium +

treductive extraction process + tisguthk + electrclysis +

extracticn columns + fissicn rroducts + heat generatlon +

molten salts + MSRE ¢ plutonium + uranium

OTHER CATEGORIES: LDX + ICC

LXX710026

McNeese LE '

ENGINEERING DEVELCPMENT STULCIES FOR MSBR PROCESSING NC., U

Oak Ridge Natiocnal Labcratecry, Tenn.

ORNL-TH¥-3139 (Aug. 1971), 124 p, 67 fig, 1€ ref.
This repcrt describes (1) exgperiments on the hydrodynamics
of packed column operation, carried out in a gild-steel
reductive extracticr facility, () measurerent cf axial
dispersion in packed columns in which immisciltle fluids
leaving large density differences are flcwing
ccuntercurrently, (3) a simplified methcd for estimating
the effect of axial dispersion on ccuntercurrent coluan
perfcrmance, (4) estimates of the effect of axial dispersion
in packed column contactors used for MSBR frccessing,
(5) measurements of axial dispersion coefficients in an
‘open bubkle column, (6) experiments related to the
development of electrolytic cells for use with mclten salt
and tiswuth, (7) the design and installation of the Flow:
Electrol ytic Cell Facility, (8) the calikraticn cf an
orifice~head pct flcumeter for use with the Flow .
Electrolytic Cell Facility, {9) the development of an
induction type of Lismuth-salt interface detectcr, and
(10) calculaticns regarding the removal of ThF4 from
nrolten-salt streams ty reductive extractiocr.

*extracticn columns + *prccessing ¢ bismuth + dispersion +

electrolysis + flov measurement + liquid level measurerent +

molter salts + reductive extraction process + ,

thorium fluorides

Accession Number LXX71C021 tc Lxx710026'
»)

Page 137

Category ¥
'~ MSRE

MAC68C0O3Y

Tallackson JE

NUCLEAR AND PROCESS INSTRUMENTATICN -- EART IIA, MSEFE

DESIGN ANL OFERATICNS FEPORT

Oak Ridge Naticnal Laboratory, Tenn. - :

ORNL-TM-729 (Feb. 1968), 397 p, 180 fiqg, 1Cz ref.
The first part of this document gives a general description
of the entire MSRE instrumentation and ccntrcl system
including ccntrol of auxiliary equipment and the instrument
power system, Considerations which influenced the decsign
are also discussed and the physical laycut of the
instrumentation system is descrited. The seccnd part is
a detailed description cf the safety instrumentation and
nuclear control systems, 1Included are neutrcr instrumnents,
-safety circuits, ccntrcl-rod system, and the heat load
control, Also described are the instrugents and
interconnections of the radiation and contamination
monitoring system, process radiation mcrnitcrs, and the
data logger-cowputer system.

*design + *design criteria + *instrumentaticn + *MSRE +

*radiation measurement + *cafety + computers + control +

control rods + control-rod drives -+

data acJuisiticn systems + electrical circuits +

health physics + monitors + off-gas systems
‘OTHER CATEGORIES: JBX :

MAD6 90004
Gabkard CH
DESIGN ANLC CONSTRUCTION OF CORE IRRAEIATION SPnCIMEN ARRAY
FOR MSRE RUNS 19 ard 20 '
Cak Ridge National Laboratory, Tenn.
ORNL-TM-2743 (Dec. 22, 1969) 23 p, 7 fig, 0 ref,
A new MSEE core specimen array was desicned ard fabricated
tc regplace the type cf metallurgical surveillance specimen
.array that was used in the MSRE througb Run 1€. The
main purpose of the_new array is to measure the capture-
to-absorption ratio of- uraniug-233 and to deterzire
‘the effects of salt velocity, turtulence, and surface
- finish on the deposition of fissicn products on graphite
- and on Hastelloy N. Two additicnal test specimerns
- were 1nc1uded, cne cf - pyrolytic graphite to determine-
" if there is permeatlon of fuel salt cr its ccrstituents
- intc the graphite ard cne of Hastelloy b to expose a
series of electron mlcrcscope screens ir a tzapge
 gas fpccket, ,
*MSRE + *irradiation + d951qn + fatrlcatlon + *uran:un-233 +
cross sections + fissicn prcdicts + adsorption + graphite +
*Hastelloy N + surveillance + pyrocarhcn + intrusicr
OTHER CATEGORIES' MBX

MaAX650019

»

Accessicn Numberx MAC680034 to "MAX6500 19

 
 

 

 

Page 138

Category M
MSRE

MAX650019 *Continued*

Robertson RC

MSRE DESIGN ANL OPERATIONS REPORT, PART I, DESCRIPTION OF

REACTOR DESIGN

Cak Rjdge National Laboratory, Tenn.

ORNL-1IM-728 (Jan. 1965), 567 p, 112 fig, 61 tables, 176 ref.
This report is one of a series descrikirg the MSRE. The
reactcr and other majcr ccmponents and systems are fully
described and detailed flowsheets show process data and
instrumentation for both the main and auxiliary systeuns,
The fuel salt was LiF-BeF2-ZrF4-UF4 and the ccolant szlt
used to transport heat from the primary heat exchanger to
the air-coocled radiator (heat sink) was LliF-EeF2,

All parts of the systems in ccntact witl the =alts weire
fabricated of standard Hastelloy K., Cperating temperatures
for the fuel salt were 1175 deg F to 12zt deg F, and for
the coolant salt were 1025 deg F to 1100 deqgq F., & dreéin
tank, cooled by boiling water in thimkles, was used tc
store the fuel salt. 1The site facilities, building
services, containment cells and cther structures, and the
varicus reactcr system electrical circuits, are also
described. (Nuclear calculaticns, operaticnal ;rccedtres
pericrmance data, and maintenance aspects are given in
companion reports ORNL-TM-729, 73C, 731, 732, 733, SOE,
9C9, 910, and S11.)

*MSRE + *design + reactors + heat exchangers + pumps +

off-gas systems ¢+ drain tanks + graphite + Hastelloy N 4

processing + molten salts + experiment +

temperature peasurement + thermal shield ¢ absorbers ¢

electrical circuits + freeze valves +

liquid level measurerment + sarppling + flowsheets +

instrumentation + maintenance

OTHER CATEGORIES: 1IAB

MBX640003

McDonald #B _

MSRE GESIGN'AND CCNSTRUCTION (PART OF MSRP SEMIANN PBRCG

REPORT 7,31/€4) '

Cak Ridge National Laboratory, Tenn.

ORNL- 3708 (Nov. 196u), FEe <42-8Z2, 44 fig.
This paper gives a brief description of the MSEKE desxcn and
considerable detail (including 3% phctecraphs) ¢f ccmionent
fabrication and installation, swhich was nearing complet1on
at the time of writing in Auqust, 19€4,

*constructicn + *MSRE + ccmpoments + description +

fabrication + progress report

OTHER CATEGORIES: MAB

MBX700002

Webkster BH ,
QUALITY-ASSURANCE PRACTICES IN CONSTRUCTION AND MAINTENANCE QF THE

Accessicn Number MAX650019 to XEX700002
")

a)

Page 139

Category M
MSRE

MBX700002 *Continueds
MOLTEN~-SALT REACTOR. EXPERIMENT

Cak Ridge kational Laboratory, Tenn,

ORNL-TN-2999 (June 1970), 106 p, 4 figqg.
The MSRE was built at ORNL to demcnstrate the practicality
of the molten-salt reactor concept., Site construction and
installation of auxiliary systems were Lty cutside
ccentractors, while the frimary reactor systems were
installed by ORNL forces. Lesign, prccurement,
ccocnstruction, and maintenance followed 2SHE codes, ORXNL
practices, and special procedures develcped fcr the MERE by
the ORNL Reactor Divisicn group primarily responsible for
gquality assurance. This report descrikes the picgram, the
proklems that were enccuntered, and the lessons that were

- learned. Four years of relialtle cperation cf the MSRE

proved the success cf the quality-assurance program.

*MSRLE + quality assurance + construction + incspecticn +

maintenance + frccedures + testing + welding

CTHER CATEGCEHIES: MEX '

MCA66C001 - '

Guyaon RH + Haubenrelch PN ¢ Engel JR

ASRE DESIGN AND OPERATIONS REEORT FART XI, TEST ERCCER2AM

Cak Eidge National Laboratory, Tenn.

ORNL-TM-911 (Nov. 1966) 84 p, 0 fig, 2 ref.
The test program for operaticn of the MSRE with U-235 fuel
is divided intc¢ 4 major phases: precritical testing,
initial critical measurements; low-power measurements, and
reactor capability investigaticons, Witkin each fhase
individual tests are briefly described to define the
objectives and general procedures to te follcwed. Internal
Test Memcs are identified which contain detailed procedures
and check lists used¢ in performinc the tests

*MSRE + *plans + *testing + *cperation + primary system +

secondary systems + containment + compcnents + exgerimecnt +

surveillarce + procedures

CTHER CATEGCRIES: KAE

MCAGBCO0Y

Engel JE
¥SRE DESIGN AND OPERA*IIONc REFOET, FAET XI-A, TEST EFCCBAH

FCR URANIUN-233 OPERATION

0ak Ridge Naticnal Labora tory, ‘Tenn.
ORNL TM-2304 (Sept. 1968), 18 p, 1 fig, 9 ref.

General rlans for cperating -the FSRE with uran1um-233
fuel are outlined.  The equiprent and procedure fcr

lcading highly radioactive uranium-z33 are described.
Experirent glans 1nc1ude measurement of the initial critical
loading, control-rod calikration, evaluaticn ¢f temperature,
fuel concentraticon and power coefficients of reactivity,
pover calibration, control-system tests, and dyrarics

tests. Continued reactcr operation will permit study

hcceséion Numker MBX700CC2 tc MCA680004

 
 

Page 140

Category M

ESRE (]
MCA680004 *Continued¥ “ '
of the reactivity behavior, noise analySLb, Beasurement
of the capture-to-fission ratic fcr urarium 233, and
additional investigation cf salt, uranium, fission-
product, and materials tehavior. . o
*MSRE ¢+ *plans + *urapium-233 + *operation +
dynamic characteristics + fuels + stability +
‘reactivity + *surveillance + *loading + criticality +
*experiment + chemistry + materials ¢+ fission products +
nuclear analysis
OTHER CATEGORIES: KAB

HCB650021

Swith AN '

MSRE TESIGN ANLC OPERATIONS REPORT, PART IX, SAFETY

PROCEDUORES AND EMERGENCY EIANS

Cak Ridge National Laboratory, Tenn.

ORNL-TM-9CS (June 1965), 46 g, 16 fiq, 3 ref.
Contains brief description ot lasic plar, erergency
philcsophy, organization and responsitilities, emergency
procedures, description of possible local emergencies and-
plans of action, and btackc¢round inforration.

MSRE + orperaticn + plans + procedures + safety .

MCB650022

Guymon RH -

MSRE LCESIGN ANL OPERATIONS REPORT, PART VIII, OPERATING

PROCEDORES (VOL. I) | | '

Cak Ridge National Laboratory, Tenn.

ORNL-TM~9C8 (Vcl. I), (Dec. 1965), 478 p.
Contains krief training material, descripticr cf auxiliary
systems (electrical, air, water, ventilation, etc.),
and coaplete startup check list fcr eacl syster.

MSIE + operaticn ¢+ operatcrs + fprocedures ¢ startug

NCB66 0029

Guymon RH

MSRE LCESIGN ANL CPERATIONS REPORT. Part VIII, OPERATING

PROCEDURES (VOL IT)

Cak Ridge National Latoratory

ORNL-TM-908 (Vcl II), (Jan. 1966), 539 p.
Contains procedures for reactcr startug, fuel cau;lzng,
instrument calibration and heat balances, response to
unusual operating ccnditions, routine okservaticns,
reactor shutdown, shutdeown operations, mainterarce.
Includes checklists for various operations.

MSRE + operation + plans + procedures

iw

MCBAICOSY , - : -
Guymon EH + Haukenreich PN | ' ) ' | L J
OPERATING SAFETY LINMITS FCR THE MSRE . |

Accession Nfimber HCAfiGGOOu tc MCB6SCCEL
-}

)

Page 141

Category M
MSRE

MCB69COSY *Ccntinued*
Cak EKidge PNational Laboratory, Tenr, -
ORNL-TM-733 (3rd revision) (July 1969), 11 p.
This document prescribes limits for parameters descriking
the operating conditicons of the MSRE. It covers all items
directly related to the health and safety cf the public.
Selected items affecting only the safety of the operators
and the protection of the Experirent against disabling
accidents are alsc included. Earlier editions were
“issued in April 1965, Rugust 1965, and Septeskter 1966,
*MSRE + *cgreration + *safety limits ¢ procedures ¢+ safety
CTHER CATEGCRIES: KAB '

MCB71€012
Guymon RH : - '
MSRE PROCEDURES FOR THE PERIGCD BETWEEN EXAFKIN2TICN 2NL
ULTIMATE LCISECSAL (PHASE-III OF LCECOMMISSION PROGRAN)
Oak Ridge Naticnal Labcratory, Tenn, ,
CRNL-TM-3253 (Feb., 1971), 41 p, 3 fig, S retf.
This document describes the condition of the ESEE and
specifies procedures to ke fcllowed after the
pcst-creration exarinations and before the ultimate
"disposal of the fissile and radicactive material in the
reactor. The fuel salt will be kept frozen in the sezled
drain tanks, within secondary containmert whcse cnly
opening is through filters to a stack. Surveillance will
consist of remote monitoring ‘and daily visits by X-10 fplant
perscnnel. Perscnnel access will be controlled Lty the
security fence around the reactor tuilding. The MSRE
Procedures specify remedial actions for abnormal
conditions, Also specified are procedures and
responsibilities fcr maintenaace, modifications, and
removal of surplus eguipment.
MSRE + prccedures + storage + surveillance +
administration + containment ¢ tlowsheets + rainterance +
operatlcn + planc + testlng '

"HCLGGOOOS

Ball &3
SIMULATCRS FCR TBAINING HOLTEN*SALT REACTOR EXPERIMENT
OPERATORS

‘Cak Rldqe National Laboratory, Tenn.

ORNL-TM-1445 (Apr. 5, 1966), 25 p, 11 fig, 4 ref.

Two on-site reactor kinetics simulators were developed

. for training operatcrc cf the PFolten-Salt Feactor

- Experiment (MSRE) in nuclear itartup and pcuer-]evel
_operating procedures. Both s+mu1atora were set up
on general purpose, portable 3lectronic Asscciates, Inc.,
TR-10 analog computers and we:-e ccnnected tc the reac tcr
ccentrel and instrumentation system. The training
program was successfully completed, Also, the reactor

Accessicon Number MCB690054 to FKCCE€0005

 
 

Page. 142

Category M
ESRE

MCC66 0005 *Continued* , -
ccntrol and instrumentaticn system, the operating
procedures, and the rod and radiator-docr drives vere
checked out. Scre rincr modifications were made to
the system as a result of the experience with tkese
sirulators

analog systemns + *HSBE + *tra1n1ng + sipulaticn +

*operators + startup + testing

MCCo70044
Guymon RH :
MSRE == TEFAINIRKRG, PREPARATION FOR OPERATION AND OPERATING
TECHNIQUES
Cak Ridge National Laboratory, Tenn,
Suppl. tc ANS Trans (10), Conf. on Eeactor Cp. Exp.
(July 1967), p. 35.
Three pericds c¢f training were given: Initial (systen
famil iarity pricr tc¢ checkout), Frecritical (2 simulator
was hooked to actual controls and instrument readcuts), and
Pre-pcver (more simulatcr work, operation of power systems).
Operations procedures and practices are alsc d1=cus=ed.
MSRE + operators + rrocedures + training

¥CD680010

Haukenreich PN + Engel JR

RECENT EXEERIENCE WITH THE MOLTEN-SALT REACTOR EXPERIMENT

Oak Ridge Naticnal Labcratory, Tenn.

Summary of paper presented at ANS Winter Meeting, .

dashington D. C. Ncv., 10-15, 1968, AKS Transactions 11 (2)
619, 1 p. '

Operating experience with 0U-235 fuel in the MSEF is
described. From initial criticality in Jupne 1, 1965 to
March 19€8, the reactcr was critical 11,51% hr and
generated 9005 equivalent full-gcwer hours cf erergy &t
pcwer levels tc 8 Mw, The shutdown in March 19€€& was tc
perecit changeover tc U-233 fuel. Cperation with U-23°¢
demonstrated system reliakility and raterials ccrmratitility.
Infcrraticn was collected on fission-product lehavior,
nuclear characteristics, and reactivity pehavicr. Operating
difficulties were associated with a very small oil legk in
the fuel circulating punrp, the fuel sampler, air-line
disccrnects, and seccndary cooling blowers. Some remote-
maintenance technhiques were demonstrated., (A ccrvenient
sumrary cf MSRE cperating experience through June 196¢ is
published in Nucl. &appl. Tech. 8, 118 {(Feb. 1970). See
MCD70000 1.) | - :

*MSRE + *operaticn + *exferience ¢+ *uranium- 235 +

redct1v1ty + components + fission products + tuelc +

rare cases + maintenance

CTHER CATEGCRIES: KRB

HCD69C017

Accession Numker MCC66C00CE to HCDGGCC17
-)

t)

Page 143

Category H
NSRE

MCD69C017 *Ccntinued*
Engel JR + Hautenreich PN
OPERATION OF THE MOLTEN SALI BEACTCB EXEERIMEN WITH U-233
FUEIL
Oak Ridge Naticnal Laboratory, Tenn.
Summary of paper presented at ANS Ccnference c¢n Reactcr
Operating Experience, San Juvan, Fuerto Rico, Oct. 1-3,
1969, ANS Transactions, 12 {(Suppl), 10, 2 p, 3 ref.
The MSRE was operated fcr 3000 equivalent full-power houts
from September, 1968 through September, 1969 with U-2:3
fuel, Static and dynamic chaacteristics of the reactcr vere
essentially as predicted. Special fuel samples were taken
to evaluate U-233 integral cross-sectiopn ratics in an MSR
spectrum. Experiments were conducted to measure cover-gas
entrainment in the fuel loop and its effect cr fissior-
product stripping. Fissicn-product distribution was
studied with a remote gamma-ray spectroceter. £mall amounts
of rlutoniur flucride uwere added to the fuel., Fuel
chemistry and materials compatibtility studies ccntinued to
shcw good system behavicr. (Freprints of the full parfer
(25 p, 8 fig, 14 ref) are availaile fros the authcrs;
similar paterial is frresented in Nucl. Appl. Tech. 8, 118
(Fek. 1970) . See MCD700001.)
*MSRE + *cperation + *experience + fuels + *uranius-233 +
*plutcniug + *nuclear analysis + dynamic characteristics +
redctlvxty + fission products + bullles + cross secticns +
corrosiocn + gacma spectrcmetry + inert gases ¢ noble metals +
rare gases + control rods + pumps + cff-gas systers + heliugm +
argon + ccver gas + void fractlonc
CTHER CATEGCRIES. KAB

MCD69COE

.Haubenrelch _EN
- MOLTEN-SALT 'REACTOR PROGRESS..

Cak, Ridge National Laboratory, Tenn.

- Nucl. Engrg. Int. Vol. 14, No. 155 (Aprll 1969),

pp. -325-329, 3 fig.

- This article tcuches brlefly on earller MSE technology
development, then describes the MSRE and its cperating
experience, (A pore recent, lengthier paper along the
same lifes is in the Fekt. 1970 Kucl. Apfrl. Tech.) '

*ex perience + %*¥MSRE +. *ope:atlon + design + MCRP +
startup :

~ CTHER carsscaxfs- KAE

MCD69C0E2

Blumberg F + Dyer FF + Houtzeel A

MSRE USES REMOTIE GAMMA SPECTRCHETR! FCE FISSICN EBCIUCT
DEECSITICN STUDIES

Qak Ridge Naticnal Laboratory, Tenn,

ANS Trans. 12{2) (Dec. 1969), p. 8ucz.

Accessicn Number ¥CD69C017 to MCL69(CCE2

 
Page 144

Category ¥
MSRE

BCD69C062  *Continued*
Describes adaptation of gamwa ray spectrosccpy and resote
handling tc investigate fissicn product behavior in
inaccessible portions of MSRE system. Enabled study cf
depcsition of fissicn products on metal surfaces of the
system wvhere heat removal is of particular imfcrtance,

Equiprent consisted of GE(Li) diode detector placed at the

exit of a oné-eighth-inch ty 12—1n. lonc aperture and a

400~-channel analyzer. ‘
*design + *gamma spectrométry + *MSKE + analysis +
dep051t10n + experience + fission prcducts + cperatlcr +
remote waintenance :
CTHER CATEGCFIES: JLX + HLE

Mcp7¢Ccoc

Haubenreich FN + Engel JR

EXPERIENCE WITH THE MOLTEX-SALT REACTCE EXPEBIEELT

Cak EKidge Rkational Leboratory, Tenn,

Nucl. Appl. Tech. 8: 118 (1970) 19 p, 6 fig, 16 ref.
The MESRE is an 8-Mvw {th) reactor in whict mclten flucride
salt at 1200 degrees F circulates through a core of
graphite bars, 1Its purpose was tc demcnstrate the
practicality of key features of molten~-salt power’
reactors. Operation with 0-23% in the fuel salt amcurnted
to SCCO0 equivalent full-povwer hours betseen June 1965
and March 1968. At the end of a 15-wcnth demcpstraticn
of reliability the reactor was shut down and the C-
235 was stripped frcm the salt in on-site fluorination
equipment., U-233 was added to the salt and cperaticn
vas resured in Octcber 1968. Cver 2500 EFFH has teen
produced with U-233 through July 1969. The. MSRE has
shown that salt handling in an operating reactor is
quite practical, the salt chemistry is sell-Lkehaved,
there is practically no corrosion, the nuclear
characteristics are very close to predictions, arnd the
syster is dynaxically stable. <Containment of fission
products has been excellent, compcnent ferfcrmarce
has been good, and maintenance of radioactive ccmponents
has been accomplished safely and witkout unreascnable
delay.

*experience + *MSRE ¢+ *maintenance ¢ *operaticn + ccm;cnent= +

descripticn + flucrinaticn + rerformance + reactivity +

reliability

OTHER CATEGORIES: KAB % KEC

MDA620001

Fngel JR + Haulbenreich PN ‘

TEMPEFATUERES IN THE MSRE CORE DURING STEADY-STATE POHER
OPERATION

Cak Ridge Rational Laboratory, Tenn.

ORNL-TM~378 (Necv. 5, 1962) 58 p, 14 fig, B ref.

Accession Numbter MCD690062 tc KLA620001
.)

*)

Page 145

Category M
MSRE

MDA620001 *Ccntinued®
Overall fuel and graphite temperature distributicns were
calculated for a detailed hydraulic and nuclear
representation of the MSRE fueled with tighly erriched
uranium- 235, These temperature distributions were
importance and volume weighted to okttair nuclear and kulk
mean temperatures fcr bcth materials., At the design fpower
level of 10 Mw, with the reactor inlet and cutlet
temperatures at 1175 deg F and 1225 deyg F, respectively, the
nuclear mean fuel temperature is 1213 deg F. The bulk
average temperature of the fuel in the reactor vessel
(excluding the volute). is 1198 deq F. For the came
conditions and with no fuel permeation, the graphite nuclear
and bulk mean temperatures are 1z%7 deg P and 1226 deq F,
resgectively. Fuel rermeation of 2% of the graphite volume
raises these values to 1264 deg F and 1231 deg F,
respectively. Power coefficients of reactivity are
calculated under various asaumptlons of system temperature
control.

*analysis + ¥*MSRE + #*cores . + *reactivity ¢ fluid flow +

heat generation + neutron flux + nuclear analysis +

*therral effects + calculations

MDA62(002
Prince BE + Engel JR
TEMPERATURE ANL REACTIVITY COEFFICIENT AVERAGING IN THE MERE
Oak Ridge Naticnal Labcratory, Tenn. '
CRNL-TM-379 (Cct. 15, 1962) 26 p, € ftiq, & ref,
Use is made of the concept of *nuclear averagye temperature!
to relate the spatial temperature prcfiles in fuel and
grarhite attained during high power operation of the MSRE
to the neutron multiplication constant. Based cr twc-groug
perturbation thecry, temperature weighting functions for
fuel and graphite are derived, frcm which the nuclear
average temperatures may be calculated. Similarly,
importance-averaged temperaturec coefficients cf reactivity
are defined. The values of the coefficients calculated for
the MSRE were -4,8 x 10 (~-5th) deg F fcr the fuel and -7.3 x
10(-Sth) for the grarhite. These values refer to a reactor
~ fueled with salt which does nct ccntain therice. They were .
about 5% larger than the values okttained freco a cne-reglon,
homcgenecus reactor nodel, thus reflecting the variation in
~ the fuel volume fraction throughcut the reactcr and the
effect of the ccntrcl rod thimbles on the flux protfiles.
*MSRE + #*nyclear analysis + *reactivity + *thermal effects +

‘models + neutrcn physics + calculatzcnc + methods

CTHER CATEGCBIE5-- EEX = L

MDA630002

Haubenreich EN

INHERENT NEUTRON SOURCE Ih CLEAN MSEE FUEL SAIT

Accession Numtgr MDAGZC001 tc MLCAG63CC(C2

 
 

Page 146

Category N
MSREF

MDA63C002 *Continued*

Cak Ridge Kational Laboratory, Tenn.

