ORNL-2684.txt

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ORNL-2684
Reactors-Power

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MOLTEN-SALT REACTOR PROJECT
" QUARTERLY PROGRESS REPORT
FOR PERIOD ENDING JANUARY 31, 1959

CENTRAL RESEARCH LIBRARY
DOCUMENT COLLECTION

.} LIBRARY LOAN COPY
' DO NOT TRANSFER TO ANOTHER PERSON
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and the library will arrange a loan.

OAK RIDGE NATIONAL LABORATORY
operated by i

UNION CARBIDE CORPORATION
for the

U.S. ATOMIC ENERGY COMMISSION

ORNL-2684

Reactors — Power
TID-4500 (14th ed.)

Contract No. W-T405-eng-26

MOLTEN-SALT REACTOR PROJECT
QUARTERLY PROGRESS REPORT

For Period Ending January 31, 1959

H. G. MacPherson, Project (oordinator

DATE ISSUED

MAR 4 1059

OAX RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
U.5. ATOMIC ENERGY COMMISSION : Lignames

I

3 Y456 03b13kY 3

- 1iii -

MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT

SUMMARY

PART 1, REACTOR DESIGN STUDIES

l.1. CONCEPTUAL DESIGN STUDIES AND NUCLEAR CALCULATIONS

A design study is being made of a 30-Mw, one-region, experimental,
molten~salt reactor. A power plant arrangement has been developed that
is based on fabricability and maintenance considerations. Designs have
been prepared of a fuel sampling and enriching mechanism, system pre-
heating devices, and off-gas system valves,

A preliminary design and costeestimate study has been completed on
a 315-Mw (e) graphite-moderated moltenesalt power reactor. In this
reactor the molten-salt fuel flows through holes in the cylindrical,
unclad, graphite moderator, which is contaified in an INOR-8 vesgsel.

The low-enrichment (1.30% U-3°, initially) fuel mixture is LiF~BeF,, -
UFM (70+10-20 mole %).

Nuclear calculations relative to the experimental reactor have been
completed in which the reactivity effects of various modifications were
evaluated. It was estimated that the experimental reactor would have an
over-all temperature coefficient of reactivity of -6.75 x 10"5
(6x/k) /°F.,

The effect of neutron and gamma-ray activity in the secondary
heat exchanger of the interim design reactor of using the salt mixture
LiF -BeF,, (63-37 mole %) as the coolant, in place of sodium, has been
studied. It was found that the gamma-ray dose would be only one-half
that expected from sodium, and that no appreciable neutron activity

would be present.
= v =

The nuclear effects of processing molten-gsalt fuels to remove the
rare-earth fission products by ion exchange with CeF3 have been investi-
gated. A comparison with performance of the system utilizing the
fluoride-voletility processing method showed that the ion=exchange
method gives 50% lower average critical inventory and a 30% lower
cumulative net burnup.

Modifications have been made to the Oracle programs Cornpone and
Sorghum to permit evaluations of heterogeneous reactors;, and Oracle
program GHIMSR has been written to expedite heating caiculations. Follow-
ing test calculations with the modified programs, initial and long-=term
nuclear performance calculations were made for two - low-enrichment, one-
region, graphite-moderated, unreflected, molten-salt reactors. The
results indicate the effect of adding thorium to the fuel mixture. The
fuel used in one calculation was LiF=BeF2=-UFh (70-10-20 mole %), as
described above, and in the other the fuel wasg LiF=-BeF2-=‘I‘hFh (67-16-13
mole %) plus sufficient U235Fh=-U238.Fh to make the system critical. For
a cumulative power generation of T200 Mwy, the critical inventory with-
out thorium was 4386 kg and, with thorium, was 1416 kg. The cumulative
net burnup without thofium was 1400 kg, and, with thorium, was 1083 kg.
Calculations of the performance of a thorium-containing graphite-
moderated and =reflected system were initiated.

1.2, COMPONENT DEVELOPMENT AND TESTING

Investigations have been continued of salt-lubricated bearings
for molten-salt pumps. Bearing and journal parts fabricated of INOR-8
are being subjected to numerous tests to determine optimum design and
operating conditions. Equipment for testing salt-lubricated thrust
bearings was designed and construction is nearly complete. A PK type
of centrifugal pump is being prepared for service tests of molten-sslt
lubricated bearings. The favorable results being obtained with thesge

-V -

bearings have resulted in the postponement of tests of the more complicated
hydrostatic type of bearing.

The modified Fulton-Sylphon bellows-mounted seal being subjected to
an endurance test in a PK-P type of centrifugal pump has continued to
seal satisfactorily for more than 10,000 hr of operation. Operation of
an MF type of centrifugal pump with fuel 30 as the circulated fluid has
continued to be satisfactory through more than 13,500 hr, with the past
11,500 hr of operation being under cavitation damage conditions.

A small frozen-lead pump seal on a 3/16-in.-dia shaft has operated
since the first 100 hr of almost 5000 hr of operation with no lead
leakage. Test operation of a similar seal on a 3 1/4-in.-dia shaft has
been initiated.

Techniques for remote meintenance of the reactor system have been
investigated further. Specifically, improvements were made in the
freeze-flange Jjoints being developed for remote connection of molten-
salt and sodium lines. Construction work continued on the remote-mainte-
nance demonstration facility.

Equipment for salt-to-salt heat transfer tests has been completed.
Data are being obtained with which to determine heat transfer coefficients.

A heat exchanger has been designed with which to evaluate triplex
tubing for use in moiten-salt or liquid metal-to-steam superheater appli-
cations. A small heat exchanger test stand will be modified for testing
of the heat exchanger.

Further tests of commercial expansion joints in molten-salt lines
have demonstrated conclusively that such joints are fiot suitable. Typical
failures -indicated, however, a high-stress point that could be eliminated
by redesign.

Operation of forced-circulation corrosion-testing loops has been
continued, and most of the 15 loops have been modified to prevent freeze-

ups of the salts being circulated.
- V] =

Operation of the first in-pile loop was started at the MIR on
December 3, 1958 and had continued satisfactorily through 860 hr, when
it was shut down because of the leakage of fission gases. The radiation
release has been traced to a partially plugged pump-sump purge=-outlet
line whicfi caused flssion gases to back-diffuse up the pump-sump purge-
inlet line to a point outside the cubicle, where they escaped through
a leak.

The assembly of an improved in-pile loop is approximately 60%

complete.

1l.3. ENGINEERING RESEARCH

Modifications in the design of the capillary efflux viscometers
and in the method of calibration have led to more precise data on the
viscosities of the salt mixtures LiF-BeF,-UF, (62-37-1 mole %) and
LiF-BeFE-U'Fh-ThFh (62-36.5-0.5-1 mole %). At the lower temperatures
the viscosities are significantly lower than those previously reported.
The solid-phase enthalpies and heats of fusion have been obtained for
three LiCl-KCl mixtures.

Fabrication, assembly, and initial calibrations are continuing on
components for the forced-circulation loop designed to ascertain the
presence and effect of interfacial thermal resistances in systemsg cir-
culating BeFe-containing salts. Heat-transfer coefficient measurements

will be made as a means of studying surface film formation.
PART 2. MATERIALS STUDIES

2.1. METALLURGY

Examinations of three Inconel thermsl-convection loops that had

operated for long periods of time and one Inconel loop that had operated
- Vvii =

for 1000 hr have been completed. Anomalous results were obtained that
are largely attributable to impurities in the circulated salt mixtures.
No forced-circulation loop tests were completed during the quarter.

Mechanical property tests have been made on Inconel and INOR-8 speci=-
mens that were carburized by exposure for LOO0 hr at 1200°F in g system
containing sodium and graphite. Carburization was found to decrease the
tensile strength and elongation of INORe8. On the other hand, carburi-
zation increased the tensile and yield strengths of Inconel and decreased
the elongation. The results of chemical analyses of the specimens were used
to establish relationships of carbon content vs distance from specimen
surface for various time intervals.