ORNL-TM-611 (Aug. 1963), 17 p, 6 ref.
Alpha particles from uranium interact vith berylliur and
flucrine tc¢ prcduce an inherent source of neutrons in the
MSRE fuel salt. The spontanecus fissicr scurce is
relatively insignificant. Calculations are described
vhich predict an inkerent source of 3 tc 5 x 1C (5th)
neutronsysec in the 25 cu ft of salt in the MSFE core.

calculations + MSRE + neutron sources + fu€ls + :

polten salts

HDA64COO01]
Engel JR + Haubenreich PN + Prince EE
MSRE NEUTRON SOURCE REQUIREMERXNIS
Cak Ridge National Laboratory, Tenn.
ORNL-T TM-935 (Sept. 1964), 37 ¢, 11 fig, 6 ref.
The alpha-n source inherent in the uranium-235 fuel salt
meets all the safety requirements for a neutron source in
the MSKE. Subcritical flux distributions were calculated
to deterrine the ccrbination cf external source strencth
and detector sensitivity required for mcnitcring the
Ceactivity., If more sensitive detectors than the servo-
driven fission chamkters are installed in the instrumert
shaft to monitor the filling cperation, the calculaticns
indicate that the required scurce strencth car ke red uwced
fror 4 x 10{7th) ny/sec to 7 1 10(6th) n/sec. BAn antimony-
beryllium source with an initial strength cf 4 x 10 (8 th)
n/sec would still groduce 7 x 1C(ێth) n/sec one year after
installation. Abstractor's ncte: The external scurce
ultimately selected for use in MSRE was americivm-curium-
terylliuz,
*calculations + *neutron sources + *MSRE + safety +
neutron flux + nuclear analysis

MDAG6U 0002

Engel J® + Prince BE

CRITICALITY FACTCRS IN MSRE FUEL STORAGE AND CRAIN TARNKS

Oak Ridge Naticnal Labcratcry, Tenn, ,

ORNL-TM-759 (Sept. 1964), 18 p, 3 fig, = ref.
Calculations indicate that there is no danger in the fuel
storage or drain tanks with uranium-2z3S fuel c=alt of rcrmal
ccncentration even if the salt is frozen and cooled to
20 deg C and the tank is suktmerqged in vater, frcvided the
uranium remains evenly dispersed in the salt. 1If
segregation of the uranium occurs durinc freezirg ared all
the uranium accumulates in a region near the center of a
tank, criticality will occur at zC deqg C tcr ccrcentrations
factors of 4 or more. Criticality can ke avoided by keeping
the salt molten or ty dividing the fuel charge agcng twc or
more tanks befcre it is allowed to freeze.

*calculations + *criticality + *freezing + *fuels +

Accessicn Number MDAE€3CO0CZz to NMIAGUCCC2

"
| H

»}

»}

Page 147

Category M
~ MSRE

¥DA6 40002 *Continued* ) - |
uranius-235 + *drain tanks + storage + density +
phase equilibria + *MSRE + nuclear analysis
OTHER CATEGORIES' BGX

MDA64 0006 '

Haubenreich PN + Engel JR + Prlnce EE + Claiborne Hc

MSRE DESIGN ANT OPERATIONS REPORT PART III -- NUCLEAR ANAIY‘IS

Oak Ridge Naticnal Laboratcry, Tenn.

CBNL-TM-73C (Feb. 1964), 199 p, 63 fig, 52 ref.
Early calculations of effects of core size and fuel-
tc-graphite ratioc had determined the core design., This

- report describes the calculated nuclear characteristics

of the MSRE with 3 fuel compositions. <(Cne had thoriur
and highly enriched U-235; a second, hichly erriched
U-23% and nc thorium; and a third, 35% enriched U-235 and
no thorium, all in a carrier salt of lithium, beryllium
and zirccnium fluorides. Calculated quantities include
critical loadings, fluxes, temperature distrilkutions,
temperature coefficients, delayed neutrcn effects,
ccntrcl red worth, dynamics, and neutron sources.

*MSRE + *nuclear analysxs + contrcl rods + criticality +

delayed neutrons + heat generation + neutron flux +

reactivity + shielding

OTHER CATEGORIES: BBX + BCX

¥DAKHK Q007 '
Beall SE + Haubenreich PN * Llndauer KE + Tallackson JF

MSRE TESIGN ANT CPERATIONS REPOBT, PART V -— REACTOR
SAFETY ANALYSIS REPORT o :
Cak Ridge National Laboratory, Tenn,
ORNL-TM-732 (Rug. 1964) , 300 g, 109 fig.. 50 ref.
"The MSRE is descrited with empkasis cn ccmpcnent design,
instrumentation and ccntrols, site, layout, and containment.
Plans and staff for startup aid operation are outlinecd.
Three different fuel,compos;txcnc and a power level
- of 10 Mw wvere considered. Nu:clear incidents that
'concelvably could cause danag: are dualyzed ard it is
cencluded that ncne could breach the containment. The
- secondary containment design is shown tc be adequate even if
~all the fuel were srilled in- the cell. _ :
*analySLS + *MSRE + *safety + acc1dents + containmert +
contrcl + instrurentaticn -

CTHER CATEGCRIES: BGX +,Jax7

MDA65C001

Ball SJ + Kerlin TW

STABILITY ANALYSIS OF THE HOIIEL SA[I FEACTCE EXEESIHENT

Cak Ridge National Lakoratory, Tenn,

ORNL-1TM-107C (Dec. 1965), 80 ¢, 23 fig, 20 ref.
A detailed analysis shows that the Mclten-Salt keactotl
Experiment is inherently stable with uranium-23% fuel.

Accession.Numbérfinbhéuoooz to MIA650001

 
 

 

Fage 148

Category M
MSEE

MDA650001 *Continued* , - o
It bhas sluggish trarsient response at low power, kLut
this creates no safety or operaticnal fptoblems. 1The
study included analysis of the transient response,
frequence response, and pole configuration. 1he effects
0f changes in the mathematical model for the systenm
and in the characteristic parameters uwere studied.
A systematic analysis was also made to find the set
of parameters, within the estimated uncertainty range
of the design values, that gives the least stakle condition.
The system was found to Le irhberently stable fcr this

condition, as well as fcr the design condltlon."Comparlbonb'

are made with previous models which underestimated
stability. Reascns are given to explain the increase
in stability with increasing pcwer level.,
*¥MSRE + *stability + *dynamic characteristics + *uranium-235 +
analysis + nuclear analysis ® calculaticns + *modelc
OTHER CATEGORIES: BCX

MDA660003

Prince BE

PERICL HERSUREHENTS ON THE MOLTEN SALT REACTOR EXPERINENT

DURING FUEL CIRCULATION: THECRY AND EXEEEIREM

Cak Ridge National Lahkoratory, Tenn.

ORNL-TM-1626 (Oct. 1966), 36 g, 8 fig, 10 ref.
A theory of period dependence cr the fuel circulaticn is
develored from the gemeral space-dependent reactor kinetics
equations. A procedure for computer evaluaticn cf the
resulting inhour-tyge equaticn is presented, together
with numerical results relating the reactivity tc the
otserved asymptotic period, both with the fuel circulating
and with it stationary. The calculated reactivity
difference betveen the time-independent flux conditioms for
the noncirculating and the circulatinc fuel states is in
close agreement with the value inferred from the MSRE rod
calibration experiments., Rod-kump pericd measurerents made
with the fuel circulating were converted to differential rod
vorth by use of this model. These results ate ccrpared with
similar rod sensitivity measurements made with the fuel
stationary. The rod sensitivities measured under these two
ccnditions agree favorably, within the limits of precision
of the period measurements., Due to the prchlesr cf
maintaining adequate precision, however, the period-rod
sensitivity measurements provide a less conclusive test of
the theoretical model than the reactivity difference letveen
the time-independent flux conditicns.

*analysis + *circulaticn + *dynamics tests + %experiment +

*¥SRE ¢+ control rods + criticality + delayed reutrcrs +

experience + mcdels + reactivity + startup + uranium-235

CTHER CATEGCEIES: ECX ¢+ MLC

MDAGECOOY

Accession Numkber MDAG6S5CO0CT1 tc MLA66CCCY

1
»

Page 149

Category M
MSRE

MDA66COCH4  *Ccntinued* -

Engel JR + Haubenreich PN + PBall SJ

ANALYSIS OF FILLING ACCIDERIS IN MSKE

Cak Ridge National Laboratory, Tenn

ORNL-TM-497 (Aug. 1966), 41 p, 14 fig, 2 ref,.
Whenever the MSRFE is shut down, the fuel salt is draired
frox the ccre. Then, -dvring a normal startup, the
graphite and the fuel are preheated and the ccntrcl
rcds are pcsitioned-sb-that the reactor remains subcritical
while it is belnq filled, Certain atncimal circusstarces
could result in criticality and a power excursion -
in the partially filled core., Varicus fcstulated
incidents were surveyed and the worse case was analyzed
in detail, This case involved selective freezirg
in the drain tanks to ccncentrate the uranium in the
molten salt fraction, Physical restrictions cn the
fill rate and safety actions of control rods and gas
control valves limited the calculated pcwer and tem;exature

" excursions sc that any damage to the reactor would

be prevented, Abstractor's note: It was subseguently
shown that the degree of uranium concentration required
for a serious filling accident cannot Le attained by
partial freezing of the salt., Protective circuits
and administrative procedurea to prevent atbncrral fills
were retained.

*MSRE + *accidents + *analysis + *freez:ng + excursicrs +

*simulaticn + *nuclear analysis

CTHER CATEGCRIES: BGX ' :

MDAGK7C038

Haubenreich EN

SAFETY ASPECTS OF THE HSRE ,

Cak BRidge National Laboratory, Tenn, _

Nucl. Safety Vecl. 8 Nc. 3 (1967) pp. 226-235, 2 tig, 12 ref.

- Fluid-fuel and solid-fuel reactors, alttough sirgilar in
‘ultimate containment regulrements, differ in the kinds of
‘accidents that can cause syster damage. Scre tluid fuels
"are susceptible tc segregation, and filling accidents are
more llkely, but afterheat is more easily bandled. The
‘surviving fluid-fuel concept, the molten-salt reactor, is
~being developed into a thermal lreeder. The MSRE, in which

molten fluoride salts circulate at 1200 deg F, is a step in
that development. Safety analyses and experience have -
shown the MSRE to be safe, and no serious prollems are
expected in designing safe molter-salt treeder :eactor

*accidents + *MSRE + *reviews ¢ *safety + afterheat +

analysis ¢+ containment + exper1ence

OTHER CATEGORIES: BGX

MDAG 70040
Kedl RJ + Hcutzeel A

Accession Number MDAGEOQOCGY to MLCA670040

 
 

 

Page 150

Category M
"MSRE

MDA670040 *Continueds
DEVELOPMENT OF A MODEL FCR CCFPUTING XE-135 MIGEATICN. IN
THE MSRE . ,

Oak Ridge Natiocnal Labcratory, Tenn. .

ORNL-4069 (June, 1967), 77 p, 22 £fiq, 3 tak, 21 ref. ,
The report deals rrimarily with developing a model for
computing the migration of Xe-13% in the MSBE ard wvith
experiments conducted tc establish the model., 2
preoperational experiment was run in the MSRE with Kr-85
tracer, and rany of the gas-transport constants vere '
inferred from the results. Equivalent transpcrt constants
for calculating the Ye-135 migration gave a poisoning of
about 1.4% without circulating Lkultkles and well belcw 1%
with bubbles. Prelirminary measurements made or the critical
reactor show xenon poisoning of C.3 to (C.4%., Sirce
physical measurements ccntirm that there are tutkles in the
system, the conclusion is drawn that the ccoputation todel,
the krypton experiment, and reactor operation agree,

*analysis + *mass transfer + *models + *MSRE + *xencn +

computer ccdes + diffusicn + graphite + krypton

OTHER CATEGORIES: BFX + 1IBA

MDAGT70041%

- Kedl RJ .

A MODEL FCR CCMPUTING THE MIGRATION OF VERY SHORT LIVED

NOBLE GASES INTO MSRE GRAPHITE

Cak Ridgye National Laboratory, Tenn,

ORNL-TM-1€610 (July 1967), 26 r, 4 fig, 1 tab, o ref,
A model descrikting the migraticn cf very shcrt-lived rcble
gases frcm the fuel salt to the graphite in the ESRE CORE
was developed. From the miqgration rate, the wcdel couafputes
(vith certain limitations) the daughter-product distritution
in graphite as a function of reactor cperaticral histcry.
Noble-gas daughter-frcduvct concentrations (EA-14C, CE-141,
SR~€9, and Y-91) vere measured in graghite sarmples resoved
fror the MSKE core after 780C ¥Whr of fpcwer cferaticn.
Concentraticns of these isotopes computed with this model
compare favorably with measured values

*analysis + ¥*experience + *fission products + #*graphite ¢

*mass transfer + *MSRE + computer codes + diffusicn +

kryptcn + gpclten salts + zxenon

CTHER CATEGCEHEIES: IEA

MDAG6BCO03
Haubenreich EFN + Enqel JR + Gaktard CH + Guywcn RH +

Prince BE
MSRE LCESIGN ANLC OFERATIONS REPORT PART V-A, SAFETY ANALYSIS
CF OPERATION WITH U-233
Cak Fidge Kational Laboratory, Tenn.
ORNL-TM-2111 (Feb. 1968), 80 g, 24 fiqg, 36 ref.
This report presents data and enalyses thkat suppcrt

Accessicn Number MDA670040 to MLA680003
e

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»)

Page 151

Category M
MSRE

MDA680003 *Continued*
the conclusicn that it is safe to load and operate the.

MSRE with uranium-233, It summarizes pertinert experiefnce
witt the MSRE and new information on materials through
December, 1967. -Procedures fcr producing, hardling
and lcading enrichirqg =alt are described and tkeir
safety assessced. The nuclear characteristics cf the
reactor with 0-233 fuel are presented and the possilbility
of breach of the primary containrent due tc credible
nuclear incidents is reexamined taking into acccunt
the different dynamics charactersitics, the acticn
of the safety system and the condition of the salt
system after two years of operaticn,
*analysis + *MSRE + *safety + *uranium-233 # containment +
dynamic characteristics + engineered sateguards % excursicns +
fuel fpreparation + Hastelloy N ¢+ lcading + radiation damage +
reactivity '
OTHER CATEGORIES: BGX

¥MDA690001

Steffy RC

INHERENT NEUTRCN SOURCE IN MSRE WITH CLEAN U—233 FUEL

Oak Ridge Naticnal Laboratory, Tenn.

CRNL-TM-2€8% (Aug. 10, 1969) 22p, 2 tig, 11 ref.
After about thriee years of nuclear ofperaticn, tte MSRE
fuel, enriched U-235, was replaced with a U-233 rfuel
mixture, Tn this new mixture there are guantities
of UV-232, U-233, and U-234., Each of these, along with
the U0-232 decay chain, is a strcng alpha eritter and
interacts with flucrine, beryllium, and lithium to
produce neutrons. - This heutrcn scurce is tire-defendent
because cf the buildup cf U-232 daughters, and at the
time of reaching criticality with the U-233 fuel, the
neutrcn source in the MSRE core was about 4 x 10 (8th)
neutron/sec, primarily from the reacticrns Be9 (Alpha,N)C 12
and F19 (Alpha,N)Naz2., Alpha—-N reactions with 11th1um
will produce less than 3 x 1046th) neutrons/sec,
Spontaneous fission W111 proauce less ttan IC(Zrd)

' ‘neutrcns/sec.,

*MSRE + *neutron soufces + *fuels + *u:an1um—¢33 +

*uranium-232 + decav + berylllum + fluorlne + llthxum +

calculatlons

HDAbQCOOZ
Steffy RC ® Wood £J '
THEORETICAL DYNAMIC ANALY‘IS CF THL narz WITH Uu- 233 EUEL
Cak Ridge National Laboratoty, Tenn. : .
ORNL~-TM-2571 (July, 1969) 42 ¢, 17 fig, 11 ref.
A study undertaken to characterize the dyndmlcs of the
U-233 fueled MSRE prior to operation revealed that
the system is inherently asymptotically statle at all

Accession Numkter MTAG680CC3 tc MLA690002

 
Fage 152

Category M
MSRE

HDA690002 ¥Continued* .
powcer levels above z2ero. The motivation for these
studies was the expected difference ketween thke MSRE
dynamic respcnse with U-233 fuel and with U-235 fuel
because of the smaller delayed-neutrcn fracticn cf
U-233. An existing syster model, previously verified
for 1-235 fuel, was modified for use in this ucrk,
The reactor system respcnse to reactivity perturtations
is rapid and nonoscillatory at high pover, and it beccmes
sluggish and oscillatory at lcwer powers. These
characteristics were determined Ly three methcd
(1) transient-respcnse analyses, including a check of |
the validity of the linear model, {2) a frequency-resicnse
and sensitivity study, (3) stability analyses, toth
by inspection of the system eigenvalues and afpglicaticn
of the recently develored, modified Kikhailov critericn.
*MSRE + *dynamic characteristics + *stalility ¢+ *gcdels +
*uranius-233 + fuels + nuclear analysis + reactivity +
delayed neutrons + computer codes + mrixing ¢+ calculaticns +
met hods |
CTHER CATEGCFIES: BCX

MDAG69C00E

Burke CW + Clark FH

ANALYSIS CF TRANSIENTS IN THE MSRE SYSTEM WITH URANIUK-233

- FUEIL

Oak Ridge Naticnal Laboratory, Tenn.

CRNL-4397 (June 1969), 47 p, 19 fig, U ref.
The uranius-233 fueled FSRE system was simulated on
the CFNL analog computer. The simulated syster was
used tc evaluate the existing MSRE ccntrcl and cafety
systers when used ¢n the uranium-233 fueled systenm,
The pertinent results and conclusions were as fclloss:
{1) The safety system will limit the *startup accident®
so that the peak power will ke 1CC kw. {2) AR quartity
of uranium-233 sufficient to cause a reactivity change
of approximately -1% delta K/K whken fprecipitated cut
of the fuel at scme pcint in the system external to
the core could be swept Lack intc the ccre in a ccncertrated
ferr without causing excessive core damage. {3) The
existing controller will control the uranius-2z33 fueled
syster in a stable smanner; however, an incredsed velocity
feedback gain will te reguired.

analysis + dynamic characteristics ¢+ computers + feedktack +

MSRE + analog systeas + stability + mathematics + models +

simulation + control + reactivity + safety ¢ startugp +

heat transtfer + ccntrcl rcds ¢ vranium-233 + accidents +

control-rod drives + delayed neutrons + dynamics tests +

excursicns + kinetic equations

MDAT700006

Accessicn Number MDAR690002 to MLA700006
»)

-l

-

Page 153

Category M
MSRE

MDA700006 *Continued*

Ulrich ®C

AN EXTENDED HYLCRAULIC MOLEL OF THE MSRE CIRCULATING FUEL

SYSTEM (THESIS)

Cak Ridge National Laboratory, Tenn.

ORNL-TM-3CC07 {(June 1970) 53 p, 6 fig, 14 ref.
The hydraulic portion of a comtined hydraulic-reutrcnic
matheratical mcdel for determining the effects of
helium gas entrained in the circulating fuel salt
of the MSRE on the neutron flux-to-pressure frequency
response was extended to include effects due tc the
fuel rumr and heliur ccver-gas system. By comparing
the computed results with experimental data, it wuas
ccncluded that rressure perturbations introduced ty
the fuel pump were the main scurce of tlte naturally
occurring neutron flux fluctuations in the freqguency
range of one to a few cycles per seccnd. 1t was also
noted that the arplitude of the neutron flux-to-pressure
frequency-response function was directly rropcrticnal
tc the pressure in the fuel-pump bowl; however, further
work will be required tefore conmrletely satisfactcry
results are obtained frcm the extended model.
Recommendations are proposed.whichk shculd prcve useful in
future mwedeling of similar hydraulic systens.

*HSRE + *models + *hydraulics ¢+ *dynamic charactericstics +

*measurement + reactivity + ccmputer codes + hydrodynamics +

primary system + nuclear analysis + cover gas + calculaticns

OTHER CATEGORIES: MDC

MDAT70G0007

Bell MJ .

CALCUIATEL RADIOACTIVITY CF MSRE FUEL SALT

Oak Ridge Naticnal Labora tory, Ienn,

CRNL-TM-2970 (May 1970) 21 p, 0 fig. 8 ref, 12z tables.
Calculations have been made of the inventory and
radiocoactivity of the fission. prcducts ard trarsvranius
isctopes present in the MSRE fuel 'salt. The calculations
included operation with koth U-23% and U(-233 fuels,

the effect of stripping of noble gases, and flucrination
of the fuel salt after the peczcd of U-235 cperation.
Results are presented which give the 1nventory and
radioactivity of individual isctores in the salt up
toc Januvary 1, 1975. after storage for 5 years, the
gamma-ray shielding requlred fcr shippirg fuel is
detersined by thalium-208 and neutrons from uranium=-232
daughters produce tbe controll1ng radiatiocn dcce throth -
'a lead shield.

*MSRE + *fission products ¢+ *Jsotopes + stcrage + fuelc +

*decay + uranium-235 + uranium-233 + uranium-232 + disposal +

processing + shielding + *inventories + rare cases ¢

radiocactivity + calculaticns

MDAT700032

Accessicn Number MDA700006 to MLA700032

 
 

Page 154

Category M
MSRE

MDA700032 *Continued*
Robkinson JC
ANALYTICAL DETERMINATION OF THE NEUTRON FLUX-TO-PRESSLRE
FREQUENCY: APPLICATION TC THE MCITES~-SRIT FEACTCE '
EXPERIMENT -
Oak Ridge Kational Laboratory, Tenn. :
Nucl. Sci. and Eng., 42(3), 382z-396 (December 197C), 15 g,
7 fig, 2 tables, 18 ref.
The neutron flux-to-pressure freguency respchcse for a
mol ten-salt-fueled reactor with a srall amcunt cf gas
entrained in the mclten csalt was determined analytically.
The one-dimensional conservation equaticns describing the
" flow cf the corxpressible molten-salt gas mixture.and the
one~-group neutron diffusion equations were written in the
linearized perturbed fcrm, and laplace transfcrmation in
time was performed. The coupled set cf equaticrs describing
the ccnservaticn cf mass for the molten salt, -ccnservetion
of mass for the gas, and conservation of mcrertum £cr the
salt-gas rixture (the hydraulic eguations) was solved Ly
enploying matrix exponential teckniques. The rerainitg
eguations were sclved by more conventional schemes. The
matrix exponential technijue was selected to cttain a
solution for the hydraulic equaticns cver the techniques
ncrrally erployed (nodal or mcdal) for stability studies in
boiling water systems Lecause the validity cf tke sclction
is independent cf the frequency of interest, and the total
number of simultaneous equaticns required tc ke sclved for
applxcatlcn cf boundary conditions (closing the flow loop)
is small. Results from the computed nectrcn flux-tc-
pressure frequency resgcnse for the molten-salt-fueled
reactor under study show that the shape of the mcdulus cof
the freguency respcnse fcr freguencies below 1 to 2
cycles/sec is independent of the void fracticr (vclume
fracticn cccupied by the yas), and the magnitude of tte
modulus of the frequency resgcnse is prcgorticnal tc the
void fracticn. Therefcre, e conclude that the amount of
void in the system can te inftferred ty ccrparirg the
analytical frequency response uwith an experimental
freguency response, {This conclusion vwas verxfled and is
repcrted in the follcwing paper,)
MSRE + bubbles + dynamic characteristics + ncise analytls +
void fracticns

MDAT71C003
Kerlin TW + Pall SJ + Steffy RC . '
THEORETICAL DYNAMICS ANALYSIS CF THE MCITEMN-SRIT FEFER2CTCE
EXFERIMENT '
Oak Ridge Naticnal Laboratcry, Tenn.
Nuclear Technology, Feb. 1971, 15 p, 24 fig, 12 ref.
The dynamic characteristics of the MSFE were calculated for
.operation with U-235 and U-23: fuels. 1Thbe aralysis

Accessicn Number MDA700032 to ¥LA7 10003
.

)y

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Page 155

Category M
MSRE

MDA710003 *Continuedx
included calculaticn cf the transient response for
reactivity perturbations, frequency resgonse fcr reactivity
perturbations, stability and sensitivity to parameter
variations, The calculations shocwed that the systern
dynamic behavicr is satisfactory for both fuel loadings.

MSRE + dynamic characteristics + feedback +

kinetic equations + sipulaticn + stability

MDB700003
Gabbard CH
REACTCR PCWFR MEASUREMENT ANL HEAT TRANSFER PERFORMANCE IN
THE MOLTEN SALT REACTOR EXFERIMENT
Gak Ridge National Laboratory, Tennessee :
ORNL-TM-3C02 (May 1970) 32 p, 6 fig, 16 ref.
The operating power of the MSRE as determired by a
heat ktalance on the fuel and coolant salt systems,
vas 8.0 MW. Changes in the isotopic composition of
uranium and plutonium in the fuel salt indicated a
power lower by about 7 - 10%. Attempts to resolve this
discrepancy have been 1inconclusive. The ccclant salt
flow rate was found to be the only potential source
of significant error in the heat talance. A calibration
check of the instruments is planned. The heat-removal
capabilities of the fuel-salt tc coolant-salt heat
exchanger and coolant-salt tc air radiator were Lelow
~the predictions of the original design calculaticns
In the case c¢f the frimary heat cexchanger, the cverestimate
was due to the use of erroneous, estimated phkysical
profperty data. 1In the case of the radiator, the
overestimate in the design was only partially exglained by
the imprcger selection c¢f an air ‘'film' temperature., There
was no decrease in heat transfer capakility cf the twc heat
exchangers cver more than 3 years of operation.
*MaRE + *heat balance + *heat transfer + ¥*heat exchangers +
molten salts + *perforrance + *cperatlon + reactors +
*experience + *flow measurement + isotores + analysis +
*therral pcwer + components + reviews + design + design datea +
physical properties + thermal propertles + 1n=trumertatxot +
spec1f1c beat + heat balance_“' L

flDB700033
Rokinson JC + Fry DN - '
EXP”RIBENTAL NEUTRON FLUX-TO-PRESSURE FRLQUENCY RECPONCE EOB-
THE MOLTEN-SALT REACTOR EXFERIMENT" BETEFPIhATICh CE VeIl
FRACTION IN FUEL SALT S
oak Ridge National Lahoratory, Tenn, .
Nucl. Sci. and Enqg. 42(3), 397-405 {Dec. 197C), S , 1C fig,
1 table, 14 ref. |
Small pressure perturkations were intrcduced intc the
prirary fuel pump bowl of the MSHE ofperating at its nc<minal
pcwer of 8 MW{th). The experimental neutron flux-to-

Accession Number MDA7100C3 to MLB700033

 

 
 

Page 156

Categjory M
MSRE

MDB700033 *Contlnued*
pressure frequency Iesfcnse uas then obtained from a cross-
power and auto-power spectral density aralysis cf the
resulting sigpals frcom a neutron-sensitive ionization

- chamber and a pressure transducer. By comgaring the

frequency derendence of the experimental frequency response
determined for the reactor operating at power with the
frequency response determined from analysis of mathemeétical
podels, the selection of the mcre approfriate bcupdary
ccnditicn set from a chcice of two possible boundary
condition sets was possikle. Then the analytical freguency
resgcnse was fitted by the least-squares methcd to the
experimental frequency response tc oktain the vcid fraction
in the mclten salt fuvel. A void fraction of 0.€1 plus or
minus 0.,04% was determined frcm the frecuency resgonse; this
value corpares favcrably with a value of 0.6 plus or minus
0.1% determined by cther techniques. Conclusions are that
the analytical model leads to acceptable results for the
neutrcn flux-tc-iressvre frequency respcnse and that
properly designed dynamic tests involving srall reactivity
perturbaticns (intrcduced by means other than rcd motion)
can be used to extract specific nuclear pararceters for a
nuclear system operation at poser.