Similar specimens exposed for 400 hr to sodium containing graphite
showed no evidence of carburization. Specimens of Inconel and INOR-8
were also exposed at l300°F to fuel 130 containing graphite for 2000
and 4000 hr. The Inconel specimens were attacked and showed reductions
in mechanical strength. The INOR-8 specimens showed no carburization
and only slight attack after 4000 hr.

Further tests have been made in the study of the compatibility of
graphite and molten-salt mixtures. Uranium oxide precipitation has
occurred in all tests with fuel 130. The quantity of uranium precipitated
appears to be an indirect function of the purity of the graphite. Attempts
are being made to eliminate the precipitation by using purer graphite.
Tests are also under way to investigate the penetration of graphite by
fuel mixtures as a function of time, temperature, and pressure.

The strength properties of INOR-8 at high temperatures are being
investigated in tensile, creep, relaxation, and fatigue tests. Prelimi-
nary data indicate that the creep properties in molten salts at 1200°F
are the same as those in air. Tests of notch sensitivity have indicated
that INOR-8 is notch strengthened at room-temperature and at l500°F.
Rotating-beam fatigue studles are being conducted under subcontract at

Battelle Memorial Institute.
- viii -

Fabrication and welding and brazing studies are being made of triplex-
tube heat exchangers. The tubing consists of two concentric tubes with
a porous-metal-filled annulus. Means are being sought for obtaining good
conductivity across the annulus and adequate permeability of helium through
the porous metal for leak detection. Prefabricated porous materials are
being obtained from commercial sources for evaluation.

Studies of composites of INOR-8 and type 316 stainless steel have
shown the alloys to be compatible at temperatures up to 1800°F. Such
composites therefore appear to be promising for use in fused salt-to-
sodium heat exchangers.

Various means for deoxidizing and purifying ingots of INOR-8 weld
metal have been studied as means for improving the ductility of INOR=8
weldments. The ductility at lSOOOF has been increased from an average
of 7% to 13 to 15%, and further improvements are thought to be possible.
No increase in high-temperature ductility was obtained by decreasing
the carbon content.

A niobium-containing coated electrode, designated Inco Weld "A"
electrode, was investigated for use in joining dissimilar metals. The
studies showed that where the metallic-arc welding process could be
used, such weld deposits would be satisfactory for joining dissimilar
metals for high-temperature service.

Molybdenum~to-Inconel joints were brazed with several materials
in connection with the fabrication of a pump-shaft extension. All joints
showed a tendency to crack because of stresses built up by the different

thermal expansion properties of the base materials.

2.2. CHEMISTRY AND RADIATION DAMAGE

Phase studies are being conducted to determine whether the NaF-
BeFQ-UFh system has any advantages over the LiF-Ber-ThFh-UFh system.

The need for liquidus temperatures of 55000 or lower in nuclear reactor
o X =

systems will restrict the ThFh concentration to the range 10 to 15 mole %.
The substitution of UFh for part of the ThFh would be expected to lower
the liquidus temperature.

Plutonium trifluoride has been found to be sfifficiently soluble in
LiF--BeF2 and NaF-BeF2 melts to form a fuel mixture for a high-temperature
plutonium-burning reactor. An apparatus was developed in which l=-g
samples can be used for phase-relationship studies.

The system KF-LiF-BeF2
coolant. Such mixtures would have lower gamma activity than similar
NaeF-base mixtures after irradiation. The mixture LiCl-RbCl is also
being examined as a possible coolant. Experimental studies of the com-
patibility of the chloride mixture with fuel mixtures have shown that

is being investigated as a possible reactor

no uranium compound would deposit as the primary phase at reactor temper-
atures.

In studies of figsion-product behavior, it was found that additions
of UF, had no effect on the solubility of CeF3 in LiF-BeF, (62-38
mole %). In esnother experiment it was demonstrated that the addition
of CeF, to remove SmF_ from LiF-BeFE-UFh mixtures was effective even

3 3

when‘the SmF3 was present in trace amounts. The behavior of oxides in
molten fluorides is being studied as part of an effort to explore
chemical reactions which can be adapted to £he reprocessing of molten-
fluoride-galt reactor fuels.