MSRE 4 bubbles + dynamic characteristics + dyramics tects +

measurement + noicse anmalysis + void fractions

CTHER CATEGCRIES: MCL

MDB71C002

Kerlin TW + PRall SJ + Steffy RC + Buckner MR

EXPERIENCES WITH DYNAMIC TESTING METHODS AT TEE MCITEN-SALT

REACTCR EXPERIMENT _

Oak Ridge National Laboratory, Tenn.

Nuclear Technology, Feb. 1971, 15 p, 1S fig, 13 retf.
A series of reactivity-to-power frequency resronse
measurements was made on the Mclten-Salt Reactcr Exgeriment,
This was done for U-233 and U-235 fuels, for a range cf
operating power levels, at several gcints in the systenm
operating histcry, and for several different test
procedures., A comparison of experimental results with |
pricr thecretical predictions confirmed the validity cf the
theoretical predictions. The test frogram ircluded
measurements using the f(seudorandom binary sequence,
pseudorandom ternary sequence, n~sequence, and the
myltifrequency binary sequence.

MSRE + dynamic characteristics ¢ dynamlcs tests + £eedback +

noise analysis ¢+ stability + experience

CTHER CATEGCRIES: MCD

MDC6EC002
Rerlin TW + PRall SJ
EXPERIMENTAL DYNAMIC ANALYSIS OF THE ECITE&-SAIT BE!CTCE

Accession Numher ¥DB7C0033 to MLC6€ECCC(C2
"

*

Page 157

Category M
MSREF

MDC66C002 *cntinued*
EXPERIMERNT

Oak Ridge Naticnal Labcratory, Tenn.

ORNL-TH-1647 (Cct. 13, 1966), 58 p, 29 fiq, 2S¢ ref.
Dynamics tests were performed on the uranium-23S fueled
MSRE for the full range of operating poxer levels tc
determine the fcwer-to-reactivity frequency response.
Three types of input disturlances were used: thke psetudc-
randcr binary reactivity input, the pulse reactivity
input, and the step reactivity input. 7The fregquerncy
resgcnse of the unccentrclled reactor system displayed
resonant behavior in which the frequency of cscillaticn
and the dacring increased with increasing powvwer level.
Measured periods of natural oscillaticn ranged frcm
thirty minutes at 75 K% to two minutes at 7.5 MW. These
oscillations were lightly damped at lcw power, tut sticnyly
damped at higher pcwer. The measured results generally
vere in good aqreement with predicticns, 1The rair
conclusicn is that the system has no operational stability
problems and that the dyndmlc charactericstics are
essentially as predicted.

*MSRE ¢+ *dynamics tests + *dynamic characteristics +

*experiment + measurement + #¥ctability + testing +

reactivity + thermal power + procedures + analysis + wethcds

OTHER CATEGORIES: BCYX

MDC670001
Engel JR + Prince BE -
THE REACTIVITY EALANCE IN THE MSRE
Oak Ridge Naticnal Labeoratcry, Tennessee
CRNL~TM-1796 (Far. 10, 1967) 54 p, 1€ fig, 1S ref.
Expcrience with a reactivity balance calculation is
described for approximately 1 year of MSRE pcwer cperatiocn
with uraniup-235 fuel, Conputations were performed every
5 minutes by an on-line digital ccmputet. Results
~were used 1n1t1a11v toc evaluvate xenon poisoning in
, the reactor and &uhseguently tc mcnitcr for arcralous
reactivity changes. Sen=1t1v1ty for detecting short-
term changes in fuel compos;t1cn is 1C times greater
than chemical analy=1 ‘No significant long-ternm |
drift in reactivity is otserved ‘at zerc power with
nc xencn present., A more detailed and comprehensive
report of the theoretical Lkas€ and the entire experierce
wittk uranium-235 fuel 1= precented in CFN1-4674, accession
- number MLC710006. - ' -
¥*MSRE + *reactivity + *experxence + txenon +

*nuclear analysis + operation + data prcces=1ng'+ bubkles +

contrcl rcds + rare earths + modelc + fission products +
fuels + uranium-235
OTHER CATEGORIES: BBX

¥DC670002

Accessicn Nupber MDC660002 to MIC670002

 
 

Page 158

Category M
MSRE

MDC670002 *Continuedx
Engel JR ¢+ Prince BE - |
THE REACTIVITY BALANCE IN THE MSRE
Oak Ridge National Laboratory, Tenn.
Abstract of paper presented at the Thirteenth Annual Meeting
of the Agzerican Nuclear Society, San Liego, Calif.,
June 11-15, 1967, REmerican Nuclear Society Tramsacticns
109, g 337
Experience with a reactivity talance calculat;cr is
described for apprcximately 1 year of MSKE power operation
with uranium-235 fuel., This alstract {and preprirnt
of Faper) is a syncpclc of OBNI-TFK-179€, same title,
accession number MLC670001. ,
¥*MSRE ¢ *reactivity + *experience ¢+ #xenon ¢+
*nuclear analysis + operation + data processing + Lbuktlles +
control rods + rare earths + models + fissicn prcducts +
fuels + uranium-235

MDC680002 : _ |

Prince BE + Ball SJ + Engel JF + Haubenreich EN + Kerlin TW®

ZERC-FOWEF FHYSICS EXPERIMEHTS ON TFE MOLTEN—-SALT REACIOR

EXPERIMENT

Cak Ridge National Laboratory, Tenn. -

ORNL- 4233 (Feb. 1968), 60 g, 24 fig, 25 ref.
This report descrikes thke techniques ané results cf
a prcqgraz of experirents designed to measure the important
neutronic characteristics of the MSRE, under ccrditicrs
of neqgligible nuclear heat generation. The program
includes the initial critical U-235 loading, the conticl-
rcd calibraticn (period-differential worth and rod
drop-integral worth measurements), determinaticre of :
the reactivity loss duve to fuel circulation, the 'static!
reactivity coefficients of excess U-223% concertraticn and
isothermal core temrerature, the fuel salt temperature
reactivity coefficient, the pressure effects cn reactivity,
and a series of system dynamics tests (frequency respcnse,
transient flow, and neutron flux noise measurerents).
These measurerments, carried out in June 1965, form much
Of the experimental base¢line for evaluation ct the
nuclear operation at full power with U-235 fuel. The
repcrt includes discussions of the comparisons of the
measurement results wvith the corresponding neutrcrnic
characteristics calculated from theoretical models.

*analysis + #*criticality ¢+ *dynamics tests + *exrerimert +

*MSRE ¢ *reactivity ¢+ *uranium-2z35 + circulation + .

control rods + delayed neutrons + experience + mcdels +

‘startup ,

CTHER CATEGCRIES: KAB + MCD

MDC68CO05:
Fry DX + Kryter EC + Robinson JC

Accessicn Numbet MDC670002 to HrCGBCCCS

 
-

»

Page 159

-Categdry M
MSRE

MDC680005 «Continued*

MEASUREMENT OF HELIUM VOID FRACTICN IN THE MSFE FUEL SALT

USING NEUTECN-NOISE ANALYSIS

Oak Ridge Naticnal Laboratory, Tenn.

CRNL-TM-2315 (2ug. 1968), 32 p, 10 fiqg, 17 ref.
Investigaticns were made at the FSRE during pover
operation with uranium-235 fuel tc determine if the
amount of helium gas in the fuel salt cculd Le measured
using peutrcn ncoise analysis. The neutron power
spectral density (NPSL) was measured at differert

. reacter coperating ccnditions and compared with analytical
model predictions of the NPSD tor the same ccrditiors.
Results showed that the principal source of small

- neutron density fluctuation cliserved in the MESRE
is heliur bubbles circulating in the fuel salt. The

- measurements showed that NPSLC in the frequency range
froom 0.5 to 2 cps varied as the sguare of heliunm
void fraction as predicted Ly the model, and that:
the rpinimum void fracticn was more nearly zero than
the previously accepted value cf C.1%. It is ccncluded
that changes in the circulating void fraction can be
inferred with good sensitivity directly frce reutron
'noise measurements, and, concequently, MESL can complement
and enhance the value of the MSRE reactivity fkalance
calculaticns -

*anal ysis + *experlence + *pnodels + *MSRE + *pcise analysis ¢

*reactivity + *void fracticns + dynamlc characteristics +

experiment + uranium-235 '

OTHER CATEGORIES: MCD

MDC690003
Robinson JC + Fry DN
DETERMINATION OF THE VOID FRACTICN IN THE KSEF USIKC SEALL
INDUCED EFESSURE PERTUREAT IONS
Oak Ridge National Labcratory, Tenn,
ORNL-TM-2318 (Feb. 1969), 58 p, 11 fig, 22 ref. -
. With the MSRE cperating at 5 Fu, sawtooth precsure
‘perturbations were introduced intc thke fuel-purp bowl
tc determrine the arcunt of helium gas ertrained in
. the circulating fuel. The pressure and neutron flux
'~ 'signals were simultaneously anplified and reccrded
- on magnetic tape. Then the signals were analyzed
using auto-power spectral density, crcss-pcwer spectral
density, cross-correlation, and direct Fourier transfcrm
‘techniques to oktain the neutrcn-flux-tc-pressuie
frequency-respcnse function. An analytical model,
developed previously to aid in the 1ntexpretat1cr
“of the fluctuaticns of the neutron flux in an unperturted
system, was used to infer from the experimental data
the amcunt of heliur veoid (interpreted as a void fraction)
entrained in the fuel salt, A descripticn cf the aralytical

Accessicn Number MDC680005 to MIC690003

 
 

 

Page 160

Category M
MSRE

BDC690003 *Continued*
nodel and its experlmental verification are included
in this report. The void fraction was deteru:ned tc
be betveen 0.023 and 0.045%. The uncertainty of this
inference is attrituted to assumpticns sade ir the
model. (AbctractOI'c ncte: This work was sulsequently
reported in Nucl. Sci. & Tech.,, se€e MDB700CCS.)

*analysis + *dynamices tests + %experiment + *podels ¢

*MSRE + #*reactivity + *void fractions +

dynamic characteristics + experience + uranium-235 ¢

pressure + theory

OTHER CATEGORIES: MCD

MDC690015

Fry DN + Kryter RC + Robinson JC

ANALYSIS CF NEUTEKCN NOISE IN A MOLTEN SALT RERACTOR

OPERATING AT PORER
Cak Ridge National Laboratory, Tenn.
Summary of fpaper presented at ANS Annual Meeting, Seattle,
Wash., June 15-19, 1969, ANS Transactioms 12¢1), 299, 2 p,

1 fig, 5 ref. .
Neutron flux noise in the MSRE was studied by Fourier
analysis of an ionization chamlker signal over the frecuency
range from 0.1 to 15 cyclessec. Peasurements at varicus
operating conditions and cross correlaticns vith cthet
reactor signals shcved marked changes with changes in the
circulating void fraction and a high decree ct ccrrelaticn
between neutrcn flux noise and pressure noise, It apgears
that cross correlation of neutrcn and fpressure rncise crfers
a ncn~-perturbing method of determining the circulating void
fraction with the reactor operatirqg at power. (This scrk
is described in detail in CRNI-TK-2315. See KLCEE000E.)

*MSRE + #*noise analysis + *void fractions + measuremert +

neutron flux + heliumr ¢ ccver gas

CTHER CATEGCFTES: MCLC

MDCHICO16
Rohinson JC + Fry LN
THE FREQUENCY RESPONSE OF IHE NEUTRCN FIUX TC ERESSUFE IN A
CIRCULATIKG FUEL REACTOR - ANALYTICAL AND EXPERIMENTAL
0ak Ridge Naticpal Laboratory, Tenn.
summary of paper presented at ANS Annual Meeting, Seattle,
Wash., June 15-19 1969, ANS Transactions 12 (1) 2¢2, 2 p,
3 ref.
An analytical rodel was develcped to cogpute the neutron-
flux-to-fressurte frequency response in a reactor fuelled
with circulating molten salt. Since entrained ccver gas
makes the circulating fluid compressible, pressure
perturbations induce reactivity, and hkerce neutrcn flcx,
perturbations. Analysis indicated that the neutron-flux-to-
pressure frequency response is gprcporticnal tc the

Accessicn Number MDC690003 to KIC6900 16

i

i 1"
»}

-

"

Page 161

Category M
MSRE

MDC690016 *Continued*
circulating void fracticn in the frequency range from 0.01
to 0.1 cycle/sec. Sawtooth pressure petturbaticns weie
impcsed on the MSRE at full power and the fregquency
response vas measured, Experimental data were best fitted
with a circulating void fraction of 0.04%. (2 detailed
discussion of the model and experimental results is

' presented in ORNL-TE-2318., See MDC-690003.)

*HSRE + *dynamic characteristics + *lkullles + *mcdels +

*nyclear analysis ¢ experiment + measurement +

*void fractions + cover gas + heliur + calculaticns

OTHER CATEGORIES: MCD

MDC700004
Steffy RC
FREQUENCY RESPCNSE TESTING OF THE MOLTEN SALT REACIOR
EXPERIMENT (THESIS)
Cak Ridge National Laboratory, Tenn, |
ORNL-TM~-2823 (Mar. 1970) 118 p, 27 £figy, 31 ref.
Tests to determine the neutron flux-tc-reactivity frecuency
resfonse were rerfcrmed on the MSEE with the reactor
at various power levels between zero and full pcver ard
with the reactcr fueled with a 0~235 fuel mixture and
a U1-233 fuel mixture. Test patterns ergployed were pseudo-
randcr binary sequences (PRBS) and pseudorandom ternary
sequences (PETS5) of various sequence lengths ard mirnisum-
pulse-duration times. 1In some tests reactivity (control-
rod position) was forced to fcllow the test pattern,
ard in other tests the neutron flux was forced tc fcllow
the test pattern. The experimental results were analyzed
by several different methods and the results were ccmjared.
The frequency respcnse c¢f the uncontrolled reactor systenm
was found to be in good agreement with thecretical
predicticns for both the U-235 and U-233 fuel lcadings.
. There were no indications of respcnse ctaracteristics that
* " migkt cause ccntrol orfcafety problems. Advantages and
'dlsadvantages of varxous test1ng and analytical xethods are
discussed. _
*MSRE + *dynamic character15t1ca + ¥peasurenerct + *experlme:t +
*testing + reactivity + ccmputer codes + data process;ng +
nuclear analysis + *analysis + contrcl rods +
contrcol-rcd drives + computerc + data acgu151t10n aystems +
nmethods . o e
OTHER CATEGORIES: MCD

¥DC700005

Steffy RC

EXPERIHENTAI DYNAHIC ANALYSIS OF THE MSRE WITH U-233 FUEL

0ak Ridge Naticnal Laboratecry, Tenn.

CRNL-THM-2997 (April 1970), 28 p, 10 fig, 11 ref.
During the startup with U-232 fuel, tests shcwed that the
syster time resgcnse to stop changes in reactivity, the

Accession Number MDC690016 toc MEC700005

 
 

 

 

Page 162

Category M
MSRE

DC700005 *Continued*

flux-to-reactivity frequency respoase, and the outlet
terperature-to-pover frequency response agreed favorably
with theoretical predictions. Time-resgonse tests 1, 5, and
8 Mw verified the prediction that, althcughk after a
perturbation the reactor returned to its original povwer
level more rapidly when the initial pcwer was high than vhen
it was low, the system was load-following at all significant
power levels, Flux-to-reactivity frequency resgcmpse uas
effectively measured using periodic pseudorandom kinary and
ternary sequences. . As predicted, for tke MSRE, the degree
of stability increased with increasing power level.

*anal ysis + *dynamic characteristics + *dynamics tests +

*experiment + *MSRE + *reactivity ¢+ *stablllty + experience ¢

nodels + uranium-233 + methods

OTHER CATEGORIES: MNMCD

MEAGU40005

Blumberg R

REMOTE MAINTENANCE OF THBE MSRE {(PRRT OF MSRP SEMIANN FROG

REPT 7/31,/€4)

Cak Ridge Kational lLaboratory, Tenn.

ORNL-3708 (Nov. 1964), pp. 190-200, £ fig, ¢ ref.
Maintainability was a primary coasideration in the
design and planning of the MSRE. Cormpoctents that
will beccme radicactive were designed and located
so that they can be disconnected Lty the use cf lcng-
handled tooles inserted through a work shield set up
on top of the containment cell. Large itees %xill be
disconnected this way, then will be removed by a crane
operated from a shielded contrcl rooe. The ucrk shield,
tools, remcte vievwirg equipment, and procedures have '
been developed and tested.

*maintenance + ¥*MSRE + *plans + design + development ¢+

equipment + procedures + shielding ¢+ tcols

OTHER CATEGORIES: MEB + FKBA '

MEB660028

Blumberg R -

MAINTENANCE CF RALICACTIVE SYSTEMS AND COHPONENTS A1 THE NSEE

Oak Ridge Naticnal Labcratory, Tenn.

ANS Trans 9(2) (1966), p. 530.
Maintenance operations are performed at MSEE with lonc tools
manipulated through access holes provided in a ;crtab]e
shield., Experience has heen good. (A later, mcre de tailed
reference on this subject is KEC700053,)

experience + MSRE + procedures + remote maintenance + tcols

OTHER CATEGORIES: KBA

MEB680001
Blumkerg R + Hise EC .
MSRE CESIGK ANL CPERATIONS REPORT, PART X —— MAINTENANCE

Accessicn Number MDC700005 to MER680001

i
")

.

*maintenance + *MSRE + *plans + desiqgn + development

Page 163

Category M
MSRE

MEB680001  *Continued*

EQUIPMENT AND PROCEDURES

Cak Ridge National Laboratory, Tenn, |

ORNL-TM~910 (June 1968), 80 p, 26 tig, 3 ref.
A record of the methods develofred for maintairing
the radicactive pecrticns of the MSRE is presented.
The maintenance system utilizes long-handled tccls
operated through a mrovable shield for most of the
in-cell manipulations. For scme radicactive tranpsier
and cetu; tasks that cannct be handled ctherwise, a crane
that is operated remotely fror a shielded ccntrcl rcorn
"is used. Overall descriptions are given of the components
and the methods of wmaintenance, Some detailed frecedtres
written freoer the standgcint of the people who perform
the vork, are also presented. Reference material that
will be useful when detailed informationm is required
is included.

equiprent + procedures + shielding + tocls

MEC70C0%3

Haubenreich EN + Blumberg R + Richardson N

MAINTENANCE OF THE MOLTEN-SALT REACICB EXFEFI MENT

Cak Ridge National Lahoratory, Tenn, .

Paper, ANS 197( Winter Meeting, Hashlngton, NOV, 1970,

29 p, 8 fig, 8 ref.

The MSRE was designed fcr malntenance of radicactive
systems by simple tools inserted througt a pcrtable
shield. This system proved practical for radioactive
maintenance and inspection jols arising in & years cf MSRE
operation, Delays in the program due tc maintenance were
not excessive and activity releases and ferscrnel
exposures were wminiwal. This paper, given at a special
session on Maintenance of Radicactive Systems describes
the MSRE design for maintenance, lists jobs done, and
discusses the experience, Copies are available from ESRP
Director's Office, GRNL; a- summary is in Ahb Trans.
Vol, 13, No. 2, p. 789. - , '

*design + *experience + *MSRE + *remote malntenance +

*tools + contamination + examxnatlons + equiptent +

healtt phy51cs + malntenance + performance + rellabxllty +

shielding

OTHLB CATEGORTES: KBA + EEAr -

HFX?OOOZO '
Haukenreich PN + R1chardscn M
PLANS FCR FCST~CEFERATION EXAHINATION OF THE MOLTEN-‘AIi REACTIOCR
-EXPERIMENT _
Cak Ridge National Laboratory, Tenn.
ORNL-TM-2974 (April 1970), 30 p, O £fig, 3 ref.
In December 1969, after more than 4 successful years, the

Accessicn Number MEB6800C1T1 to MFX7CCCZC

 
 

Page 164

Category ¥
MSRE

MFX70€020 *Continued%

nuclear operation of the MSRE was concluded and the plant
vas placed in standby. Work planned for early im FY- 1971
includes removal of some core graphite; viewirg inside the
reactor vessel and inside the fuel-pump bowl, inspection of
portions of the salt piping, the ocffgas charccal bed, the
coolant salt pump, and the control rods; and testing the
coolant salt flowmeter. Each study is justified by 1its
benefit to the Molten-Salt Reactor Program. [Frocedures and
tools are available for some jcks; fcr cthers, they are
currently being develoged. - )
decommissioning + examinations + MSRE + plans + reacte raintenance

Accession Numker MFX700020 té EFX700020
.}

Page'165

Category N
Miscellaneous

NXX590002

{(Staff Rerort) a , :

REPORT CF THE FLUID-FUEL REACTOBRS TASK FORCE

United States Atoric Emergy Ccmmission LRDT

TID-8507 (Feb. 1959), 188 p, 7 fig,
A critical evaluaticn was made of the 3 flflld fuel concepts
under development Lty the USAEC: aquecus hcocgerecus,

- molten-salt, and ligquid-metal-fuel. The task force
concluded that all 3 could breed in the thorium-U-233
cycle, with the AHR having the greatest potential gair.

- Maintepance was ideptified as the most important factor
influencing the practicability cf ary of the thiee. The
molten-salt reactor was judged to have the highest
probability of achieving technical feasibility.

*AEC + *development + *plans + *reactors +

breed1ng performance + fue€ls + LMR + mpaintenance + MSER +
optimizaticns + reviews

CTHER CATEGCEIES: AEX

NXX63C001
Voznick HE + Uhl VW
MOLTEN SALT FOR HEAT TRANCFEB \ _
Atlantic Fesearch Corp + Irexel Institute of Technclcgy
Chem. Engrg. Vcl. 70, 129 (May 27, 1963) 8 p, 4 fig, 37 ref.
Heat-transfer salt (HIS) composed of 40% scdivur ritrite,
7% scdiur nitrate and 53% potassium nitrate has been used
- widely since 1937 fcr heating and cooling in the petrcleum
and chemical industries. HTS is inexpensive, has gcod heat
transfer prcperties, has a very low vapcr pressure ané is
non-toxic. The free21ng point of dry HIS is 26(C deg F, and
is defressed by water in the =alt. HTS is not highly
reactive with air, tut a rlanket cf steam cr inert gas 1is
reccerended., Carbcn steel is satisfactcry tc €€C deg F,
- with stainless steel recommended for applications to
1100 deqg F. HTS is commercially availatle frcr several
- sources and at least twc firms manufacture complete heat
- transfer systems using this salt.' Pumps up tc 17,0C0 ggnm,
valves, and steam generatcrs up to 20 M% are in- ser71ce. '
*coolants + *molten salts ¢+ *heat exchangers +
*heat transfer + *heat t:eatments + *nitrates +

*physical prcperties ¢ *cafety + %*secondary salts +

*steam generators + *thermal’ conductJV1ty +
accidents + aprlicaticns + - -
bearings + behavior + compat1t111ty + ccmponents +
containers + ccrrcsicon + cover gas + density +
experience + failures + flanges + fluids +

freezing + inert gases '+ liquidus + materials +
melting + NaK ¢ piping + fumps ¢ reviews +
specific heat + stability + thermal insulation +
tubing ¢+ valves + vapor pressure + visccsity

NXX69C046 -

Accession Numker NXXS900C: to'NXXGQCCHé

 
 

Page 166

Category N
Miscellaneous

NXX69C0U46 *Continued*

{Staff Eeport)

THE USE OF THORIUM IN NUCLEAR PFCWEF KEACTORS

United States Atomic Enerqgy Commission DRDT

WASH-1097 (June 1969), 144 p, 28 fig, €% ref. -
This repcrt identifies the factors involved in thoriun
utilization and descrites the status as of mid-1568. It
was prepared under the direction of UFDT1 by a task force
from industry, national laboratories ané the RE(, and
contains a fcreword by ERDT DiCfector M. Shaw. 1he rejocrt
treats first the geperal features of the thorium cycle
(resources, nuclear characteristics of thoriur and U- <33
in thermal- and fast-neutron spectra). Then it discusses
thorium utilization in specific reacter types: high-
temrerature gas-cocled, mclten-salt, licht-vater, and
heavy-water thermal reactors and a fast reactor, The MSER
considered is a single-fluid treeder with reductive
extracticn prccessing fcr Pa and rare earths. No
conclusion as to the relative Eerits cf varicus reactcr
types is explicitly rresented.

reviews + thorium + uranium-233 + treeding perforrance +

natural resources + develcrment + MSER

CTHER CATEGCRIES: AEX

NXX69C0E7

{Staff Eeport) ‘

COST-BENEFIT ANALYSTS OF THE U.S. BEEECER FEACTCEF EFCCFAM

"United States RAtomic Energy Commission DRD1T

WASH- 1126 (Apr. 1969), 98 g, 8 fig.
This report weighs the quantifialle tkenefits cf breeder

- reactcecrs against the ccsts incurred by the government in

their developrent. A model of the U.S, electrical foser
econcmy was used tc comfpare cases without a breeder and
with the LMFER plus converters. Large lenefit/ccst ratios
for LMFBR develcpmert were found in all cases.
Development of a parallel Lbreeder was also considered and
appeared desirable under most sets of assumptions. Tte
light-vater breeder, the molten-salt kreeder and the
gas—ccoled fast breeder are mentioned as candidates fcr
development.

*economics + *electrical fower + *optimizations +

*reactors + AEC + LMFER + MSER + power costs

NXX70C011

Bond VP

EVALUATICN CF FOTENTIAL HRZARIS FRCM TRITIUM WATER

Brookkaven National labocratory, N.1Y. -

Paper TAEA S¥ 146/13, IAEA Symposium on Envircnmental Aspects of

Naclear pcwer Staticns, New York, Aug. 10-14, 1970, 21 p, 44 ref,

This paper analyses possitle tiolcgical effects cf tritium,
revieving frcecm thecretical and experimental standpoints all

Accession Number NXX6900846 to XNXX700011

in

(1
«b

Paye 167

Category N
Miscellaneous

NXX700011 *Continued*
factors involved. Factors conceivably increasing toxicity
are considered in detail. These include selective
concentration {actually discrimination) in the human tody,
and possible effects due to incorporaticn intc¢ nclecules
such as DNA. Ccnclusicns are: these factors dc not
significantly increase the dose expected frcs tritium in the
envircnment cr effects c¢f that dose; a dose of radiation
from tritium has the same radiokiological pearirg as the
same dcse of x-rays; the ICRP-AEC rax permissible body
burden of 1000 microcuries is quite ccnservative; the MPC
fcr water is ccnservative by a large factor. Anticipated
population exposure from reactcr-groduced HTO is very small
compared to that frcm existing HTC and other sources cf
radiation,

*health physics + *tritiuc + beta decay + concentration +

environment + reactors + reviews + safety + wastes

NXX70C0&87

Deonigi DE

A STMULATION OF THE UNITEL STATES ECHWEEF ECCNCPY

Facific Northwest Latoratory, Washingtcn

Proc. American Power Conf., Vcl. 32 (1970), pp. 105-115,

10 figq, 3 ref.

A ccrprehensive sipulation of a typical U.S. utility system
was made and the optimal qrowth pattern, usirc fcssil-fuel
plants and varicus reactor types, was calculated using
projected availability dates, fuel utilizaticr jerfcrrance,
and ccsts estirated in 1967. 1In the case vhere all reactor
types were allowed, ry the year 2(10 cver bhalf cf the
capacity was in mclten-salt converter reactors using e€xcess
plutoniun from fast treeders.

*electrical power * *econcmics + *natural resources +

*reactors + *optlmlzatlons + *systems + converters +

pover costs + sirulaticn

CTHER CATEGCRIES: BEFX

NXX70C058

(Staff FReport)

POTENTIAL NUCLEAR PORKER GBO&IH PATTEENS

United States Atomic Energy Commissicn DRDIT

HASH—1098 (Dec. 1970) 249 g, 33 fig. o
This report descrites the develo;ment and dpp11cat1cn of a
mcdel of the U.S. electrical power econcmy by the Systems
Analysis Task Force, whose activities centered at the
Pacific Northwest laboratories. 1Input data, includincg
costs appropriate for 1967, were provided by cther tacsk
forces established by DEDT to evaluate various reactor
concepts., Most consideration was given tc cctbinaticrs of
a few reactcr types considered most likely to be developed
in the U. S. 1In the one case in which thke ccrngetitior
included all reactcr types, over half of the reactors built

Accession Numker NXX7C0011 to NXX7C(CCSS8

 
 

Page 168

Category N
Miscellaneous

NXX70C058 *Ccntinued* '
after the year 2000 were plutcnium-fueled molten-salt
converters (Fig. 6.9).

*economics + *electrical power + *optimizaticns +

. *reactors + AEC + carpital costs + converters + fuel cycle +

fuel cycle costs + LMFER + MSER + natural rescurces +
pover costs :

NXX700060

(Staff Rercrt)

FEPOBT CF THE EFI REACTOR ASSESSMENT PANEL ,

Edison Electric Institute, N.Y. _

EEI Publication 70-30 (April 1970), 53 p, 4 fig, 7 ref.
A panel cf 6 utility executives (plus a working group)
reviewed power reactor developments and sucgested the
direction, pricrities, financial requirements, and timing
of utility involvement. Major emphasis cn the LMFBR is
indicated. With regard tc molten-salt reactors, the fpanel
concludes that they promise low costs it the future, Lut
the current lack of a supplier and the small scale of AEC
development deter utility involverxent,

*development + *electrical pcwer + #*reactors ¢ *utilities +

costs + energy ¢+ industry + FTGR + LMFBR + LWBR + MSBER +

natural rescurces ¢ fpower ccsts + processing + reviews

Accession Number NXX7CCOSE to NYXX7CCCEC

o
-l

169

KEYWORD INDEX

The reviewers who prepared the abstracts for MSRIS had a list of about
600 keywords from which to select a set for each abstract. In the pages
which follow are listed the keywords that were actually used, each followed
by the identification numbers of the abstracts-to which that keyword was
assigned. '_
The user bf-this‘index shpuld be aware of a‘péculiarity of the listing:
all keywords beginning with a capital letter (Hastelloy N, for example) are
listed after all the other keywprds. This is a consequence of the prepa-
ration and print;out of the index by‘the‘computer.

 
9

-

o

Keyword Index

Page 171

The Follcwing Index States The Key Term Ané Gives Referetces

Tc Each Article Which Was Keyed To It

absorbers
MAXEECO019
absorption
ACB690029
ACCEEC00E
BAX700008
accidents
AAX670010
CAX€SC(C53
IBA710005
LIX670013
MDAELCCOT
actinides
- ACDEEQC17
ACD670020
ACD670027
" ACDEBCOD15
ACD680016
ACDESCO2¢4
administration
- MCB71C012
adsorption
HIX660026
afterheat
- BGX670045
CAX€9CC53
C1X700010
IAC70CC47
aging
- PCCT7100C1C

alloy composition

ACE660018
ACEE7C02E
ACE680017
ACE€8GC02¢4

 'ACE690026

ACETCC025

ACE700039

. PAX€9C(3S

FBC610001

FBCESCO17

.- . FBE69004 ¢

allo¥s B

~ ACE650008

- ACEE€ECO014

-~ ACE€€CC12

"ACE660018

- ACE€7C028
aluminue

ACE€€CO1E

-analog systens

JAB690018
JAB70C017
JAB710008

LGX650002
LKX620003

MDA660004
MDA 670038

-MPA690005

NXX630001

ACD700038
CAX680032
CAX69005:
CAX710023
CXX70004¢
LDAT 00046

MAD69000U

IAC7 10013
IBAT710005
MCA670038

FBE7 10017
FBE710018

FCC700040

FCC700044

- ECC710010°
FCE690043
FCE7 10004

FCX690033

 FCX700026

FXX650047

 GXX680039 -
. ACE680017
ACE6800214
FAX620004

FAX632000€
~ FAX690035

" ACE670021

JCX690019
MCC660005

MCA690005

argon

analysis
AAX6700 10
ACA660008
ACAG€CO14
ACR6700 16
ACA€7C023
ACA680012
ACAEECC1S
ACA€S0021
ACA690028
ACAT7(0021
ACA700035
ACB7 10029
ACL660017
ACD670019
ACD677C02C
ACL670026
ACD€7C027
ACL700038
BGX67CO4S
CCX670035
CLX700010
- ECX710011
FCL7100 16
HCX71C0z2

analytical chemistry

ACD6E5CO11
ACD660011
ACDEECQ17
ACL670019
ACD670026
ACC6C027
applications
CAX€ECCOES

architect-engineering

ADXESCO€3

MCD6SC017
barium

ACA680012
 ACBETCO024
ACB6EQO13
tehaviecr =
ACD6€CO17

ACL670026
ACDEEQ(OZ3
CAX690053
CDX670035

beryllium
ACC660016
ACD6ECO15S

| ACD7(C024
~ bearings !
- ACA670023 .

IABE70043
IBD €80036
JAB630018
JAB700017
JCX690019
MCD€90062
MDA €20001
MLA640007
MLCA€50001
MDA €60003
MILAEG0004
MDA €70038
MLA670040
MDA €70041
MDA €80003

"MLCA€90005

MDB 700003
MLCE60002
MDC 680002
MLCE€80005
MDC €90003
MDC 700004
MILC700005%

ACD680022
ACD€90024
ACD69G0O31
ACD700024
ACD700038

NXX€30001

‘ACE€80024

ACE€90026
ACE700025
NXX€30001

- CXX700049
JAB€90018

JAB700017
NXX€30001

ACDEB0022

ACD€30031

 
 

 

Page 172

berylliunm

*Continued*
ACD70003€
LYXX660031

teryllium fluoride

ACD660017
ACDETCO1S
ACD670020
ACDET7CO2¢€
ACD670027
ACDEBCO1S
ACD680022
ACDE€9C024
ACD690031
ACD7CCO03¢
CAX680032
keryllium oxide
ACD690024
Freta decay
NXX70C011
kismuth
ACCEEQCTIE
ACC670025
ACC€9C023
ACC690030
ACC70C023
ACC700037
ACDE€7C01%
ACD670026
ACDEBCO1E
ACD680022
ACDé€9002¢%
ACD690031
ACD700024
ACD700038
ACEE€9CC2¢€
ACE690032
Elanket
AAX670004
ACBE€7C017
ACDETCO2¢€
ACD690024
BFX€E8C009
blovers
ACAE€ECC14
boiling
IAFE€7C047

boron trifluoride

ACDE6EQC17
ACD670026
ACD680015
ACD€EEC023
ACD690024
krazing
AAX670006
ACEESCCOE

Keysord Index

MDA690001

CAX690052
CCX680033
CXx700049
LAX710019S
LCA680008
LCA6S90037
LCB680007
LCB710007
LCC710024
LDA690012
LDA700046

CXX700049

ACE7C0025
ACE70003S
CXX700049
GLX7 10028
LDA690012
LDA6900123
LDA690038
LLDA7000 14
LDA7C0015
LLA7C0046
LKX7C0030
LKX7 10001
LXX700029
LXX710021
LXX710026

IAC6€0025
IBB670039
LJX660006
LJX670032

ACA670016
IAF6700u48
ACD690031
ACL7C0038

CRX69005:
CXX700049

ACE6B0024
ACE690026

ACE660012
ACEG6€001E
ACE670021
ACE67CO2E
ACE6 80017

treeding perforeance

AAX670009
ABX6&L00CY
ABX670049
ABXEECO3S
ABX690007

ABX65C0S¢€

ABX700054
ABX71C02C
ACE660009
ACBGECO1S
ACE670024
ACBe€CO13
ADX640021
ALX690063
BAX6GECOCE
bubbles
ACB6EDO2C
ACE690022
ACB7C0036

ACLC680015

ACDEECO22
ACC690024
ACDEC003
cCcx680033
turnup
ACL680022
calculaticos
AAX670009
BBX67C012
MCA6 20001
MDAGZCOC2
KCA6 30002
MDAELCOC
MFDRAELCOC2
cagpital costs
ABX7C00%4
BFX700056
IAAGECQ3C

capital eguipment

IACE6€0025
IAC700047
capsules
ACL650007
ACD660017
ACD67CC1S
ACL670026
ACD6ECOD1E
ACL680022
capture
ACB690029

ACE€90032
ACE700025
ACE700039
FBX640015

BEX€70012
BFX680009
BEX 7000 16
BFX700056
ELCX€80031
EDX€90051
TAAES50024

- IAA660030
- IAC€60025

TAC700047

-I2C710013

IAF €70047
IAFE70048
NXX%90002°
NXX690046

IACT710013
IEC€80036
MCD€90017
MLAT700032
MDB700033
MLC€70001
MDC€7C002
MLC€30016

BEX €70012

MLCA€S50001
MDA €90001
MLA€90002
MDA 700006
MLA700007
MDC €900 16

IACT710014
LKX€20003
NXX700058

IAC700051
IAC?710013

ACDEB0023
ACD€90024
ACD700038
CEX€640018
GDX€90042

EAX 300008
")

a

)

carbides
ACDEECC1?
ACD680022
ACE€7C028
ACE680017
ACE€EC024
carbon
ACDES0024
EXX70004E
carbonates
CAXES0053
carriers
CAX€80032
casting
: FBXEUCO15
cells '
ACB€9C022
ACB690029
ACDEECC17
ACD670019
ACD€7C02¢

ACD680015

ceramics /
FBX64001
cermets
ACF€8C02K
cesiurm
ACD70€024
charcoal
HIXE€CC2€

chemical properties

ACD670019
ACDEBOC1S
ACD680022
ACDEGC024

ACD630031
chemical reacticns

- ACD670019

~ ACDE€7(C26

" ACDEBO001S

“ACD680022

~ACDE€8(C023

ACD690024

- ACDE€9C031
chemistry. -

~ AAXE€7C00%S

- ABX580001

. ACD650011

ACDEECCIT

ACD670019
ACDG67CC2E€E
ACD680015
ACDEBCO1E€
- ACD68B0022
ACDE€80023
ACD690024
chlorides

Keyword Index

FBE7.10017
FCC700040

FCC700044

FCC710010

FB2660020

ACD680016
ACD680022
ACD690024
ACT690031
ACD700038

GGX670034

1CC710024

ACD700024

ACD7C0038

CAX68003z
 CAX690053
CXX700049

ACD7000724

ACD700038 =
CAX690053
 CXX700049
LGX650002
LXX660031 .
LXX7C€0029

ACD690031
ACD7C0024 -
ACD700038 -
CAX680032.
. CBX69005Z
CCX680033
CXX64002Z0

CXX7C0049
LXX660031
MCA680004

ACC7€0037
CAX690053
clrcriurm |
ACD670026
ACDEECO22
circulation
ACD6ECO17
ACL670020
ACD6 70027
coatings
- ACB680013
~ ACE€7C02¢8
ACE6 800 17
cckalt ;
ACE6B0024
ccke
EXX700048
columns
ACC690030
ACC7(C023
ACC700037
ccmpatikility
AAX670005
AAX€7CO0C6
ACLC650007
ACD660017
ACD65CC1S
ACL670020
ACDET7002¢€
ACLC6700 27
ACD6300 15
ACD6ECO1E
ACL680022
ACDEECO23
"~ ACL690024
ACD7C0024
. ACL700033
 ACE6E500CE
- ACE650014
" "ACE660012
ACEE€€CO1E
ACE670021
ACE67002E

ACE680017

ARCEEECO24

ACE6900 26
components

AAX67C0CT

ARX670008 .

ABX6SC0%6
ACA6600 14
' ACA€7C023
ACA680012
ACB7C0022
ACE700036
CXX7CCO4S
HEX620006

Page 173

IAFE€70047
TAFE70048

cXxX 700049

CCX€80033
MLCA660003

‘MDC €80002

ACEEB0024
ACE690032
ACE700025

ACEE600 18
ACE630024
LEX €80027

ACE€90032
RCE700025
ACE700039
BGX €70045
CAX680032
CAX €90052
CAX€90053
CEX€40018
CXx700049
FAX€20004
FAX €20005
FAX€90035
FAX €90045
FBD€90036
FEE€70031
FBE€90034
FEE700027
FEX €40015
FCX€90033

- FCX 700026

GDX690042

. GFX€60023
NXX630001

HBX700012

HIX €60026
HXX€640019
IEB€70039
JDX€90060
MEX €40003
MCA €60001
MCDE€80010
MCD 700001
MLR700003

 
 

 

Fage 174

components
_*¥Continued*
NXX€3C001

~

Keyvord Index

conpressive Properties

EBX€SCO036S
EBX700042
computer ccdes
- AAX670009
ACCESCC23
BBX€7C012
FCD710016
HCX71C022
IBB670039
JARTIC0CY
computers
JAB€SCO1E
JAB700017
JAB710008
JCX€9C01¢
concentration
ACDE€ECOYV7
ACD670019
ACD680015
ACDE€S80(22
ACD690024
ACD€9C031
ACD700024
concefptual design
ABX670049
ACBEECQOS
ACB66001S
ACB670017
ACBE€7C024
ACB680013
ACB€8002C
ACB690022
ACBE€SC02¢
ACB700022
ACB7CC03€
ACB710029
ADXE€LHCO21
ADX670046
GDX71C025
construction
ACAESCO04
ACCH650006
ACC650012
contactors
ACE700039
containers
NXX63000 1
containesent
- ARX670010
ACA€6CCO8
ACBEBCO2C
- ACB690029
ACB70003¢€

EDX640016

LKX700030
EDA67004C
MDAET70041
MDA69000 2
MDA 700006
BEC700008&

JEX650020
MAC68003%
MDA69000CS
MDC7C0004

ACD7(€0038
CAX680032

CRX 690053

CDX670035
CXx700049
NXX700011

HCX7 10022
TAR650024
T22660030
IAC660024
TAC660025
IACT700047

- IACT7C0051

IAC710015
IAC710014
IAD700052
IAE700059
IAF670047
IAF670048
IAF690014
LJX660032

LEX690011
MEX640003
MEX7C000z

LKX7C00230

TIACT700047
IACT7 10013

MCA660001

MCB7100 12
MDA 640007

CAX6¢€C053
IAC6€002Y

ccntamination

JDX€7C037

caontrol

AAX6700CE
ACE680020
ACB€<0022
ACE6€0029
ACE700022
ACB7CCO3¢
IAC710013
IBB710015

control rods

ccntrol-rod drives

AAXE7C0C?
ACR650010
ACR660008
ACAECSC021
ACR690028
ACBEE002ZC
FEX640015
GGX670034
TAC7CC047
IAC710013

AAX670007
MAC6EQO34

converters

ACB7(C022
ACE710029
ADX690063
GXXE6ECO038
IAR650024

coclant locps

ccclants

ACB7 10029
ACD67C02¢€
ACL680015
ACDEECO22
ACL680023
ACD6SCO24
ACC690031
ACD7(0024
ACD7000 38
ACE690026
ACE€SC032
ACE700025

cooling

COres

IAC660025

AAX670007

ACA71C028

ACEEECO13
ACE680020
ACB6S0022
ACD660017

FCL7 10016

MDA €70038
MLA 680003

~ MEC 700053

' JAAR 710009

JAB€90018
JAB 700017
JAB710008
MACEBO034
MDA €40007
MLAE90005

MACE€80034
MCD690017

MDA €40006

MDA €60003

‘MLA€90005

MDC 670001

‘MLCC€70002

MDC €80002
MDC 700004

MLA€90005
MDC7C0004

IAC €60025
IAD700052
LKX €20003
NXX700057
NXX 700058

JLX €90060

CAX €30053
CXX€eu00290
CXx700049
FBD 690036
GFX €60023
IACE60025
I2C700047
IAC710013
IAC710014
JCX €90019

‘NXX€30001

TEAT710005

ACDE70020
ACDE€70026

" ACDE70027

ALX €70046
HAX700050
IAC710013

 
o

")

cores
*Continued*
IAF670047
IAF670048
corrosion
ABX58000 1
ACDEEQC17T
ACD670019
ACDE€7C02¢C
. ACD670026
ACDE€7C027
ACD680015
ACDE€BCO22
- ACD680023
ACD70C024
- ACD700038
ACE€EC008
ACE650014
ACEE€CC12
ACE660018
ACEE€7C021
ACE670028
ACEE€8C017
ACE680024
ACE€9C02¢
ACE690032
ACE7CC02E
. CAX680032
CAX€9C053
corrosion products
- ACC€ECQOE
ACD660017
ACD670019
ACDETC026€
- ACD670027
ACDE€BCO1S
ACD680022
ACDESCO2Y
corrosion protection
- ACBETC(C17
- ACC660010
ACCEECCT1E
costs o
- ARXETCO004
~_ AAX67000¢€
AAX670009
- AAXE€7CCI1C
‘ABX680035
ABX€9CO5€
ACB660009
cover gas
ACA690028
ACDEECC17
ACDE€7C02¢€
ACD680015
ACDEEQCTIE
ACD680022

Keyword Index

MELA620001

CXX700049
FAX620004
FAX620005
FAX690035
FAX690045
FBD690036
FEE67003 1
FBE690034
FBE700027
FBE710018
FEX6400 15
FCD710016
FCX690033
FCX700026
GAX700045
GCX610002
GLX 690042
GFX660023
GXX680039
LBX680027
LIX690008
MCD690017
NXX63000 1

ACD690031
ACL700024
CXX700049
LAX690010
LAX700013
LEX680027
LHX6900 11

ACC680021
ACE660018

ACC660010

ADX640021
TAR650024
LJIX660006

LJX660032

1JX670032
NXX 700060

ACD69003 1

ACD700024
CCX680033
MCT6900 17
MDA 700006
MLC690015

Page 175

ACD6ECO23 MDC 690016
ACL690024 NXX€30001
cracks ’
"ACAG6€CO014 EDX€4 0016
ACE650008
Creep
ACE650008 FEC€10001
ACE6£0014 "FBCE€S50017
ACE660012 FEE€600 19
ACEE€€CO1E FBE€70029
ACE670028 FEEE70030
ACE6EC017 FBE670031 -
ACE6 80024 FEE€80025
ACEE€<C0Z6 FBE€B0026
ACE690032 FEE€90034
ACE7C002°% FBE 690044
ACE700039 - FEE700027
EBX7CC042 FBE710017
ECX7 10011 FEE710018
EDX6ECO31 FBXE40015
ELCX690051 FCC700040
- EXX7CCOuE FCC700044
FBEF660021 FCC710010
FBB6€C022 FCE€90043
FEE690040 FCE710004
FBBESCOU1
criticality
ACAG6EC001C MDA €4 0006
BEX670012 MLA€60003
MCAGEQOCY MDC 680002
MCA640002
crcss secticns
AAXETCCCS BBX€70012
BAX680006 MADEYOOOU
- BAX7C00CE MCD€90017
crystallization |
 CAX€ESCO€1
cutting tools
: ACAT710028 KBB €90006
data '
ACBEECO13 CAX€90061
-ACE690029 CCX €80038
ACC7€0037 "IBDEBOO36
ACE650011 LEAT700014
ACD660011 LDA700015
| BAX7C00CE o
data acquisition systems
JEX€EC02C MDC 700004
FAC6B0034 -
data processing o
' JEX650020 MLC€70002
- MDCEICOCT MDC 700004
decay
ACE690022 ACDET0027
ACB7(C03¢ MDA €90001
ACLh60017 MLAGT00007
ACDETCO2C

 
 

Page 176

decommissioning

MFX700020.

decomposition
ACD670026
ACDEEBCC1S

decontamination
LIXEICO08
LIX690009

defects
ACE€9C02€
ACE690032

delayed neutrons

ACD680023
MDAG64D006
MDA €E€C003
density
ACD€€ECC17
ACD680022
ACD690024
ACDEY0031
ACE660012
ACEEECO1E
ACE670028
ACE€80017
ACE680024
ACE€9C032
deposition
ACDEEOCY?
ACD690024
ACD700038
ACE€7C028
ACE680017
ACE€80024
ACE690032
descripticn
ACD650013
HBX€9CO05€
HBX€9CCS9
TAFP690014
design
ABX640004
ABX70CC54
ACCEESCO0E
ACC660010
ACC€9C03C
ACC700023
HAX70005C
‘HBX620006
HBX€7C042
HBX690058
HBX€9C056
HBX700012
HCX€8C037
IAB670043
IAET0CO05¢
IBAT10005

Keyword Index

CAX690053

LJX660006

ACE7C0025

MDA690002z
MDA 690005
MDC680002

ACE7C0039
CAX690053

- CCX680038

CXX7C0049

ELX6400 16

EXX700048
FBX6 40015
MDA 640002
NXX630001

ACE700025
ACE700039
BGX67004°
CDX67C035
cxx700049
MCD690062

JFX660027
MBX640003
MCD7C0001

IBB670036S
IBB71001¢%
IBD680036
JAA7 10009
LHX690011
LJX6€0006
LKX620003
MAC680034
MAD690004
MAX650019
MCD690055
MCL690062
MDB7€0003
MEAG640005
MEB680001

MEC7000523

design criteria
ACB6€C01ES
BGX6 70045
CDX670035
design data
AAX€70006S
ACE6700 24
ACB7 10029
BAX6EQOCE
BEX670012
develofrment
AAX670003
AAX6700C4
AAX67C005
AAX670006
ARX67C007
AAX670008
AAX670011
- ABXSE€00C1
ABX6 40004
ABX6£C035
. AEX690056
ABX7C00E4
AEX700055
ABX71002¢C
ACE670024
ACB6ECO13
ACE700022
ACB700036
ACB71C029
ACC700037
ACDEEC013
ACE650008
ACE6E5C014
ACE660012
ACE€€CO01E

ACE67C028

ACE680017
ACE€ECQ24
ACE690026
ACE6¢<C032
RCE7(002S
diagrams
ACD6€C017
ACL690024
diffusicn _
ACE670028
CDX€7C035
FCC690048
FCCE<CO4S

~disconnects

ACR67001¢€
dismantling
' ACD67C02C
dispersion
LXX71002¢

IAE700059

~ MACE€B0034

HCX710022
IA2€50024
TIAET700059
LDA €90012
MLB700003

ELX €80031

- EDX€90051

FAX€20004
FAX €20005

- FAX€90035

FAX €90045
FCC 710010
FXX€90047
GLX 710025
HAX700050

- HEX €20006
"HBX€70042

HEX €90058
HBX€90059
HEX 700012
HCX€80037
HEFX €20007
HIX €60026
HXXEL 0019
JLX €90060
KBB €90006
LHX €900 11
LJX€70032
LKX 700030
LKX710001
LXX700029
MEAE40005
MEB€80001
NXX £90002

NXX€30046

NXX7C0060

ACD700038

FCD710016
MDA €70040
MLAE70041

ACD670027
-
*

o

&)

disgosél

- dissol

MDAT700007
ving

ACDEECO17
CaAx680032

distillation

AAX670004

 ACCEECO01C

ACCE€001€
ACC670018

- ACCE€7C02¢

ACC680014
ACC€80021

- ACC690023

ACCE€SC03C
ACC700023
ACDEECO11

- ACD660017

distri

ACEEECO01E
bution

ACCE€8C014
ACC680021

‘ACC690023

drain

ACC€9C€030
ACC700023
ACC700037
ACD670026
ACDESOO1S
tanks

ACBE€9C02¢S

‘ACB70003¢€
ACE650008

d:yihg

ductil

IACTCCOL47T

ACALESCOO0Y
ity

ACE£ECO0E
ACE650014

ACE66C012
 ACE660018
ACE€7C021

ACE670028

 ACE€80017

. ACE680024

 ACEEYC02¢€
. ACE690032

ACE700039

 EBX700042
FBBE60C21
FBB660022

FBBES00LC
FBB69004 1
FBB7CC028
FBC61000 1
FBCEU40017

"~ Keyword Index

AT X
ek

CAX690052

CXX700049

ACE680024

- GDX69004z2
LCA670014

LCA680008

- 1LCA690037

LCB680007
1LCB710007
LCC7 10024
LIX660006

LIX660032

LJX670032
LXX660031
LXX710021

ACD680022
ACL690024
ACD690031 .

ACD7C0024
ACD700038
CDX670035
LDA6900 12

LDA700046

IACT10013
MAY650019
MDA64000 2

FBC650017 _ -
- FBE6%0015
‘'FBE650016
FRE66001S -
‘FBE670029

 FBE6T70030
FBE680025
FRE6B00Z6 =
'FBE690044
~ FBE710017
" FBE710018
FRX6400 1%
- FCC7C0040.~
FCCT00044
'FCC710010

FCE690043
FCE710004
GAX670033

GAX680028

dynamic characteristics

AAXETCO09

MDA650001

cartbhquakes -

econonics

elasticity

electrolysis

Page,!??

AAX670010 MLCA€80003
ACAEECO12 MDA €30002
ACR6ECO1S MLAE90005
ACE680013 MLAT00006
ACB6<0029 MDA 700032
ACE700022 "MIAT710003
BFX6€ECOCS MDB 700033
BGX67C045 MLE710002
JAE690018 MLCE€B0002
JAB7C0017 MDC €80005
JAR7 10008 - MLCE€90003
JCX6S001¢9 MDC€900106
KCAa680004 MLC700004
MCD6<0017 MDC 700005
dynamics tests
"~ ACA6ECO010 MDB 710002
AC2660008 MLCE60002
ACA€SCO21 MDC €80002
KDA660003 MLCE90003
MDA6SCOCS MDC 700005
MDE700033 ' |
ACB690029 ACB 700036
ABX€40004 CAX€90053
AEX670049 IAR€50024
ABX6<€0007 TAA€60030
AEX710020 NXX€90057
ADXE€SCO63 NXX700057
BFX700056 NXX 700058
ACEEC028 EDXE4 0016
EEX700041 FEXE€40015
ECX710011
electrical circuits
MACEE0034 MAX€50019
electrical conductivity
ACD6ECO1S - ACE€60018
ACL680022 - ACE€70028
ACDE€<SC024 ACE€80017
ACD690031 "ACE€80024
ACD7C0024 ACE700039
ACD700038 FEX€40015
A ACE6€C012 GCX680030
electrical power 7
- ABX6SC0C7 - NXXE90057
~ AEX700055 NXX 700057
ACB6€CO15 - NXX700058
" ADX690063  NXX 700060
electrical prcperties :
ACD660017 ACL€80022
- ACD6EQO1S - ACD 500038
- ACC690023 ACD670019
ACC690030 ACD€80015
ACC7C0023 ACDEB0022
ACC700037 ACDEJO0024

 
 

Page 178

electrolysis
*Continued*
LKX70003C
LKX710001
embrittlement
ACD6T70027
ACEEE000¢8
~ ACE650014
ACEE€€ECC12
ACE660018
ACE€7C021

emergency cooling

CAX€9CO053
enerqgy -
NXX700060

engineered safeguards

MDA68000 3
engineering
LXX700029
entrainrent
ACDEE0017
ACD700038
CCX€80033
environment
JDXE&7C037
equilibriunm
ACD660017
ACDETC02¢
ACD680015
ACD€80022
ACD690024
ACDEICO031
egquipment
ACDETC02¢€
ACE6 50008
ACEG660012
ACEE€CC1E
ACE670028
ACEES80017
ACE680024
ACEE€9C032
ERX70004 2
erosicn
FBX640015
errors
BAX€8000€
examinaticans
ACA 700035
ACA710028
ACDE€SC007
ACD660017
ACDE€7C02¢C
ACD6T70027
ACDE€SBCC15
ACD680016
ACD€8C022
ACD69Q024

Keyword Index

LXX710021%

"LXX71002¢€

ACE670028
ACE680017
ACE680024
ACE690026
RCE690032

. ACE700025

LCB710007
LCC710024
NXX700011

ACD700024
ACD700038

CXX700049

LCA680008
LCa690037
LDA690012

GDX690042
JFX660027
JFX670036
LTX650023
LIX670012
MEA640005
MER68000 1
MEC700053

EBX670012
ACE660012

ACE6€0018

ACE670021
ACE670028
ACE680017
ACE6B80024
ACE690026

ACE690032

ACE700025
RCE700039

ACD690031
ACC7C003E
ACE650008
ACE€EC014

i =
excursions

IAF6SCO14
JAE6900 18
JAB7C€0017

expansion

ACB67C024
ACL680022
ACE6€0012
ACE6600 18
ACE680024
ACE€SC026

experience

ABX67C04S
AEX6 90007
ACRA6ZC0CY
ACA6S001C
ACR660008
ACAEECO1Y
AC2670016
ACAG67CC23
AC2680012
ACR680019
ACR6SC021

" ACAa690028

ACAT7CCO021
ACA700035
ACA71002¢E
ACL650011

 ACD660011

HEX7000 12

HCX680037

LCC71C024

experiment

ACC7C€0037
ACD650011
ACDE€E€CO017
ACLC670020
ACD67C02¢
ACLC670027
ACD6ECO15
ACLC6EQDYE
ACC680022
ACD€<C024

ACDb90025

ACD7(C038
CCx680033

. CEX64C018

CLX700010
CXX7C0049

extraction columns

GDX710025
1XX700029

EDX €40016

‘MEC700053

MFX 700020

MDA €6 000k

MCAEB0003
MDA €30005

ACE6€90032

ACE700025
EXX700048
. FEXE€40015
HCX 710022

MCD €80010

MCD€90017

- MCD €90055

MCL€90062
MCD700001

MDA 660003

MLAE€70038
MDA E€7COu4

MLB700003

MDB 710002
MDCE€70001
MLCE€T0002
MDC€80002

- MLC€80005

MDC €90003

- MLC700005

MEB€60028
MEC700053

- NXX €30001

EBX700041
EEX700042

- IAF€90014

LXX€60031
MAX€S50019

- MCRAE€60001

MCA €80004
MEA€60003
MLC€60002
MDC 680002
MLCE80005
MDC €90003

MLCE€90016
. BDC 700004

MEC700005

- LXX710021
LXX 710026
)

"

[ 2]

fabrication
ACE670028
ACE€8C017
ACE680024
ACE€9C026€
ACE690032
ACE70002%
ACE700039
EDX€4C01¢€
failures
ACAEGECO14
ACE€E0N00E
- ACE650014
ACE€€C012
"~ ACE660018
ACEe€7C02¢
ACE680017
fast neutrcnhns

IAF670047

TAFE€TCO4E
fatique
EBX€9CC39
. PBCE90001
feedtack
- JCX€9CC16S
MDA690005
ferrcalloys
ACE690026

fertile materials

BAX7CCOO0E€
films
CDX€7CC3€¢
filters

~ ACREE€Q01U

HIX660026
filtration

HIXEECD2¢€

LAX690010

LDAT7CC014

fissile materials
ACD€7C02¢

ACD68B0015S
_ACD680016

" 'ACD€EC022

~ ACD700038
fission ' '
ACD660017
ACDETCQ2C

fission products

‘AAX670005
- AAXE€ICCOS
ACA660014
ACB€390022
ACB700036
ACD€ECC17
ACD670020
ACDETCC2¢

Keysxord Index

EXX700048

FBB650018
FBX6400 15
GDX710025
GGX 670034
GXX680039
MADE9000 4
MBX640003

- ACE680024

ACE690026
ACE690032
ACE700025
CAX 690053
NXX630001

IAF6900 14
FBX640015

MDAT710003
MDE7 10002

ACE700025

LHX690011

LIX690008

- LIX690009

BAX7C0008
CAX690052

CCx68003:

| CXX700049

~ ACD670027
- BAX7C€0008
- €CX680033
CDX670035
.CXX700046
- IACT7C00047

IAC7 10013

JDX690060
LCA680008

LYX660031

ACD670027
ACD6EQO1E
ACL680022
ACD6EQO023
ACL690024
ACDESCOI
ACE700024
ACD7CQ03¢€
BEX670012
BFX7C001¢
BGX670045
flanges
NXX€:00C1

flexural properties

EDX€UC016

flow measurement

AAX670008
ACA710028

HAX7€005C

flovsheets
ACBE€COCS
ACE6600 15
ACB6800 13

ACB6¢C029

ACE710029
ACCE€CO1€
ACC690023
ACC6S003C
ACC7C0023

fluid flow
HaX7C00S0
: HCX7 10022

fluids

- CPX670035

flucrides
AAX€TCOCS
ACD650011
ACDE6€C017
ACLC670020
ACDE€7C027

- ACL6800 15
RCD680016

ACD6ECO22
ACL6900 24
ACD6¢SCO31
ACE700024
ACD7(CCO3E
ACE6ECO14
ACE660012
ACEEECQ1E

ACE670021

ACE67C028
ACE680017
ACEEECO24
ACE690026
ACE€S0032
ACE700025

Page 179

LXX 200029
1¥Xx710021
MADEIQ00Y

MCA 680004

MCDEBOO10
MCD€90017
MCD€90062
MDA €70041
MLCAT700007
MDC67C001
MLCE670002

LXX710026
MLB700003

ACC 700037
124660030
LIX€50023
LIX670013
LJX €60032
LXX700029
M AX €500 19
MCB 710012

MDA €20001

NXX €30001

ACE700039
CAX €80032
CAX€90052

~ CRX€90053
- CAXE90061

C2X 710023
CCX €80033

- CCX€B80038

CXX €40020

CXX700049
- FAX€20004

FAX620005

 FAX690035
 F2X €900U45

FBE€70031
FEEEI0034
FBE 700027
FEE710018
FBX€40015
FCD710016
GCX€10002
LAX €90010

 
Fage 180

fluorides

*Continued+*
LCB7 10007
LCC710024

fluorination
ACALE€80019
ACB670017
ACC660010
ACCEECC1E€
ACC670018
ACCE7C02¢E
ACC680014
ACCEB8C021

ACC700037

fluorine
ACA700035
ACD65C007
CEX640018
fluorcborates
AAX670005
ACB€8C02C
. ACB€SC022
ACB700036
ACD€SCO11
ACD660017
ACDETCO2¢€
ACD680015
ACDEEBCO22
ACD680023
ACDEY0O24
ACD690031
ACD7CC024
ACD700038
ACE€6CO1E
ACE670021
foaming
ACD690024
foreign
ABX700055
ABX71C02¢C
ADXEY4CO021
forming
ACEBE9CO2¢
ACE690032
freeze flanges
ACA680012
ACA€8001S
ACA690021
freeze valves
ACA650004
HFX€2C007
freezZing
ACD€9C024
CAX690061
MDAG640002
fuel cycle
ACB680013

Keyvord Index

LHX6500 11

ADX640021

TAn650024

LBX680027
LTX650023
LIX670013
LIX690008
LJX66000€

1JX660032

MCD700001

CLX700010
LGX650002
ADA69000 1

ACE670028
ACE680017
ACE680024
ACE69002¢€
ACE690032
ACE700025
ACE7C0039
CAX 690053
cCCX680038
CXXx700049
FBD690036
FCX690032
FCX700026
GAX700045S
HCX710022

CcCx680033
ADX670046
IAF670047
IAF670048

ACE700025

ACA69002z8

~ ACAT00021

IAC710013
MAX650019

MDA660004
NXX630001

LXX700029

fuel cycle costs

CXX7€004S

AAX67C0C3
ABX690007
ABX7C00%54
ACE660015
ACB670024

BFX6€E00CS

fuvel preparation

fuels

gamga

gaema

gamma

EFX700016
BFX7C00%6

ACDG6EQO1S
ACD680022
ACDESCO24
ACD690031
ACD7C€0038

AAX67COCY
ACC670018
ACD670019
ACL6ECO1E
ACD690024
ACD7C002¢4
CaY680032
CAXESCO0S2
cCcx680033
CXxei002Q
CIX700049
FAX€200C4
FAX6 20005
FAX6SC035
FAX690045
radiaticn
ACL660017
ACD67C02C
sources

ACD6EQCO23

spectrometry

ACA67C023
ACA6S00ZE
ACA700021
ACD6€0017
ACL670019

- ACDE€7C02C

ACL670027

gas analysis

ACL670026
ACDEECO1E€

gas flow

CDX67C035

gas injection

ACBEECO2C
ACE690022
ACB690029

gas separation

ACE680020
ACBESCO22

NXX700058

 IAC660025

1aC700047
IAC700051
I2C710013
IAC710014
IAD700052
LIX €60006
NXX700058

LAX690010
LAX700013
LIX€S0009

- MLAE€80003

TAC€60025
IAFE€90014
LCA€70014
LJX€70032
MCAEBOO004

" MCDEB0010

MCDEIDO 17
MDA €30002
BLCA€40002
MDA €90001
MLAE90002
MDA 700007
MLCE€70001
MDCE€70002
NXX £90002

ACDE70027
ACD€80023

 ACD€80015

ACD680022
ACD€90024
ACD 690031
ACD700038
MCDE9S0017
MCTC€90062

ACDEI00 24

ACB 700022
CCX€80033

- ACB€90029

ACB700022
gas separation

*Continued*
~ACB700036
ACB7CCO03E
ACB710029

gases

ACD660017
ACDE€7CC26
ACDEBQO1E
ACD680023
glass
. JPX670036
graphite
ARL670006
ACB€7C017
ACBE€7C024
ACB680013
ACB€8002¢C
ACB690022
ACB700022
ACB700036
ACDE€EC007
ACD660017
ACDE€7C019
ACD670020
ACDE€T7C02¢
ACD670027
ACDE€ECO15S
ACD680016
ACD€80022
ACD690024
ACD69(CC2%
ACD690031
ACD70C024
ACD700038
ACEES000¢
ACES650014
ACE€€CQ12
ACFE660018
- ~ACE€7C021
ACE670028
- ACE€8C017
hafnium

- ACE€B0024

ACE690032
hardness
' - ACE700039
bealtt physics
e MACEBCO3Y
- MEC700053
heat '
- ACB690022

- heat tkalance

ACAT1C02E
MDB700003
heat exchangers
AAX670007

Keyword Index

CCY680033
IAC710013

ACD690024
ACD700038
Ccx680033
CXX700048

ACE680024
ACE690026€

ACE650032

ACE700025
ACE700039
CAX680032
CDX670035
CEX6 40018

CXX700049
FBX690039

EBX7G0041
EBX70004¢2
EBX700043

- ECX710011

EDX640016

- EDX680031

EDX690051
EXx700048

FBX640015

FCX690033
FCX700026
IACT700047
IAC710013

IAE700059
MAD690004 -
MAX650019
 MDA670040
MDA670041

ACET00025

. NXX700011

'MDB700003

HCX 680037

"helium

ACA710028
ACB6600 15
ACBE7C017
ACE670024
ACB6£0020
ACE690022
ACB7(0036
ACE66001€
ACE700039
CAX6SC0S3
FEE6500 18

FBXE4CO15

heat generation

"IAC710013

LKX7100C1
LXX7 10021

heat trapsfer

ACA660014
ACA67C023
ACE660015
ACB6ECO2C
ACE690022
ACB6¢SC029
ACB700022
ACB7(€003¢
"ACC690030
ACDGECO1S
heat treatments
ACAG6ECOCE
ACE650008

- ACE€ZCC14
ACE660012
ACE6E€CQO1E
- ACE670021
ACE6TC02E
ACE680017
ACE€EC024
ACEG690026
ACEESC032
ACET700025
RACE7CC03¢
- ELX640016

AAX67C0CY

 ACD670026

ACL66C031
IACT10013

hot cells
ACD670019
ACDE5C02C
ACL670026
ACD670027

hydraulics
| AAX670007
HBX6200C6

~ EXX7C00ug
“heaters ' '

Page 181

HCX710022
IACE60025
IAC 700047
1AC700051
IAC710013
I2C7100 14
IBA 710005
IEB670039
JCX €90019
MAX650019
MEB700003
NXX€30001

MDA €20001
MEA €40006

ACD680022
ADX670046
CAX €90053
CLX700010
1EA710005
JCX€90019
MTA€90005
MDB 700003

- NXX€30001

FBA 660020
FEAEB80029
FBB€60021
FEBE€60022
FBB 700031

“FEC€10001

FBC640017
FEEEB0026
FBX€40015

- FCC700040

FCC 700044

- FCC710010

GCXEB80030

':sxx€30001

HCD €900 17
- MEC€90015

MLCE90016

" ACDEBDO 1S
"ACD €80022

ACD€90031

HEX 700012
HCX€80037

 
 

Page 182

hydraulics
*Continued*
MDAT00006
hydrocarbons
"ACD660017
ACDE70026
nydrodynamics
MDA ICCO06

hydrofluorination

ACDEBQOO01S
ACD680022
ACD690024
ACDES0031
hydrogen
- ACDE€8001S
ACD680022
ACD690024
ACDE9CO031
- ACD700024
LAX€ESQO1C

hydrogen compounds

ACCEECO0E
ACD680022
ACDE€9C024
impregnation
ACEE€7C02€
ACE680017
ACEE€E8CC24
ACE690026
impurities
ACD670026
ACDESCC31
ACEESCO0E
ACE650014
ACE€€CC12
in-pile tests
ACDESCO07
ACL650013
ACD660011
ACDETCO1S
ACD670027
ACDEBOO1E
~_ACD680016
ACD€SC023
-~ ACD690024
ACDE€9C02¢
ACE650008
ACEEEC014
inconels
ABXS80001
ADX640021
FBX640015
FCD71001¢€

industrial studies

ACB70C03¢€
industry
ADX690063

Keyvord Index

ACD680023
HIX660026

LA X6S0010
Lax700013
LAX71001S
LIX650009

LGX650002
LGX650002
LIX650023
LIX670013
LIX690009

ACD700038
LIX650023
LIX670013

ACE690032
ACE70002%
ACE7C0039

ACE69002¢€
ACE690032
RCF700025
FELC69003€

ACE660012
ACE660018
ACE670021
ACE670028
ACE680017

ACE680024

ACE690026
ACE690032
ACE70002E

CEX640018

JFX67003¢€

GAX700045
GCX610002
GCX68003C
GGX670034

ACB710029

NXX700060C

inert gases
ACD660017
ACLCEeT7C026
ACD6800 16

ACDEEQOZ3

ACE690024
ACDE6S0025
inspection
AAX65COCE
ACE650008
ACE650014
ACEE€C012
ACE660018
instrumpentaticn
AAX670008
- ACB680020
IAC710013
JAB6900 18
JAB7C0017
JLX670037
JDXESCO€EC

interfacial tension

ACD6ECOQ15
ACL680022
ACD6<SC024
intrusion
EDX64CO016€
inventories
ACDET0026
ACD680023
ACD7C003E
iodine '
ACD650011
icns
ACL690024
iron
ACE700039
GCI6 10002
iron alloys
ACE6S50014
ACE6€C012
RCE660018
ACE6C02E
| "ACE690026
irradiation
ACD670026
ACD6ECO1E
ACB700038
ACE6ECOCE
ACE650014
ACE660012
ACEG6€C01E

ACE670021-

ACE€C02E
ACE6800 17
ACE680024
ACEESCOZ6

ACD€90031
ACD 700038
CCX€80033

. CDX€70035

MCDES0017
NXX€30001

ACE 670028
ACE€80017
ACE€80024
EDX€40016
MEX 700002

- JEX €50020

JFX €70036
MAC 680034
MAX€50019
MDA €4 C007
MLB700003

ACD 690031
CCX€80033
MAD€90004
ADX640021
ALX €70046
MDA 700007
LXX 660031
ACD700024

GFX€60023

ACE700025

- FBDE90036

FCX€90033
FCX700026

ACE€90032
ACE700025
ACE700036¢

- CEXEu0018

ELX€40016
EDX €80031
EDX€90051
EXX 700048
FBB€90040

FEBE90041

FBB 700028

" 'FBE€50015

 
-
®

»

irradiation

*Continuedx
FBEE€SC01¢
" .FBE660019
FBE670029
FBEE€7C03¢C
FBE670031
FBEE€B8002S
FBE680026
PBE€9C034
.FBE690034

FBE€90044 -

- FBE700027
isotoges
BAX700008
MDA7CC007
joints
ACB680013
ACEEEC00E
ACE650014
 ACE€€C012
ACE660018

kinetic equaticns

ACD690031
MDA €9CCOS
krypton
ACA650004
ACB€9C022

laboratory equipment

ACDEB0016
lattice
EXX700048
layout
ACB660009
ACB€8C013
lead | .
ACEESCO014
- RACE660012
ACEE€(CC1E
lead cooling
' ACB€€C015
7 ADX670046
leakage
- CAX€S0053
leaks. -
"ACA 700021
~ ACD670020

~limits

~ IAB670043
linings
- GFX660023

liquid level measure

GRX700045

LCB71C007

~ LCC71C024
liquid wetals

ACDEECO11

Keysord Index

FBE710017

FBE710018
FBX6400 1S
FCE690043

FCE710004
- GAX670033

GAX680028
IAE700059

- TAF6900 14

MAD690004
MDB700003

ACE670028
ACE680017
ACE680024
FBX640015

MDA7 10003

MDA670040

HDA6700u1

ACD690024

ACE690029
ACB700036

GFX660023

IAC660025

GXX680039

ACD670027
| CAX690053

fiént

LXX71002¢€
MAX6S5001S

CAX690053

ACD6 70019
ACEESC014
ACE6600 12
ACE660018
liquids
- ACD690031
liguidus
ACD670026
- ACDEECO1S
ACD680022
ACD690024
ACD700024
lithiunm
~ ACC660010
" ACC6ECO16
~ ACCE7C025

lithium chlcride
ACD7C0024
lithium fluoride

ACD6ECO17
ACL670019
ACDE€7C02C
ACL670027
ACDEECO15
ACLC6E0022
ACL690024
ACDESCO03Y
ACT700024
ACD7C003¢
CAX680032
CAX690052
- CAX6¢00E€3
loading
BACA€ECOCY
ACA650010
 LAX710C19
loop- '

ACE660012
ACE6€C018

ACE670021
ACE670028

ACE6ECO17
ACE680024
ACE7(€GC39
"FRX620004
lcosses :
. LIX690008
machining
| FRX640015
maintenance
- AAX670011
ACAEECO14
ACA67C023
AC2680012
ACA6ECO1S
ACRA690021

ACEESC014

Page 1813

CYXX 700049
FCD710016
GFX 660023
HBX 670042

ACD700038
CAX€90052
CAX €90053
CXX700049

'~ NXX €30001

LL2€90012
MDA €30001

CAX€90053

CCx680033
CXX 700049
LAX€90010
LAX 700013
LAX710019
LCR€80008
LCA€90037
LCBE€80O007
LCB710007
LCC 710024
LLAEY0012
LDA 700046
LIX690009

MCA 680004
MTAE80003

FAXE20005
F2X €90035
FAXE90045

"FEX€40015

FCD 710016
GAX700045
GCX €10002
IAC 710014

ACA€90028
ACB €80013
ACB 680020
ACB€90022
ACB€90029
ACR700022

 
 

Fage 184

maintenance
*Continued+*
ACB700036
- HBX 700012
HXX640019
TACT7CC051
IACT10013
JFXE7C03¢€
KBB690006
KAXESCC1S
manipulators
JFXETCC36
mass transfer
ACB67002¢4
ACBE€SQ02C
ACB690022
ACBE€SC02¢
ACB700022
ACB7CCO03€
ACE650014
ACEE€€C012
ACE660018
ACEETCO2E
ACE680017
materials '

: AAX670006
ABX700054
ACD670019
ACDETCO2¢€
ACD6B80015
ACDEEBCO1E
ACD680022
ACD€80023
ACD690024
ACDESCC3
ACD700024
ACD70003€
ACE650008
ACEE€S0014
ACE660012

materials testing

ACB680020
ACDETCC1S
ACDE€7C02¢€
ACD670026
ACDE€7CC27
ACD680023
ACE€SCO00¢E
ACE6500184
mathepatics
FCD710016
JCX€9CC16S
measurement
ARX670008
ACA€EC01C
ACD700024
ACD70C03E

Keyword Index

MBEX70000¢2
MCB710012
MCcD680010
MCD700001
MEAG6LOOOS
MEB680001

"MEC700053

NXX590002

ACE680024
ACE690026
2CE69003:
ACE700025
CXX700049
FBD690036
FCL7 100 1€
GFX660023
IED680036€
MDA 67004 0
MDA67004 1

ACE66001¢
ACE670028
ACFE680017
ACE680024
RCE69002€
ACE690032
ACE700025

CDX670035

CEX640018
GDX710025
HIX660026€
IAC700051
IaDp70005%¢
nca680004
NXX630001

ACE660012
ACE660018
ACE670028
ACE680017
ACE680024
IAE700059

IAF690014

MCR690005

CLX6700135
LCA680008

MLCA700006

MDB700033

ACEG€C01E MDCE60002
ACE670028 MLCE€90015
ACE6EC017 MDCE€90016
"ACE680024 MIC700004
CAXESCOE
mechanical properties
AAX67C006 - ACE690026
ACE650008 ACE€90032
ACEEECO14 ACE700025
ACE660012 ELX €80031
ACE6€001E EDX€90051
ACE670021 FEAE660020
ACE670028 "FCX€90033
ACE6ECO017 FCX700026
ACE680024 - GXX €80039
- melting |
ACD6900 24 FEX €40015
ACDE6SCO31 NXX€30001
CXX700049
mercury o
IAF670047 TAFE70048
setal transfer proccess ]
ACC700023 - ACD€90031
ACC7C0037 ACD700024
ACL690024 -
wretallograghy
ACE650008 "ACE€90026
ACEEECO14 ACE€90032
ACE660012 ACE700025
ACEGECO18 FBAE60020
ACE670021 FECE€40017
ACEESC02¢ "FBX640015
ACES680017 GCX €10002
ACEGECO24
metallurgy
ACE6EC0CE ACE€80024
ACE650014 ACE€90026
ACE6€0012 ACE€90032
ACE660018 .ACE70002¢%
ACE6€Q0017 ,
metals
| ACD67C016§ ACE€80017
RCEEESC0CE ACEE€BO024
ACES650014 ACEE€90026
ACE6€0012 ACE690032
ACE660018 ACE700025
RCEE€C02¢ ‘
methods
A2X67C0CS MDC€60002
MDA6 20002 . MLC700004
o MDA€SCOC2 - MDC 700005
picrostructure
- ACE650008 ACE70002¢
ACEESCO014 ACE700039
ACE660012 EEX 700042
ACE6<0026 EDXe€4Co16
ACE690032 EXX 700048
o

microstcucture
*Contlnued*
FBA660020
FBA680029
FBCE900C1
FBE660019
FBE€7C02¢
FBE690034
FBE700027
- FBE710017
mists
ACD670026
ACDEEBCO15S
- ACD€80022
- ACD680023
ACDESCO024
mixing
MDA €9CCO02
mixtures
CAX690052
nodels
FCD710016
FCD71C01¢€
HAX700050
"IBDEBCO3E
JCX690019
MDA€2C002
- MDRA650001
"MDAEECOC3
MDAG70040
‘MDA €9C002
moderators
- BFX€8C00¢
modlfled Hastelloy N
AAXETCO0E€E
ACE650014
ACE660018
" ACE€7002¢€
. ACE680017
"ACE€80024
~ RCE690026
- ACE€9C032
- ACE700025
. ACE70003¢
- FBE690034
FBE€ICOUY
modular design
- ACB€€0C15
. ACB670017
molten salts
- ARX670005
- ABXS5E&C001
ACAES5C004
AC2650010
ACDESCQO07
ACD5650011
ACDEGCC11

Keyword Index

FEE710018

FBX640015
FCC7C0040
FCC700044
FCC710010
FCE7 10004
GAX670033
G2X680028

ACD700038

CCX680033
CXX7C0049
HIX660026

CAX690053

MDA690005
¥DA 700006
MDC67000 1
MDC 670002
MDC68000 2

MDC 680005

MDC69000 3

MDC690016
MLC70000%.

CDX670035

FBE700027
FEE710018
FCC 690048

FCC690049

FCC700040

FCCT00044
FCC710010
FCE690043

FCE710004

FCX690033 -

FCX70002¢

FXX690047

ACB670024

FAX62000%

FAX690035
FBD690036
FEE7 100 18

FBX640015 -

FCX700026
GAX7CO0045

ACD6€0017
ACL670020
ACD670026
ACL670027
ACL680015
ACD6ECO16
ACD680022
ACD6€0023
ACD690024
ACD6SC025
ACC690031
ACD7C0024
ACD7C0038
ACE6500 14
ACE6€0012
ACE660018
ACE€7C021
ACE6700 28
ACE680017
ACE6EC024
ACE690026
ACE650032
ACE700025
ACE7CC03¢
ACX640021

"BGX6T700KS

CAX710023

CCX6€£0033

CEX640018

CLX7C0010

CXX640020
EBX7CC043
PAX620004

molytdenum

ACLC670019
ACD67C026
ACD67C027

‘ACE700038
- ACE6ECOCE

ACE660012
ACE6€C018
ACE670028

- ACEE€EQ0017

ACE680024
ACE€SC02¢€

monitors
-JDXG?COB?

- JEX690060

natural [eEsources.
fAAX670003

ABX67C04S

"ABX6S0007

ADX690063

" BFX7CC056
neptunium

BAX7CO0CE

Page 185

GCX610002

GLX €90042
GDX 710025
G XX€80039
HEXE€ET70042
HBX€90058
HEX €90059
HBX700012

- 1ACE60024

IAC700047
I2FET0047
TAFE€70048
IAFE90014
IEB€70039
LAX710019

"LEX€80027

LCA 670014
LCB680007
LCB 710007
LCC 710024
LTA€90013
LGX650002
LJIX€60032
LKX€20003
LXX €60031
LXX700029
LXX710021
LXX710026

MAX €50019

MDA 630002
MLA€70041
MDB 700003
KXX€30001

ACE€90032
ACE700025

ACE 700039

CCX €80033

~ CEX€40018

GAX €70033
GAX€EBCO28

 GCX€10002
 GDX690042
" GLX 710025

 MACE80034

NXX€90046
NXX700057
NXX700058
NXX 700060

 
 

 

Page 186

neutrcn fluence
ACB€7C024
neutron flux
ACBE€7C024
IAC700047
IAC710013
MDA€2C001
neutron physics
ACBE80013
ACB680020
ACB690022
ACBE€9C029
ACB700022
ACB70003€
BAX700008
neutrcn scurces
IAF690014
MDAE3CO002
neutron spectra
ADX640021
neutren yield
AAX670009
nickel
ACDE7CO01S
ACE700039
ADXE€40021
FAX620004
FAX€2C00€
nickel alloys
ACDE€7C019
FAX69003°€¢
niobium
ACE700036
GCX610002
nitrates
NXX630001
nitrogen
ACEE€SQ00E
ACE660012
noble metals
ACB690022
ACDE€ECC17
ACDE7CO01¢
ACD670020
ACDETCC27
ACD680015
ACDEBCO1€
ACD680022
ACDEB(CO023
ACD690024
noiSe anal¥Yeis
ACA680012
ACR€9CC21
ACRLE9C028
ACAT700021
MDAT0CO032
nuclear analysis

Keywvord Index

IRF690014

MDA64O001
MDAGRL0006
MDC6900 1%

BFX680009
I2C660024
IAC6€0025
TACT7C0047
IARF6900 14
MDA620002

MCA64000 1
MDA69000 1

IAF6900 14
BaX700008

FCC7C0040
FCC700044
GCX610002
GCX68003C
LDA7C0015

GGX670034

GXX680039

FEE690034
FEE700027

ATD69003 1
ACD700024
ACD7C0038
EFX7000 16
CCXx680033
CXX700049
IAC7C0047
IAC7 10013
IBA710005
MCD6900 17

MDB700033
MDB7 10002
MDC68000S
MDC6900 15

ARX67C009
BAX680006
BBX67C012
BFX680009
BFX700016
IAF6S0014
MCA680004
MCDES0017
BLA620001
MDA6 20002
MDAE€4COC
off-gas systenms
ARXET001C
ACR660008
ACA6€001Y
ACA670016
ACAE67C023
ACR6EQO12
AC2680019
ACRESCO21
ACA690028
ACB67C017
ACLCE€COMN
ACL670026
ACD6EQO1E
ACL680022
operatinc costs
LEKX6 20003
cperaticn
ACA6500 10
ACA660008
ACAG6€CO14
ACA670016
ACA€70023
ACA680012
ACAGECO1S
ACRECCO21
ACR690028
ACA7CC021
ACC650012
IRBETCO43
JEX6E5C02C
1CC710024
LIX6£0023
1TX6700 13
cperators
ACA650010
MCBE€E£C022
Optics
JFX67C036
optimizations
BFX7CC056
KXX5¢00C2
NXX690057
oxidation
ACD680015
- ACDEE(C(Q22

MDA 640002
MLA€40006

- MDA €50001

MLAE60004
MDA €90002
MDA 700006
MLEC€70001
MDC €70002
MLC€90016
MDC 700004

ACD€80023
ACDE€90024
ACD700038
CCX€380033

HIX€60026

1ACE60025
TAC700047
IAC700051
IAC710013
JDX€70037

- JLX690060

MACEBOO34
MAX€50019
MCDE€Y0017

LIX€90008
MCA €60001
MCae80004
MCBE€S0021
MCB €50022
MCE€60029
MCB €900°54
MCB710012
MCD€80010
MCD€90017
MCD€30055
MCD €90062
MCL700001
MEBJ00003
MDC€70001
MLC€70002

MCCEB0005
MCCE700u4

NXX700057
NXX700058

ACDE€Z0024
ACD 700038
t

)

oxidation
*Continued*

ACE680017
ACEE€8CC24

Keyword Index

FBX6uooas

oxlde precipitation process

ACC70C037

. ACD660017

oxides
ACCEECO0E

oxygen

parametric studies

ACDE€70026
ACD680015
ACDESC022

ACC650012
ACD65001 1
ACDEECCTI
ACD660017
ACDE7C01¢
ACD670026
ACDEB0015

ACD680022

FBE690034

ACRBETCO24

performance

phase

AAXE7C003
ACA670023
ACBE6CC1S
ACB670017

-ACB€8C013

ACB680020
ACB€9C022
ACB690029
ACB70C022

~ACB700036

equilibria

ACD660017

- 'ACDETC026

'ACDEBO01S
ACD680022

 ACD€9C024

physxcal properties

~ACD690031

ACDT7CCO024

ACB€8C02C
ACBE€9C022

~ ACB690029

ACB70C022

~ ACB700036

ACD€ECC17

ACD670026

ACD€8001S
ACD680022

‘ACDESCO024

ACD69003 1
ACD700024
ACD700038

ACD6S0024

ACD690031
ACD700024
ACD700038
CXX700049

ACD690024
ACD690031

ACD700024

ACD7C0038

CAX68003z

CXX7C0049
EBX70004 3
LIX650023
LXX660031

FBE700027

TAC660024
TAC660025
TACT7C0047
TIAC700051
IAC710013

- IAD700052

MCD700001

MLB700003
MEC 700053

ACD700038

CAX690052

CAX690053
CXX640020
cXX700049
MDA64000Z

ACE690026
ACE690032
 ACET700025

CAX680032
© CAX690053

CCA680038

CLX670035

CXX640020
CXX700049
EBX700041
ELX640016

-EXX7C0048

FEX640015

ACE6€0012
ACE660013
ACE€7C028
ACE680017
. ACEGECO24
Fiping
ACAE€COCE
~ ACE680020
Flans |
: AAX670003
AAX6€7C0CS
AAX670006
AAX670007
- AAX67CC0CS
AAX670009
AAX67C01C
AAX670011
ABX€LOOCY
ABEX670049
GDX7 10025
HEX690058
HBX6¢COE9
I2aA€50024
glant
ACCE8E0012
IAC700051
JAB6SC018
JAET700017
plutcniurn
ACC690030
ACC7(C0023
ACL700024
BAX7C00CE
BEFX700016
Flutcniur flucrides
ACL680022
ACDESCO24
- ACLC6SC031
ACL700024

potassiup flucrides

ACD690024
ACDE€SCO024
pouer costs
AAX€7C0C3
BAX68000¢
BFX700056
IACEECOZ4
1AC660025
IAC7C00L47
TACT700051
prec1pltat10n
ACD650011
.ACD6€E0O 11
ACD68V022
ACDECQ024
ACD690024
ACD6S0031

Page 187

12C660025
I2C700047
IAC710013
MLE700003
NXX630001

NXX€30001

IABE70043
JDX€E90060
JEX 650020
LIX€50023
MCA €6 0001
MCAEB000Y
MCB 650021
MCBE6D029
MCB710012
MEAEL0005
MEB €80001
M EX 700020
NXX590002

LIX650023
LIX€70013
LIX690008

IAD700052
LBX€80027
LXX 710021

'MCD 690017

CAX €80032
CAX€E90052
CAX710023

ACD€90031

CRX€90053

IAC710013

“IAC710014

NXX €90057
NXX7500057

U NXX 700058
- NXX700060

ACEE50014
ACE€70028
ACE€80017
ACE€80024
FEA€80029
FCC700040

 
 

Page 188

precipitation
*Continued*
FCCT70C0uy
pressure
- MDC€90003
primary salt
ACB680020
ACB€9C02¢S
ACD700038
CAXE€BC032
primary systenm
MCRA€€0001
procedures
- MBX700002
MCAEE€CO01
MCB650021
MCB€5C022
MCB660029
MCB€9SC0S54
processing
AAXETCO04
AAX670010
ABX690007
 ABX70C0S54
ACB660009
- ACBE€ECO1S
ACB670017
ACCE€ZC00e
ACC650012
ACCEECCIC
ACC660016
ACC67C01E
ACC670025
ACCE€BCO014
ACC680021
ACC€9C023
ACC690030
ACC70C023
ACC700037
ACDE€80015
.ACD680022
ACDE€SC02Y4
ACD700038
ACEE€9(002¢
BFX700016
CAX€B8CO032
CAX69006 1
procurerent
ACE660012
ACE€6(CC18
ACEETC028
ACE680017
ACE€8C024
ACE690026
production
LAX710019
proqress regort

Keyvord Index

CLX700010
CXX700049

HCX7 10022

MDA 700006

MCB7 10012
BCCO670044
MEC660002
MEA640005
MEE66002E
MEB680001

GDX690042
IARR660030
IAC660024
IAC700051
IACT7 10013
LCA6€E0008
LCB710007
LCC710024
LIX650023
LIX670013
LIX690008
LIX690009
LJX6€0006
LJX660032
LJX670032
LKX620003

LKX700030

LKX710001
LXX660031
LXX700029
LXX700029
LXX710021
LXX71002€
MAX650019
MDA700007
NXX7C0060

ACE690032
ACE700025
ACE7C0039
EDX6400 16
FBX6U40015
LAX710019

ACE660015 ACEEB0017
ACB67C017 ACEEB0024
ACE670024 ACE€90026
ACBEECO13 ACEE€90032
ACE6800 20 - ACE700025
ACB690022 ACE700039
ACEBE€SC02S ACXE40008
ACE700022 ACX €40014
ACB7C€003¢6 ACX€50003
ACE7T 10029 ACX €50005
ACD6E00C? ACX€60007
ACL6500 11 ACX 660013
ACD6 60017 ACX €70015
ACL69C01¢S ACX€70022
ACD670020 ACX €80011
ACD67C027 ACX€B0018
ACL680015 ACX €90020
ACD6ECOD22 ACX€90027
ACD690025 ACX 700018
ACD7C0024 ACX70003¢
ACD700038 ACX 710027
'ACE650008 EDXEL0016
ACEGECO14 FBXE4 0015
ACE660012 FCX€90033
ACEGECO18 FCX700026
ACE670021 MEX€40003
ACE67002¢
protactinium
AAXETCOCHY ACD€90031
ACE660015 ACL700038
ACCE6EQCO14 BFX700016
ACC680021 CXX700049
ACC€SC023 IAAE6CO030
BCC690030 IACE€60025
ACD6ECO11 LBX€80027
ACL660011 LLCA€ES0013
ACD6€ECC17 LJX670032
ACC670019 LKX €20003
ACDE7C02¢ LKX710001
ACDH580015 LXX€60031
ACDEECO22 LXX710021
ACL690024
protactinium flucrides
ACL660017 ACDEY0024
ACDETCC19 - ACD€90031
BCL670026 ACD700038
ACD6ECO1E CXx7C0049
ACL680022 - LLAa7000u46
ACD6SC024 B
Frototypes ' _ '
|  ADX690063 HEX €20007
HBX€200Cé6 HXX€40019
FURES
~ AAX67C0QC7 - ACB€70024
ACA660008 ACB€80013
ACA71C028 ACBE€B0Q20
ACB6C017 RACB690022
-
-

pumps
*Continueds
ACB€9C026
ACB700022
ACB70003€
CCX680033
HBX€20006
‘HBX670042
HBX€9C(05¢
HBX690059
HBX700012
pyrocarhon
ACE70C02¢
-EXXT700048
gquality assurance
"HBX€9005¢
radiation damage
ACB67C024
ACD670020
ACD670D27
ACDEB0023
ACE650008
ACEESCO14
ACE660012
ACE€60018
ACE670028
radiation heating
ACB690029

AAXETCOOE
JDX670037
radioactivity
MDA700007
radiolysis
- ACA€6CO1L
ACAT700035
ACDESC007
ACD670026
rare e€arctks
ARX670009
ACCE€€CC1C
ACCETCO1E
- ACCK70025
ACCEB80014
‘ACC690030
ACC700023
ACD660011
"ACDEE€CO17
- ACD670019
ACDE€7C026
- ACD6870027
~ ACD€BCO1E
- ACD680022
ACDE€9C024
ACD690031
ACD70C024
ACD700038

Keyword Index

IAC660025
IAC700047
IAC700051

- TAC710013

"IAC710014

MAX650019

. MCD6%0017
NXX 630001

MAD690004

MBX700002

ACE680017
ACE680024
ACE690026
ACE690032
BCE700025%

CXX700049 .

ECX7 10011
TAB670043
MLCA680003

- IBRAT10005
radiation measurenment
MAC680034

ACD680023

 CEX640018
CLX7€0010

EFX700016
CAX69005Z
CDX670035

CXX700049

FBX640015
GGX670034
LCA670014
LCA680008 -
ICA690037

- LCB680007
1LCB710007
LCC710024 -
 LDA690012

LDATO00046
LKX7€0030
LKX7 1000 1

LXX660031

MDC67000 1

ACD700038

rare gases

ACL660017

-ACD&?GOZG

ACL680015
ACD6ECO22
ACL680023
ACDESCO24

reaction rates

ACD6E0O1S

-ACL680022

ACDESCO24

"ACD690031
reactivity

AAX€TCO01C

ACA650010
- ACA670016

ACRA670023
ACA670023

ACA6EC012

ACR690021

- ACAT7(C021

BAX680006
BBX670012
BFX680009

BGXETCOUS

¥CA680004
MCD6ECO1C
MCL6<C017

"MCL700001
reactor vessel

‘ACB6700 17
ACB6ECO13

‘ACEGGOOZC
reactors '

- ABX7(CC0E

} ACE680020
ACB6SC022

ACE700036
ADX€4C021
ACX670046

~ BAX680006
BFX7CC0S6

CAX680032

' _IACBEOOZQ

I14C660025
IACIC0047

- IACT700051
- IAC710013

recomblnat1on

- CEXe64C€018

reduction

ACD6E€0011
ACL670026
ACDEECO1E
ACL680022
ACD6SCO24

Page 189

MTCE€70002

ACD€S0031
BFX700016
CXXx 700049
MCD €80010
MCD€90017

- MDA 700007

ACD 700024
ACD700038
CLX7C0010
CXX 700049 -

MDA 620001
MCAE20002
MDA €4 0006
MCA€60003
MDA €8000 3
MDA €90002
MLCA€90005
MDA 700006
MECE60002
MDC 670001
MLCE70002
MDC 680002
M LC€80005
MDC €9000 3
MIC 700004
MLC 700005

- ACB€90022

HAX700050
1ABE€7C043

IAE700059
IAF€70047

 IAFE€70048
" IAF€90014

KBB €90006
LKX620003
M AX €500 19
%DB 700003
NXX 590002

- NXX€90057"
NXX700011

N XX700057

NXX 700058
 NXX7€0060

~ CLX700010

ACD 690031
ACD700038

"LAX700013
LGX €50002

LIX€90009

 
 

 

Page 190

Keyword Index

reductive extraction process

ACCEECO1E€
ACC67002%
ACC680014
ACC€8C021
ACC690023
ACCE€9C03¢C
ACC700023
ACC1700037
ACD670019
ACDE7C02€
ACD68B80015
ACDELCD22
ACD690024
Celiakility
ACE660018
ACEE€T7C02¢€
ACEE€80C17
ACE680024

Tremote maintenance

ACR660008
ACA€6C0O14
ACAETCO1€
ACA670023
ACA71002¢
ACE690032
ACE70002°¢
Femote ¥elding
AAXE€7C011
ACE680017
replacenent
ACB€7C017
CAX680032
Lesearch
AAX670005
ACDEECC17
ACDE€7C02C
ACD670027
ACDEBOC1E
revievs
AAX€7C003
AAX670005S
AAX670006
AAXETCO007
ABX580001
ABX€4C004
ABX670049
ABX€8C03E
ABX690007

ABX69CO05€ -

ABX700054
ABX70C05¢
ABX710020
- ACYXE4CO14
BBX670012
CXXE€H002C

ACD690031
ACD700024
ACD700038
GDX710025
LDA690012
LDA690038
LKX700030
LKX710001
LXX66003 1
LXX700029
LXX7 10021
LXX710026

ACE6900232
KCD700001
MEC700053

JFX6€0027
JFX670036
MCD690062
MEB660028
BEC7C0053
MFX700020

ACE680024
KEE690006

CAX650052

ACD680022
ACD690031
ACD7C0024
ACE69003:

EDX640016
EDX680031
EDX 690051
FBX640015
FCX690033

FCX7C0026 -

FXX690047
HBX670042
IED680036
MDA67C038
MDE700003

- NXX590002

NXX630001
NXX690046
RXX700011

'NXX7C0060

rupture
ACE650008
ACE6500 14
ACE6€C012
ACE660018
ACE670028
ACE680017
safety :
AAX670010
ABX6€C03S
REX690056
ABX7C00c4
BGX670045
CAX6%00€3
1IX650023
LIX€7¢013
FAC6EQO3Y
safety limits
MCBESCOCSY
samplers
ACRAGECO1Y
ACR€7C023
samgling
ACD6€C017
ACL670019
ACDE€C02C
ACL670026
ACDGT7C027
ACL6800 15
ACD6800 16
ACD6ECO22
sealing
ACEEESCOCE
ACE670028
ACE6ECO17
ACEEECO24
ACE690026
seccndary salts
ACC670019
ACDESCO024
ACL700024
CAX690053

secondary systems

GFX660023
JCYX€<SCO19
separations
CAX6£0032
CAX690061
CXx7€c04s9
LCA690037
shielding
MCA640006
MDAT7C00C?
MER6 40005
shrinkage
ACE670024

ACEE€80024
ACE€90026
ACE6€90032
ACE700025
EBX700041
FEE€30034

MCBES50021

- MCB€90054
"MCA€E40001

MDA EL40007
MLCA€70038

- MDA 680003

MLA€90005
NXx€30001
NXX700011

ACA €80012
ACA €80019

ACDE€90024

ACDES0031
ACD700024

ACD700038

CXX7000u49
JLX€70037
MAX €50019

ACE690032
ACE700025

FCX€90033

FCX700026

HCX710022
JCX€90019

" NXX€30001

MCAE60001

LCB710007
LCC710024
LXX€660031
LXX 700029

MERE30001
MEC 700053

o
o

simulation

JAB€9CO18
JAB700017 MDA690005
JABT71C008 MDA710003
. JCX690019 NXX700057
MCCEECCOS ' S
single-fluid reactors -
"ACBESBCO13 BFX7C0016
ACB680020 CXXT00049
ACB690029 HCX710022
BFX€ECCOS -
sites
IAC71C013
sodium
FCD71C01¢€
sodium fluoride . -
ACCEECO0E ~ ACD7C002¢4
ACC660010 ACD700038
ACD680015 ADX640021
ACD€8C022 CAX690053
ACD680023 LIX690009
ACDE€S0024 LJIX660006
~ ACD690031 LKX620003Z
solidus |
ACD670026 ACD700024
ACDEECC22 CRX69005z2
ACDE€9C024 CAX690053
ACD690031 CXX700049
solubility | o
- ACC690023 CAX690052
ACC€9C03C CAX710023
ACD670026 CCX680038
ACDEBCO15 CXX700049
- ACD680022 FBLC690036
ACD€9C024 LDA690013
ACD690031 LDA7000 14
ACD7C0024 LDA700015
ACD700038 P
specific keat T
- ACB680020 cCcxe80038
ACDEECC17 EDX6400 1€
ACDEECO1S FBX640015
"~ ACD680022 MCB700003
- ACD7C0024 NXX630001
CARX650053 e
Specific inventory - =
" BFX680009 EFX700016
speciticaticns o
| - ACE660012 FEX640015
. EDXE4CO016 GGX670034 .-
spectrophotometry . . -
ACD670026 ACD690031
ACD€8C022 ACD700024
- ACD6920024 ACD700038
spheres ' -
ACB680020

sprays

Keyword Index

MDA6600OY

ACB70002Z

stabil

CCX€ECO33
ity

AAX67000S

AAX670010

ACA€E€CCCE

stack

stainl

startu

ACL670026
BGX6T7CO45
Cax6.80032
CAX690053
JAB6SCO18
JAB700017
JCYXEe<CO01S

JDX€7C037
ess steels
ACEE6SCO4
ACE6600 12
ACE660018
ACE€C021
ACE670028
ACE6EC017
ACE680024
ACEECC02¢€

P
ACAGECOCY

ACA65CO1C
AC2660008
 ACAGECO14

stean

steanm

MCE650022
cycle
ACB660015
HCX710022
generators
ACBE€ECO2C
ACE690022
ACB6SC02S

 ACE700022

ACB71C02¢

ACEb6S0032

ACE7(C0025

“ACE7CC039

st€an

storag
“ TCLX7€001C

. stress

systers
AAX67C007
ACE660015

ACB€<SC022 -
IaC660025

IACTCO0LT
e | )

BCE7 10012

ACAEECCCE
ACE670024

- ACD67C0ZC

ACD670027
ACE€S5C02¢

Pagye 191

MCA €80004

‘MCA€50001

MDA €90002

MEA€EI0005

MDA 710003
MLB710002
MDC €60002

- MDC 700005
NXX€30001

ACE700025
ACE700039
GAX €70033

GAX€80028

G2X 700045
GCX610002
GFX€60023

- MCC €60005

MCD€90055
MLA660003
MDA 690005
MIC 680002

TEBE70039

CAX€90053
HCX 710022
IAC 660025
IAC700047
IAC 710013

- IEB€70039
‘IBB710015
- NXX630001

 IAC700051
IAC710013

TAC710014
IEB710015
JCX€90019

MDA €40002

MLA700007

ACE€90032
ACEJ00025
ECX710011
FEB700031
HCX710022

 
 

 

Keyword Index -

Fage 192
stress rupture
ACDE€7C02(C ACE6600 18
ACDE€7C027 ACE670028
ACE650008 ACE6800 17
ACEEE0014 ACE680024
ACE660012 IAB670043
structures
ACB680020 IARC660024
ACBE€9C022 IACT700047
ACBESCO29 IAC710013
ACB700036
surface tension
ACD690024 CCX680033
ACD€9C031 CCX680038
ACD700024 o
surveillance
ACD670019 ACE6B0024
ACD€7CC26 ACE690026
ACD€8CO022 ACE690032
ACD690024 ACE70002%
ACDESCO31 FBE670031
ACD700024 FBE690034 .
ACD70003€ FBE7C0027
ACE650008 FBE7 100 18
ACE€5C014 FBX640015
ACE660012 MAD69000Y
ACEE€C0O1E MCA660001
ACE670021 MCAS80004
ACEE€7C02¢E MCB710012
ACE680017 ~
systerns
AAX670008 JAB700017
JAB€ICO18 NXX700057
tantalum
ACE650008 GXX 680039
ACE7CC039
technology
AAXETCO03 ABX700054
ABX580001 ACD700024
ABX€8CC35 ACX64L0014
ABX690056
temperature measurerent
AAX670008 ACB68002C
ACA€7C023 MAX650019
tensile properties
ACE660012 FRE65001¢S
ACE68C017 FBE650016
ACE680024 FRE670031
ACET70003¢ ~ FBE670031
EBX690039 FBE6B8002€
EBX700042 FBE6S0034
EDX640016 FEE69004 4
FBBEEC(21 FBE7€0027
- FEB660022 FEE7 100 18
FBBESCO4C FBX640015
FBB690041 FCC7100 10
FBB70002¢€ FCE710004

FBCE€1COCH
FBC640017
FEC650017
test tacilities
" ACE680020
ACB€S0022
ACF690029
ACB7(0022
ACE7(C003¢
ACL680015
ACD6ECO22
ACL700038
testing
ACR650004
 ACAGECO1C
ACA660008
ACDETCO1S
-~ ACL€7C02¢6
ACD6800 15
ACD6ECO22
ACL680023
'ACD€%0031
ACL700024
ACD7C0038
ACE650008
ACE6EC014
- ACE660012
ACE6ECO1E
ACE670021
ACEE€7C02E
" ACE680017
ACEE€ECO24
ACEECSCO2¢6
ACE690032
ACE?7C002¢
theory , '
MDC€<COC3

thermal conductivity

ACBEECOZ0
- ACE690022
ACB€¢(C02¢S
ACE700022

ACD6E0022

ACE7(CO3S
CAX690053

therral ccrvecticr

ACD670020
ACDe7C027
ACL6800 16
ACE6SCO1yY
thermal effects
ACB€T7C024
EBX700041
FBE710017

thermal insulation

AAX67C0C7
AC2650004

GAX€70033

GAX680028..

ACE€90026

ACE€90032

ACE700025
GDX710025

"HEX€90058

HEX €90059
IAE700059

CLX€700135
EBX€90039
EEX 700041
FBB 700031
FEX€40015
HEX €20006
HBX€90058
HEX700012
HCX680037
HEX €20007
HIX€60026
BXX €40019
JDX€90060
LCB710007
LAHXE€9001
MEX 700002
MCAE60001
MCB710012
MCC €60005

- MLCE€60002

MLCC700004

CCX€80038
CXX 700049
£DX€40016
EXX 700048
FBX€40015
NXX€30001

ACE€60012
FBD€90036
GCX€10002
GFX€60023

FCC710010
MLA€20001
MDA 620002

ACBE€90022
NXX€30001
-

thermal pover
ACBE€ECO1E
MDB700003

therral prcrerties

. ACD€8C01E
ACD680022
ACD700038

thermal shield
MAXESCO19

thermal shock
EBX70004 1

thermodynamics
ACB660015
ACD70C038

thorium
ACC€80014
ACC680021
ACCES0023
ACC690030
" ACDE€7C02¢€
ACD680015
ACD€8C022
ACD69002%
ACD7CC03¢
BAX680006
BAX7CCO00E8

thoriur fluorides

ACDET?C01¢
ACD67002¢€
ACD680O0 15
ACD€8C022
ACD690024
ACD€90031
ACD700024
titaniurm
FCC690048
tools .
MEAEUCCOES

~ MEB660028
~training |

ACA650010

MCC€€C00E

tritium
ACAT700021
ACB700022

ACB71C02¢
 tubjng" '

~ NXXE€3C001
tungsten ,

. ACE®50008

-~ ACEE€CO1E

ACE690032

two-fluid reactor

AAX670005
AAXETCCO7
ABX€40004
ACB660009

Keyvword Index.

¥DC660002"

EBX7C0041

HCX7 100272
MDB700003

FAX 620004

FAX620005

- BBX670012

EFX7000 16

- IAA660030

LT26900 12
LDA690038

‘LDA7000 14

LDAT7C0015
LJX66003¢
LKX7C0030
NXX690046

ACD7C0038
CAX680032
CAX69005z2
CXX7C0049
LCA6S500137
LDA7C00U6
LXX7100z¢

FCC690049

MEE680001

MEC70005 3

wcce70044

ACD7000284
~ TAC710014
NXX700011

"ACE700025
- GAX67C033

GAX680028

HBX670042
TAA660030
IAC660024
IAC66002¢

ACB6€CO1E
ACE670017
ACB670024
ACC6€CO01C
BEX670012
uranium
ACC650006
ACC6€CO1C
ACC6700 18
ACC680014
ACCEECO21
ACL670026
ACD6ECO1E
ACD6 80022
ACDEECO2Z3

uranium fluorides

ACD6€CO17
ACL670019
ACD67002C
ACL670026
ACD67C027
ACL680015
ACDEECO022
ACD680023
 ACDE€SQ0024
uranium-~232
ACD7(C0024
LAX6900 10
- LAX700013
uranium-—-2<33
ACA670023
"ACA6ECO12
ACA680019
ACAESC021
ACA690028
ACC67002¢
- ACC&€EOQOO1L
ACC680021
- ACD7C0024
- BAX680006
BFX7C0016
- LAX690010
uranium-23¢
- ACD670020
ACD7CC024
CXX700049
MCD6ECO1C
 MLDA640002
MDA650001
MDAGECOC3

utilities

ADYX€ECSCOE3
valves
- AAX670007
vapcr pressure
ACL660017
ACD&7C0Z6

Page 193

IAC 700047

IEB67C039

LJX €60006
LJX€70032

ACD€S0024
ACD 700038
E2X 700008
LBX €80027
LDA €90012
LIX€70013
LIX€90008
LJX€60006
LXX710021

ACD690031
ACL700024
ACD700038
CRX €80032
CXX700049
LAX 710019
LGX€50002
LIX €90009

MDA €90001
MLAT700007

LAX 700013
MAD 690004
MCAE80004
MCD 690017
MLAEB0003
MDA 69000 1
MLA€90002
MLA€90005
MDA 700007

 KLC700005

NXX€90046

MLA700007

"MDC€7C001

MLCE€70002
MDC 680002

- 'MLCCEB0005S

MDC €90003

NXX7C0060

NXX€30001

ACD€80015
ACD 680022

 
 

 

Page 194

vapor pressure
*Continued*
ACD7CCO03¢
CXX700049
vibration
HCX680037
viewing devices
JFXEECQ027
viscosity
ACDE€0C17
ACD680015
ACDE€E0022
ACD690031
CAX€9C053
void fractiomns
ACBE€7CC17
ACB670024
ACBEBC02C
- -ACD69003 1
CCX€80033
MCD690017
volatility
ACB670017
ACCEECO1C
ACC660016
ACCET001€
ACC670025
ACCEBCOTY
ACC680021
ACDEB8001S
volume fractions
ACBE€7C017
vaste disposal
ACB590029
vastes
NXX700011
weigh cell
ARAX670008
velding _
ACB€9C022
ACB690029
ACB7CC022
ACB700036
ACEE€EECO0E
ACE650014
ACE6€C012
RCE660018
ACFETCO2¢€
ARCE680017
ACE€8C02¢4
ACE690026
wetting .
ACD680023
"ACDESCO02¢
X-rays
ACE670028
ACEE80017

Keysord Index

NXX630001

HCX7 10022

JFX670036

CCX680033

CCX680038
CXX700049
NXX630001

MDA7C0032
MDB700033
KDC 680005
MCC69000 3
KDC690015
MLC6900 16

ACC680022
ACD690024
ACD700038
LCA67001%4
LCA6€0008
LCR690037
LCC710024
LKX620003

ACB670024

CLX700010

ACE700025
RCE700039
FBA660020
FBE6E500 1€
FEB660021
FBB66002Z
FBB690040
FBB69004 1
FBB700028
FEX6 400 1S
GGX670034
MEX70000%

EEX700043

"ACE69002¢€

ACE690032

- ACE680024
X€enon
ACA680012
ACRESCOZ2E
ACA700021

 ACB670017 -

ACE670024
ACE680013
ACB6EC02C
ACE690022
ACDE€C017
Zirconium
ACD6C0024
ACL690031
ACE6€C01E

zirconium fluoride

ACD67C01S
ACL680015
ACD6EQ022

ACL690024 -

ACD7C€003¢8
CXX700049
REC ' |
AEX700055
BFX7C00%6
NXX590002
ARE
ABX580001
BEastellcy N
AEX580001
ACDEECOCT
ACD660017
ACD670020
ACLC670027
ACC680015
ACD6EQOD22
ACL700038
ACE€ECCCE
RCE650014
ACE660012
ACEEECO1E
ACE670021
ACE670028
- ACE680017
ACE€EC0O24
ACE690026
ACE6SC032
ACE7 00025
ACE7C0039
ADX640021
CAXEEQ032
CXX700049
FAX€z00CH
FAX62C0CE
FAX690035
-~ FAX€SCOLS
FBA660020

EXX 700048

CLX €70035
EDX€80031
ELX €90051
IBD€80036
MCA€70040
MLAE€70041
MDC €70001

MECE70002

- ACE€70028

LKX 710001

LAX710019
LCBEBO0OT

 LCB710007

LCC71002¢4
LDA 700046

NXX €90057
NXX700058

AEX 700054

FEA€80029
FBB 650018
FEB660021
FBB€60022
FBB€90040
FEBEJOO4 1
FBB7C0028
FEB700031
FBCS90001
FECE€10001
FECE€40017
FBC6€50017
FEDES0O0 36

FBE€50015
- FEEES0016

FBE€60019
FEEET0029
FBEE€70030
FEE€70031
FBE€80025

"FEE€80026

FBE 690034
FEE€90044

- FBE700027
- FEET10017

FEE710018
FBXE€40015
FCD710016
L))

L

Hastelloy N
*Continued¥*

HTGR

LMFBR

LMR

LWBR

MSBE

MSBR

FCE690043
FCX690033

"FCX70002¢6

FXX690047
GAX€70033
GAX680028
GCX€8C03C

JDX€9006C

CAX€90053
NXX690057

NXX590002
NXX700060

ARX670003
ACB€8CO02C

- ACB690022

ACBESQ02¢
ACB700022
ACB70C03€

AAX€7C003
ARX670004
AAXE67C00€
AAX670007
AAXETCOO8
ARL670009
AAXE€ETCO1C
AAX670011

ABXELCOO04

ABX670049
ABX€EBCO3E
ABX690007

.;ABXE9GOSE
- ABX700054
ABX71002C

ACB660009

 ACB66C01E

ACB670017
ACB€7C024

- ACB680013
- ACB€8002C

'ACB690022

" ACB€9C02¢

ACB700022
ACB70003¢
ACB710029
ACCEECCI1C
ACCH660016
ACCETCO1E
ACC670025
ACCEECO1Y4

Keyword Index

HCX7 10022
IAB670043
IAC710013
LCB680007
MAD690004
MAX6500 19

 MDA680003

NXX700060

NXX700058
NXX700060

CAX69005¢%
HBX690058
HBX 690059
IAE700059

TEA710005

CAX690052

CXX700049
ECX710011
EDX680031

EDX690051

GDX69004 2
GDX710025
HRX 67004 2
HCX710022
IAA660030

TAC660024 =
. IAC660025

TACT7C00u47?
IAC700051

IAC710013
IACT7 10014
IAD700052
~ IAE700059
IBA710005
IEB670039
IBD680036

JART7 10009

JAB690018
JAB700017
JAB710008 -
JCX690019

LBX680027
LCA6700 14
LCA6E0008
LCA690037
LJX660006

MSRE

ACCEECD21
ACC690023
ACC690030

"ACC7C0023

ACC700037
ADX6S0063
BAX680006
BBX€7C012

- BFX680009

BFX7C0016
BGX670045

AAX670008

AAXeiCo11

AEX6 40004
ABX€7C04S
ABX690007
ACA€EESQ0CH
ACA650010
ACAGECOCE
ACA6 60014
ACA67C016
ACR670023
ACAEE0012
ACA680019

ACA6C0021

ACA690028
ACAT7C0021
ACA700035

ACA710028

ACB7 10029
ACCe€E:00Ce
ACC650012
ACC67CO1E
ACCH670025
ACCEECO1Y
ACC680021
ACC6SC023
ACC690030

ACC7C0023

ACLC650011
ACDEECO 1
ACL660017
ACDETCO1¢S
ACL670020
ACD670026

ACE670027

ACDEECO1E
ACL6800 16

ACDGECO22
 ACL680023

ACDESCO2Y4

ACL700024

ACD7C0038
ACE660018
ACEETC021
ACE670028

Page 195

LJIX660032
LJX€70032
LKX 700030
LKX710001

- LXX 700029

LXX700029
NXX'€90002
NXXEI004U6
NXX €90057

- NXX700058
- NXX 700060

ACEE80017
ACE680024

~ ACE€90026
ACE690032

ACE700025
CAX710023
CCX €80033
CDX€70035
CEX€40018
C1X700010
CXX €40020
CXX700049
CXX 700049
EDX640016
FEB€50018
FBC €50017
FEE€70031
FBE€90034
FEE700027
FBE710018
FEX €40015
FCX690033
FCX 700026
GGXE7CO34
HAX 700050

"HBX€20006

HEX 700012
HCX680037
HEX €20007
HIX 660026
HXX €400 19

TAB€70043
- I3D700052
"JDX€70037

JEX €500 20
JFX€60027
JEX €70036

" LAX690010

LAX 700013
LAX710019
LCBE800OT

'LCB710007

LCCT710024
LHX€90011
LIX€50023

 
Fage 196

MSRE

*Continued#*

LIX670013
LIX€9C008
LIX690009

- LKX71€001

LXX7%0021%
MACEBCO34
MAD690004
MAXEECO1S
MBX640003
MBX700002
MCRA660001
MCA€80004
NCB650021
MCB€ESC022
MCB660029
NCB€90054
MCB710012
MCC€6CCOE
MCC670044
MCD€8CO1C

‘MCD6380017

MCDE€SO0055
MCD690062
MCD70C001
MDA62000 1

Keyword Index

MDR66000 3
MDAG660004L
HDR6700138

¥DA670040
KEDR67004 1

MDA680003
MDA690001
MDA690002
HDA69000¢
MDA 700006
MDA700007
MDAT700032
MDA7 10003
MDB700003

MDE700033

MDB710002
MDC6€0002
MDC670001
MLCC670002
¥DC680002
MLC68000¢
MDC690003
MEC690015
MDC690016
MEC700004

-ESFE

¥ak

MDA62Q0C2
MBA630002
MDAGUO0OCT

- MDA640002

MDAGULCOCE
MCA6 40007
MDAGEGOCH

AAXETICOCE
AAX670011
ABX64COCY
ARX670007
AAXEICOCE
ARX670009
AAX67C01C
AEX580001
ABX€4C00C4
AEX680035
ABX6<000Q7
ABX690056
ABX7CCOSy
ABX710020
ACXEeucocCs
ACX640014

CaAX690053

MDC 700005
MEAELO00S

MEB€60028

RAX €70008

AAXETQO011

AEX€40004

ACX€50003

ACX €50009 -
 ACX660007

ACX €60013

- ACX€70015

ACX€70022
ACX680011
ACX€80018

ACX€90020

ACX €90027
ACX700018
RCX 700034
ACX710027

CXX €40020

TAA€50024
MCDEJD055

NXX€30001
k1]

CATEGORY INDEX

'The category structure used in the MSRIS appears in outline form on

the next three pages, followed by the index.

The significance of ‘the letter X, which does not appear in the outline

but does appear in category designations elsewhere, requires some expla-
nation. - | |

There are two meanings or uses of the letter X. One is simply as a
"filler." For various reasons associated with the computer, every category
designation, whether it be the primary category‘in'the identification num-

ber or an "other category” in the last line of the entry, must have exactly

3 letters., Thus for category'N which is not subdivided, every abstract is.

designated NXX, Similarly, all abstracts falling into second-order cate-

.gories which are not subdivided have X as the last letter in their category

designation (ABX700054 for example)

The other use of X is to denote a general or broad treatment. This

is the meaning if X appears in place of the letters for established second-

or third-order categories. To illnstrate,'MAX690019 is a comprehensive

description of the MSRE design'(category MA) which discusses general con-

‘siderations and embraces information in third-order categories MAA, MAB,

MAC, and MAD

 
198

SUBJECT CATEGORIES IN MSRIS

 

 

A Molten-Salt Reactor Programs D Analytical Chemistry
AA MSRP -~ Plans and Organizations :
AB MSRP - Technical Summaries
AC MSRP - Progress Reports

ACA MSRE
ACB Large MSR's
ACC Salt Processing
ACD Chenistry
ACE Materizals
AD MSR Ac¢tivities Outside MSRP

 

E Graphite
| EA Fabrication
EB Unirradiated Properties
EC Irradiation Effects
ED Applications |

 

F Hastelloy N and Related Alloys
FA 'Alloys Leading to Hastelloy N
FB Standard Hastelloy N
FBA Microstructure
FBB Fabrication

FBC Mechanical and Physical
- Properties

 

B Reactor Analysis
BA Nuclear Data
BB Static Neutronics
BC Dynamics
BD Thermal%Effects
BE Activation, Radiation and

FBD Corrosion
FBE Radiation Damage
FC Modified Hastelloy N

Shielding
BF Fuel Cycle and Economics FCA Microstructure
BG. Safety FCB Fabrication

FCC Mechanical and Physical

BH Computer Programs Properties _

 

FCD Corrosion

C Reactor Chemistry
‘ FCE Radiation Damage

CA Phase Relations
CB Thermodynamics and Equilibria

 

G Materials Other Than Hastelloy N

CC Physical Properties and Graphite

CD Rates and Diffusion GA Stainless Steels

CE Corrosion Reactions GB Steéls other than Stainless

CF Fission Product Behavior GC Nickel and NiQBase Alloys other

than Hastelloy N
GD Molybdenum and Mo-Base Alloys
GE Brazing Alloys'
GF Other Metals
GG Nuclear Control Materials

CG Tritium Behavior
CH Oxide Behavior
Cl Crystal Studies

Surface Effects

 

CcJ
CK - Electrochemistry
CL. Radiolysis

 
T

199

Subject Categories in MSRIS
. (continued) )

 

 

HA
HB
HC
HD
HE
HF

HG
HH
HI

H Reactor Component Development

Core
Pumps
Heat Exchangers

Steam Generators

‘Gas Injection and Removal

Valves _

HFA Freeze Valves
HFB Mechanical Valves
Flanges

Heaters_’

Other Cemponents,

 

TA

1B

I Reactor Design;

Reactor Plant

IAA Early Molten-Salt Reactors

IAB MSRE B

IAC One-~Fluid MSBR (Reference
Design)

IAD Other Thermal Molten-Salt
Reactors

IAE MSBE D

IAF Fast and Epithermalfi
Molten-Salt Reactors

Systems

IBA Fuel

IBB Coolant

IBC Steam

IBD Gas

IBE Containment

 

J Instrumentation and Controls

JA

JB
JC
JD

JE
JF
JG

General

JAA Instrument Development

JAB Plant Control

Nuclear Control and Plant Safety
Process

Radiation and Contamination
Monitoring

Data Collection and Analysis
Communication and Surveillance

Electrical and Pneumatic
Systems

 

K Operation and Maintenance

- KA Operation

KB

KAA ARE

KAB MSRE . | _

KAC Other Molten-Salt Systems
Maintenance

KBA MSRE Maintenance

KBB cher Molten~Salt and
Radioactive Systems

 

 
 

Subject Categories in MSRIS
(continued)

200

 

 

 

 

L Fuel Preparation and Processing M MSRE
LA Salt Procurement and MA Design
Preparation MAA Plant.
LBV Fluorination _ MAB Méjot Component.
LC Distillation , MAC Instrumentation and Controls
LCA Experimental Basis fiAD Auxiliéry Systems and
LCB Engineering Development: Components
LCC Operating Experience MB Construction
Lb Reductive_Extraction | MC Operation
LDA Experimental Basis MCA Program
 LDB Engineering Development MCB Procedures
LE Metal Transfer MCC Training
LEA Experimental Basis MCD Experience
LEB Engineering Development - MD Analysis
LF Oxide Precipitation | - MDA Theoretical
LFA 'Experimental Basis MDB System Performance
LFB Engineering Development 'MDC Nuclear Performance
LG Adsorption and Reduction . ME Maintenance
LH Salt Purification MEA Principles
LI MSRE Salt Processing MEB Procedures
LJ Plants for Two-Fluid MSBR MEC Experience
LK Plants for One-Fluid MSBR MF Decommissioning
N Miscellaneous

 
Cateqgory Index Page 201

The Followirg Index States The Category And Gives Feferences
To Each Article which Was Keyed To It

ACX

CLX7€001C

| AAX o ACX64LCOCE ACX€80011
- AAX€7C003 RAX670008 ACX6 40014 ACX 680018
AAX670004 AAX670009 ACX6500C3 ACXE£90020
AAXETCOO0S AAX670010 ACX650009 ACX €90027
3 AAX670006 BAX670011 ACX€€00C7 ACX700018
AAXE€70007 ' ACX660013 ACX 700034
ABX ACX€7C015 ACX 710027
ABX580001 ABX7C0054 ACX670022
ABX640004 ABX70005% ADX .
ABXE7COUS ABX710020 ADX6 40021 ALX 630063
ABX680035 ACX 640014 ADX67C0U46
ABX€9C007 NXX590002 BAX -
ABX690056 NXX690046 BAX6ECOCE BAX700008
ACA . BBX |
ACR650004 ACA680019 BBX67C012 MDA €4 0006
ACAE€SCO1C ACA690021 MLDA6 20002 MTCE70001
ACR660008 "ACA690028 'BCX
ACA€6C014 ACA700021 . JAB6300 18 MCA€60003
ACA670016 ACA70003% JAB7C0017 MDA €90002
ACA€7C023 ACA710028 MCA640006 MLC€60002
ACA680012 ' MDAES00C1 |
ACB BF X
. ACB660009 ACB690022 BAXEECOCE BFX700056
ACBEECCTS ACB690029 BEX670012 MLAE70040
ACB670017 ACET700022 BFX6EC0CS NXX700057
. ACBE€7C024 ACB7C0036 BFX700016 .
ACB680013 ACBE7 10029 BGX
ACBERCO2( AAX670010 MLA660004
ACC BGX67C045 MDA €70038
ACCES0006 ACC680014 MDA640002 MLA€80003
ACC650012 ACC680021 - MDAEU4COCT
ACCEECCTIC ACC690023 RXX
ACC660016 ACC690030 AAX6700CS
ACCE€7C01¢€ ACC7C0023 CAX | -
- ACC670025 ACC700037 . CAX6EC032 CAX€90061
ACD | C - CAX690052 C2x710023
~ ACD650007 ACD68001E : CAX€SCQE3 :
" ACDESCO11 - ACD680016 -~ CBX -
ACD650013 ACTD680022 ~ LDA6SCO12 LDA 700015
~ ACDE€CO011 ACD680023 -  LLCA690038 R
- ACD660017 ‘ACD6900 24 ccx - : L
ACDETCC1¢ ACD690025 " CAX690053 CCX €800 38
ACD670020 “ACD690031 CCXEEC033 A
= ACD67CC2€ ~ ACD700024 - CDX
" ACD670027 'ACD700036 . ~ CDX670035S
. ACE _ S - CEX . , -
' ACE650008 ACE6800z4 CAX6SCOE3 CEX€40018
ACEE€SCO014 ACE690026 CFx |
$ ACE660012 ACE65003Z ~ ACAT7C003¢% CCX€E80033
o ACEE6CC1E ACE7C0025 CJdXx |
\fi# ACE670021 'ACE700039 CCXE6€C033
ACEE7C02E€ EDX640016 CIX
ACE680017 FEX 640015

 
 

 

Fage 202

CXX

DXX

ERX

EBX

EXX

FAX

FBA

FBB

F8C

FBD

FBE

AAX€7C005
ACD650011
ACDE€€CC11
ACD670019
RCDE7CO2¢€
ACD680015

ACD650011
ACDEECO1T
ACD670019
ACDE€7C02€

ACEEECO0E

CDX€7C03S
EBX690039
EBX70C041

ACE€S0014
ECX710011

ARX670006
ACE€7C021
EDX640016
EDX€8C031

ACEEEC012
ACE660018
ACE€7C02¢8
ACE680017

FAX620004
FAX€2COCE

FBAEE€C02C

FBBE6SCO1E
FBB660021
FBBE€€C022
FEBR69004O

FBB66002 1
FBBE€C022
FEC59000 1
FBCE10001
FEC640017
FBCE50017
FBE650015
FBE€SCO16€

FAX€2C004
FAX620005
FAX€9003¢S

FBBE€9COUC
FEB69004 1

Category Index

ACD680022
ACC690024
ACD690031
ACD700024
CXX640020
CXX700049

ACD68002
ACD690024
ACL&90031

ACD700024

EBX700042
EBX70004 3

FCX690033
FCX70002¢€

ECX690051
FCX690033
FCX70002€

" ACE680024

ACE700039
FXX7C00u48

FAX690013%

FRX690045

. PBA680029

FBB690041

FBER700028
FBB700031

FEE660019
FBE670029
FEE67003C
FBE680025
FEE68002¢€
FBE690044
FEE7 100 18

FRX690045

FED6900 36

FBE710018

- FBE680026

FEE690034

FBX

FCA

FCC

FCD

FCE

FCX

FXX

GAX

GEX

GCX

GDX
GEX
GFX
GGX

GXX

HAX

HEX

FBEE€SC01ES

FBE650016

FBEG6€CO1S
FEE670029
FBE67C03C
FEE670031
FBE€EQO2C

ACE6E00CE
ACE6500 14

- FCC700040

FCC690048
FCC€<C049
FCC700040

FCL7100 16

FBE690034
FBE7C0027
FEE7 10017

AAX670006
ACE7(C0039
FCX690033

ACE660012
ACF€€CO1E
ACE670028

ACE670021
GAX€E70033

- FBDESCO3€

FCD71001¢
GCXh 10002

GLCX690042
ACE650008
GFX660023
GGX670034
ACE650014
ACEE€0012
ACE660018
ACEeiC02¢
HAX7(C00EC
HBXEZ2COC6

HEX670042
HBX6<S00EE

FBEE€9004Y

FEE700027
FBE710017
FEE710018
GAX670033
GAX €80028

- ACE€7C021
FEXELO015

- FCC700044
- FCC710010

FCE710004

FEE710018
FCE€90043
FCE710004

FCX 700026

FXXE9COu7

ACE€B0017
ACEE€80024
FXX €90047

GAX €80028
GAX700045

GCX€80030

GLX 710025

ACEE€80017
ACE€80024
ACET700039
G XX€80039

HBX€S0059
HEX 700012
¥

HCX

HDX
HEX
HFX
HIX
HXX

IAA

IAD

IAE

IAF

1BA

1BB

IBC
I8D
JAA

JAB-

JCX

JDX

JEX

FBB650018
HCX 680037

HCX71€022

"ACB71C0C2¢

HPX€20007
HIXE€002€
ARXE€TCCO7
1AA 650024
IABET7C043

ACBE€60015

ACB670017

ACB€7C024
ACR700036

TACE6€C024

ACB€80013
ACB680020
ACB€9C022
ACB690029

ACB710029
ADX640021
ADXETCCULE
TAF670047

IBA710005
MDAETCCHU(

 IBBE7C03S

IBBETCC3S

IBD€EECO3E

JART1C00S

~ JABE9CO1E
 JAB700017
Jsx
, "JAB710008
MAC6BCO34

JCXESCC1C

JDXE7CC37
JDX690060

Category Index

HCX7 10022
IBA710005

IBD680036

HXX640019

- IAA6E60030

MAX650019

TAC660025
IAC700047
IAC700051
TAC7 10013
IAC710014

ACB7€0022
ACB7 10029

IAD7C0052

IAE700059

TAF670048
IAF690014

MCA670041

IBBT10015

JDX65006C

'aghbfl9007 '

MCD 630062

JEX

JXX

KAB

KEBA

KBB

LAX

1BX

ICA

1CE

ICC

IDA

LEB

LDX

o Lex
JAB710008
JCX6S0019

IIX

1JX

JEX6 50020
JFX6 60027
AAX670008
acA65000“

ACA65C01C
ACA660008

ACA6€0014

AC26700 16
ACA67C023
ACA680012
ACA6€0018
ACA690021

ACA65C028

ACAEECOCE
ACA660014
ACAETCO 1€
JFX660027

AAX670011

1AX690010

LAX7CCO13
LBXEEQD27?

ACCETC025
ACL660011

‘LCA6TC014

LCB6ECOCT

LCC710024

 ACD6€CO 11

CAX680032

N ‘LD§6§C012

1CA690013

- GDX710025
1XX700029

| - LGX650002

LHX .
" ACA680019

~ ACCE500CE

LCC710024

[IX650023

LIXETCO13
{

ACCEECQ1C
RCC660016

JFX €70036

ACAT700021
MCA 660001
MCAEBOOOU
MCB 690054

MCDEBOO10

KCD€90017
MCDEJ0055
MCD 700001
MLC€80002

JFX670036
MEAELO00S
MEB €60028
MECT700053
KEBE€900Oe6

LAX 710019
LIX690009

LCA €80008
LCA€900 137

LCB 710007

LXX710021

. LDA€90038
 LTA700014

LDA 700015
LLAT700046

1LXX 710021

ACCES0012
LHX690011

LIX€90008
LIX €90009

LJIX€6C006
LIJX€60032

 
 

 

 

 

Fage 204

LJX

*Continued¥

LKX

LXX

MAA

MAB

MAC

MAD
MAX

MBX

MCA

MCB

ucC

MCD

LJX670032

ACC690023
ACC7CC023

LEKX620003

LXX660031

LXX70CC2¢

IABE70043

LIX650023

HBX620006

HBX700012

HCX680037

JDX670037

JEXE€EC02C

HXXELOO01S
MAXESCO1S

ACCESCO0E
MAD690004
MBX€4CO03

MCAEECCO

MCBESC021
MCB650022
MCBE€CC2¢C

MCCE6CCOE

ACCeECO12
ACD650011
ACDEECOC11
ACD670019
ACDET7CC2¢€
ACD680015
ACDE€B0022
ACD680023
ACDE€SCO24
ACD690031
ACD70C024
LAX690010
LAX70CC13
LIX690008

Category Index

LKX6 20003

-LKX700030
LKX710001

LXX710021

LXX710026

LIX670013

HFX620007
MBX640003

MAC680034

MAD690004

MBX7C0002

"MEX70000%2

ECA680004

MCB6S0054
MCE7 10012

MCC670044
MCD68001C

MCD690017
KCD690055

MCL690062

MCD700001
MCE700033
MDB710002
MLEC68000:2
MDC680005
MLC690003
MDC690015
MLC6900 1€

‘MDC700004

MLCC70000%

EDA

EDE

MLCC

FDX

MEA

EEB

MEC

FFX

EXX

NXX

LIX€<COCS

MDR6 20001
MD A6 20002
MDA6ZC0C2
MDA640001
MDA6LCOO2
MCA640006
MDAELCOCT
¥DA650001
MDAGECOCI
MDA6 60004
MDA6GCO3E

MCDESCOE2
MDE700003

#DA660003
MDA7C00COE

- MPC660002

MDC€700C1
MEC670002
MDCEEQQC2

ACA€7C023
ACA680012
ACA6EQO19
MEAELCOCE

JFX6€C027
MEA640005

ACR670023

ACA6EQ012

AC2680019
ACA€S0021

ACA71002¢€

RCAESCOCH
ACRA650010
ACA6ECOCE
ACA660014
ACA67C01e
AC2670023

NX¥590002
NXX€3COCT
KXX690046
NXX65C0E7

MDAE70040
MLAETO041T
MDA €80003
MLA€90001
MDA €90002
MLA€90005
MDR 700006
MLAT700007
MDA 700032

- MLAT10003

MDB700033
MIB710002

MLC€B0005
MDC €90003

- MLC€90015

MDC€90016
MEC700004
MDC 7C0005

ACA€90021

- ACA€90028

ACA 700021
MEC 700053

MEB€60028
MEB€B0O0OO1

ACA€90028

MCD 700001
MEC70005%3

MFX7C0020

ACA€80012
ACA€80019
ACA€90021
ACAE30028
ACAT700021

ACA700035

NXX 700011

- NXX700057

NXX 700058
NXX700060
 

205

i

INDEX OF AUTHORS
wi

Following Is A List Of Authors Whose Documents
Have Peen Ahstracted In This Pukblicaticr

{Staff Regcrt)
ACBEECCO9
ACC660010
"ACCEECC1€
ACC670018
ACCETC02¢S
ACC680014
ACC€8C021
"ACC690023
-ACCE9003¢C
ACC700023
ACC700037
ACD650007

ACDESCO11

ACD650013
ACDEEQC11
ACE660017
ACDET7CCI1¢
ACD670020
ACE€ECCOE
ACE650014
ACE€€ECC12
ACX640008
Atrakhbimuv U
EBX70004 2
Adamscn GY
- ACE€€CC1E
Alexander 1G
TAAESC024
Altom DW
' ECX71GC11
Ball SJ ,

" MCC660005
MDAESCOO1
-MDA66000Y
MDAT1CCO03

Barbanov VN

- EBX70CC42
Barton CJ |
- CAX690052
Bauman HF .
- BFX700016
~ IAA€EECC3C
Bauman, CC
~ JDXE€9CO06C
Beall SE
- MDA640007
Beatty RL

Bell MJ
MDAT700007

Bennett LL
BBXE€7CC12

FCX690033

Personal Author Index

ACX640014
ACX65000 3
ACX650009
ACX660007

A€ X660013

ACX 670015
ACX670022
ACX680011
ACX680018
ACX690020

ACX690027
ACX700018

ACX7C0034
ACX710027

BGX670045

IBE67003S
NXX590002

 NXX690046

NXX6S0057

NXX700058 «
-NXX7C006¢C

GCX610002

MDB710002

MDC660002

MDC68000Z

I1AC660024
IAD7C0052

FCX700026

~Cantor S

Bettis CE
HCX7 10022
Bettis ES
ARX690056
EDX6EC031
EDX690051
TAA€€C03C
IAC660024
Blankenshiyr FP
CEX6 40018
Blumkerg R
~ AAX670011
MCDéSCOE2
MEAGU0005
Bchlmann EG
ACD6T70027
ACDEECO 16
Bokros JC
EXX7(C0O4E
Bolt SE '
LAX6€001C
Bond VE
NXX7C0011
Briggs BE
AAX€7C0C3
AEX€LCOCY
AEX700054

ACB6€0015

ACE670017
ACBEGCOZY
ACE680013
ACB680020
Buckner MR -

MCE7 10002
Burke OK '

JCX690019

Cancnico DA
' ACE670021

FBB690040

€CCX680038

‘Carlsmith RS :
- BBX670012

Carlson RG

FBC590001
~ Cearpenter JA L
ECX71C011

Carter Wl
’ - AAX6700CY4
- TAR650024
TAREGECO3C
IAC6E€0024
- "LCB680007

Page 207

- TAC€60025

IAC 700047
12C700051
IAD 700052

LXX 660031
MEB660028
MEB 680001
MEC700053

ACDEBOO23
ACD 690025

LAX700013

ACB€90022

ACB690029
ACB700022
ACB700036
ACB710029
IAB€70043
1AC700047

- MLCAE90005

FEB€90041

FBB700031

 BEX 700056

LJXE60006

L3X€60032
LJX670032

- LKX620003

LXX 700029

 

 
 

 

Page 208

Cathers G1
LBX€80C27
Cepolina AG
FBB700031
Chandler JH¥
LAX690010
LAX7C0013
Chang SIY
JAAG1000S
Chang SJ
ECX710011
Claiborne HC
MDAG40O006
Clark FH
JCX690019
Cochran HD
LCC710024
Cook #®H
ACE670021
EDX€4CC16
EDX€EECO31
Craven CH
BAX7CC008
Crouse ERS
GCX€1€C02
Crovley WK
HCX710022
Dahlke C
LDAEYCO3E
Deonigi DE
NXX700057
Dergunov NN
EBX70004 2
DeVan JH -
FAX620004
FAX€2CCCES
FAX690035
Ditto SJ

AAX670008

Donnelly RG
: FBB650018
Dyer FF

MCD69006 2

Eatherly &P
ACE690026
ACE690032

Edison GE
BBX670012

Fngel JR
JEX650020
MCREECCO
MCAEB0004
MCD680010
MCDESCC17
mcD700001
MDAE2CO01
MDA620002

Personal Author Index

LIX690009

MDR69000E

EDX 69005 1
FCX69003 3
FCX7C0026

IAR650024

FAX690045S
FCD710016

ECX 680031
EDX69005 1

MDA640001
MDAG64000 2
MCRA640006
MEA6€0004
MDA 680003
MLCC670001
KDC670002
MLC680002

Engle GB

EXX700048

Evans(III) RB
CDX670035
FAX620004

Ferris LM
CAX7 10023
LCA€¢CG037
104690012
LDA690013

Fontana A
EEX700043

Fraas AP

- IBB710015

Fredricksenr Ja
CAX690052

Fry ON
MCB700033
MDCEEQOCE
KDCESCOC3

Gabbard CH
MADESCOCH
ML 2680003

Gans W
1CA690038

Gat U
ADX67004¢€

Gehllach RE
FBA680029
FBE7 10017

Geist JJ
AEX7 10020

Gift EH
BEX670012

Gilliland RG
FEEH60021

Gilpatrick LO
CAX690052

Greenstreet WL
EBX690039

Grencn M
AEX710020

Grimes HWR
AAX670005
ACDE€7C026
ACLEE001S
ACL680022

- ACD6SC024

Grindell AG
AAXE€TC0CT
HBXE6<SCOEE

Gunkel E¥
FBB7(CO2¢€

Guymon FH
MCA6€0001
MCE650022
MCBEECO2S

- FAaX€20005

FCD710016

LLA700014

" LDA 700015

LLA700046

. LXX700029

HMICE90015
'MDC €90016

#4DB 700003

FCX €90033

FPCX700026

- FER€60022

ACDEY00 31

- .ACD 700024

ACD700038
CXX €40020
C Xx700049

- HBX€90059

 MCB 690054

MCB710012
MCC 670044

")
w}

~Guymon RH
‘*¥Continued*

MDA68B0O003
Hannaford EA

LCC710024
Haseltine DM

ACE670021

Haultenreich PN

ABX690007
ACAESCO04
ACA650010
ACAE€ECOCQE
ACA660014
ACAE7C016€
ACA670023
ACA€8C012
ACA680019
ACA€9C021
ACA690028
ACAT700021
"ACA700035
ACAT1C02€
C1X700010
MCA€6CCO1
Hightower JF
LCA€ECO0E
LCB710007
Hise EC
MEB680001

"Hitch BF
HIX660026

Holmes DK .
EDX680031
Bolz PP
KBB690006
Houtzeel A
MCDE€9COE2
Janney DB
- IAA€5C02y
Jensen RJ o
. ADX€9CC63
Kasten FR

‘AAX670010

ABX7CC054
-ACX640021
 BAXTICCO008
- .EDX680031
Kedl RJ
- EDX680931
EDX€E9CCS 1
HAX7CCCSC
Kennedy CEF
ACE€7C021
EDX680031
-EDX€9C051
Kerlin TW
MDAE€S0001

Petrsonal

MCR690054

MCD680010
MCD690017

MCD690055

MCD70000 1

MbA620001

MDA63000z2

MDA640001
MCA64000€
MDAG40007
MCA6€0004
MDA670038
MLA6B0OQOS
MDC680002
MEC70005:
MFX7C0020

LCC710024
LIX66003Z

EDX69005 1

. MCA670040

EDX690051 -
IAA660030
TAC6€0024 -
IAC660025

HCX 680037
MDA67CO4C.
MCA6T0041

FCX690033

FCX700026

MDC 660002

 

Author Index

MCA710003
MDB71C0C2
Kerr HT =
BFX680009
Kirslis SS
EDX680031
Knacke O
- "LCA690038
Knowles DG
JEX670037
Koger JW
FEL690036
'FCD71001¢
-FCX690033
Kohn HW ,
CCX680033
Kowalew A
IAFO70047
Kryter RC
MDC680005
Kurakin VM
EBX700042
Landis J¥
ABX700055
Lane JA
BFX700050
~ Laney RV
- REX700055
Lavrence JJ
! "ICA700046
Layman WH =~
‘BAEX700055
Lindauer RE
| ACC6500006
ACCoEC01¢
LCEGECOCT
IHX6900 11
- LIX6E£C023
- Iitman AFE -
’ 'ACE€7C021
FBL6900 36

Lundy 1S
- FCC690048
MacPherscn HG

ABX580001

- ABXETC0US
“Failen JC

LEX€E€ECO0Z27
Malinauskas A
: CDX65C03%
Fanly WD
GCX€100C2

Fartin WE
FBE6ECO1E
FEE6500 16

- FCX6390033

CAX710023

Page 209

MLCEBOOO2

ELX €90051

FCX 700026
GAX700045
GLX €90042

IAFE70048

MECES0015

LIK670013
LIX€900J8
LIX€90009

LJIX€60032

MDA E4 0007

FCX700026

G2X 7060045
GDX €90042

FCCE90049

 AEX€80035

LDAE90013

LTA700046

FBE€6G019
FEE€70029

 

 
 

Payge 210

McCoy HE L
AAX670006
ACE€7C021
ACE670028
ACEE€80017
ACE680024
ACEESC02¢
ACE690032
ACET70002¢
CAX690053
EDX€8C031
EDX690051
FBRAEEC02C
- FBA680029
FBBE€9CO4C
FBEB69004
FBB7CCO2¢€
McDonald ®B
IAAE6€003C
TAC660024
McDuffie HF
CAXE€9C0O53
McElroy DI
GCX€BCO3C
McGlothlanm CK
HBX690058

HCX€B8CC37

BcNeese LF
LCAE€7CC14
LCA68000¢8
1LCB680007
LC371CC07
LCC710024

) LGX€5C002

KcWherter JE
IAE70CCS¢

Metzger GE
GXX€B8C03¢

Mielcarski M
IAF670047

. Milford RP _

-~ LKX620003

Moore RL
AAX6700038
JFXEECC27

Muller F
LDA€9CC3E

Nelms HA
HCX710022

Nicholson EL
GDX710025

Nogueira ED
LDA690013

Peekles FN
IBDEBLO3E

Perry AWM
AAXETCOO0C

Personal Author Index

FEC640017
FBE670030
FEE670031
FBE680025
FEE680026
FBE690034
FEE690044
FBE700027
FEE7 10017

FBE710018

FCE690043
FCE710004
FCX690033
FCX7C0026
FXX690047
GCX680030

MBX640003

HI X660026

LEX690011

LJX660032
LKX7C003¢0
LKX710001
LXX7C0029
LXX710021
LXX710026

IAF67004€E

JFX670036

LXX700029

LLCA700046

EDX680031

BAX680006
BFX6ECOCS
BFX700016
Pickel TN
HCX7 10022
Poturaj-Gutniak
IAF670047
Price RJ
EXX700048
Prince BE
MDA6 20002
MDAG6L4COC1
ML 2640002
MDAG4COCE
MLC2660003
Ricci JE
CAX690061
Richardscn M
HFX620007
MEC700053
Robertson RC
ABX690056
CAX6¢SC0E3
ECX680031
EDX690051
IAAGECO3C
IAC660024
Rokertson RC
, IAC7 10013
Rctinson JC
MDAT7C0032
MDE700033
MDC6ECOQE
Fosenthal F¥
ABX€<COC?
ABX690050b
Ross RG
HIX€€00z¢€
Russell JA
JDX67C037
Futherford JL
CDXe67C03%
Schilling CE
LD2A7C0014
Scott CLC
LGX65C0C2

1JX66000€ -

Scott D
AAXE€TC0CT
CAX690053
EDX€€00 31

Sessions CE '
FCCec<Cousg
FCCe6<CO04S
FCC7C0040
FCC7(CO4y

m

(editcr)

ELX €90051
IAF 690014

IAF€70048

MCA€B0003
MDCE€70001
MLCE€70002
MDC €80002

MEX 700020

TACES0025
IAC700047
IEC700051
TIACT71001Y4
MAXE€50019

MDC€90003
MLC€90015
MDC€90016

ABX700054
AEX7C0055

‘LJX€60032

EDX€90051
HXX €40019

FCC 710010
FCX€90033
FCX 700026
 

 

Shaffer JH
LAX71C016
Shaw ¥
~ ABX70005¢%
Sides WH
JABESCO1E
JABT700017
Siman-Tov M
HCX710022
Slauqhter GM
FBB700028
Smith AN
MCBEECQ21
Smith FJ
CAX71C023
LCA690037
Smith J®
EBX690039
Smith OL
BFXEEQOODS
Smith PG
HBX€2C00¢€
HBX670042
Stanstury EE
FCCT7190010
Steffy RC
MDA E9CO001
MDA690002
- MDA71C0OC3
stiegler JcC
GAXE7C033
Stockdale WG
LKX€2C003
Stoddart WC
HCX710022
Strehlow RA
EDX680031
Strokecv VI
EBX70004 2

Swanscn E

ADXE9CC63

-Swindeman ERW

S PBCETCCO1
Taboada A |

- FBXE4CcCc1¢

GFX660023

‘Tallackscn JR

AAX670008
IBA71C00E
Taube M |
IAF€7CC47
Thoma RE
CAX€EEC032
CAX690053
Thomas ¥WE

 

Personal Author Index

JAB710008

LDA700015
LLA700046

IAC7C00u47

HBX7C0012

MDB710002
MDC700004

MDC7C0005

" GAX680028

ECX690051

GGX670034

MAC680034
MDA640007

IAF67C048

CAX69006 1

White JI

BBX€7C012
Thompson CT

LCA€SC037
Tolson GF

GFX6€C023

Uhl VW
NXX€Z00C1
Ulrich WC
MDA7CCOCSE
Valachovic ES
EBX6¢€C03S
Van Winkle &
IAA650024
Venard J1
FEE660021
FBBE€€C022
Vvoznick HE
NXXێ:C0C1
Webster EH
- MEX700002
Weir JR
AAX670006
- ACE€IC02E
ACE680017
ACEE€EC024
ACEESCO2¢
ACE690032
ACE7C002¢
ACE700039
FBE6ECO1E
FBE6S00 16
Welfare FG
REX6700 12
Kerner WJ
ACE67C02z1
Westsik JH
IAAE€C03C
Whatley VFE
AAXG6T7COCY
LIX€€ECO32

EXX7C004€E

'Eilson v

HBX€CSCCEC
kinand F
EBX7CC043
Wood BJ
MDA€SCOC2
wright B¢ '
BAX7C00CE
Yahr GT :
EBX&6<C03S
Zaitsev GG
EBX7C0042

Page 2])

GGX€70034

FECE50017

FEE€60019
FBE€70029

"FEEET0030

FBE€80025S
FCE€90043
FCX€90033
FCX700026
GAX€70033
GAX€EB0028

IACEe60024

LJX€70032
LXX700029

EBX700041

 
 

R

 

 

 

b

 

 
-

(h -

241,
242,
243,
244,
245,
246,
247,
248,
249,
250.
251.
252.
253.
254,
255.

- 256~258.
259-260.

 266-267.

268.

269.

270.

 

213
- ORNL-TM-3595
Internal Distribution

W. B, Cottrell

H. C, McCurdy

A, J. Miller

M. R. Sheldon

D. A. Sundberg

G. ‘D. Whitman

Central Research Library

Y-12 Document Reference Section
Laboratory Records Department
Laboratory Records Department (RC)

External Distribution

D. F. Cope, Atomic Energy Commission, RDT, Oak Ridge, TN 37830
Merson Booth, USAEC, DRDT, Washington D. C. 20545

A. R. DeGrazia, USAEC, DRDT, Washington D. C. 20545

David Elias, USAEC, DRDT, Washington D. C. 20545

Ronald Feit, USAEC, DRDT, Washington D. C. 20545

J. E. Fox, USAEC, DRDT, Washington D. C. 20545

Norton Haberman, USAEC, DRDT, Washington D. C. 20545

P. A. Halpine, USAEC, DRDT, Washington D. C. 20545

W. H. Hannum, USAEC, DRDT, Washington D. C. 20545

R. Jones, USAEC, DRDT, Washington D. C. 20545

J. Neff, USAEC, DRDT, Washington D. C. 20545

T. G. Schlieter, USAEC, DRDT, Washington D. C. 20545

R. M. Scroggins, USAEC, DRDT, Washington D, C. 20545

F. N. Watson, USAEC, DRDT, Washington D. C. 20545

M. J. Whitman, USAEC, DRDT, Washington D. C. 20545

Director of Division of Reactor Licensing, Washington D. C. 20545
Director of Division of Reactor Standards, Washington D. C. 20545

'Executive Secretary, Advisory Committee on Reactor Safeguards,

Washington D, C. 20545
Division of Technical Information Extension (DTIE)

Laboratory and University Division, ORO

R. C. Steffy, Jr., TVA 303 Power Building, Chattanooga, Tenn.
37401
A. Houtzeel, TNO, 176 Second Ave., Waltham, Mass. 02154