In the study of the chemistry of the corrosion process, the
activity coefficients of the fluorides of structural metals in dilute
gsolutions of molten fluorides have been measured. The data are useful
in understanding and predicting the corrosion reactions which take
place in systems in which molten fluorides are in contact with alloys
of these metals.

Techniques have been developed for determining the self-diffusion
coefficients of chromium in chromium-nickel alloys with Cr51 as a

radiotracer. The results of recent experiments have shown that the
o X =

grain structure has a marked influence on the gelf-diffusion rates of
chromium in Inconel. Grain size appears to be a controlling factor.
Hydrogen firing and annealing in helium had the same effects on the
over-all diffusion rate.

Salt samples taken from operating forced-circulation corrosion=
testing loops with decreasing frequency were analyzed. Samples taken
from an INOR-8 loop during the first 1000 hr showed a slow but steady
increase of chromium from about 420 to 530 ppm. Samples from an Inconel
loop showed an increase in chromium content from about 350 to 450 in
500 hr, The chromium content of neither salt has reached a steady-
state value,

Vapor pressurescfi'CsF-Bng are being measured to obtain values
on which to base estimates of the vapor pressures of LiF-BeF2 mixtures.,
Activities in the LiF-BeF2 system cannot be obtained from vapor-pressure
measurements because of complex association in the vapor phase. The
vapor pressure of liquid UFh wag measured between 4 and 180 mm Hg
(1030 to 1300°C).

Further studies have been made of gaseous aluminum chloride as
a heat transfer medium. Calculations indicate that very low pumping
power would be required and that it would be useful as a turbine
working fluid. Experimental investigations of the compatibility of
aluminum chloride and structural metals have shown very little attack.

Reactor-grade graphite impregnated with LiF=-MgF2 has been tested
in LiF-BeF,-UF, at 1250°F for 800 hr and examined for uranium pene-
tration. Complete penetration of beryllium ard uranium was found in
decreasing amounts with distance from the surface.

A series of tests was run to determine the effect of thermal cycling
on salt stability. It was found to be important in the handling of
beryllium~based fuels to ensure complete melting of a batch before any

portion is transferred to another container or test rig. Thermal

cycling under static conditions must be avoided.

- X1 -

The in-pile thermal convection loop that was being readied for
insertion into the LITR was found to have a deposit at the bottom of
the lower bend. Examination of the fuel from which it was filled
revealed the presence of UO2 on solid pieces. A second loop is being
assembled.

Methods have been developed for synthesizing simple fluorides by
reactions with stannous fluoride. The following fluorides have been

MoF,, ALF,, FeFp, VF, and UF,.

prepared: CrFE, 7 ) 3

2.3. FUEL PROCESSING

The solubilities of fluorides of neptunium, corrosion products,
uranium, and thorium were investigated to evaluate their behavior in
the proposed EF dissolution process for recovering LiF-BeF2 salt. On

the basis of this work, the recovered LiF-BeF, salt after processing

2
by the HF dissolution method would contain, at most, 0.001 mole %

neptunium. The solubility data indicate that the fluorides of neptunium

and possibly plutonium behave similarly to the rare earth fluorides.
Separation of the recovered LiF--BeF2 from corrosion~product fluorides
would probably be less effective. However, the solubility of any
particular contaminant fluoride is usually less in the presence of

others.
- Xjiidi -

CONTENTS

SUWARY P B0 0O C P OB EOCE S BN AEADSSOPERVRIOCO0OCG0CSONCAIBS S LONGCOO0D0CHEOSASD000E0

PART 1. REACTOR DESIGN STUDIES

1.1. CONCEPTUAL DESIGN STUDIES AND NUCLEAR CALCULATIONS ssecaecescoces
Design of a 30-Mw, One Region, Experimental Molten-Salt |

ReacCtor seseeeecsescecnoesccncsococoescssosencooonooosoesceccoas
Reactor System ceeececcccccscsscccossncanscsecoecsccososcoose
Power Plant Arrangement coccoccococcccseecsssccscoococcoscsosssaoc
Fuel Sampling and Enriching Mechanism ccccccccccaccoscecean
System Preheating Device c.cccocecoccecoccocsccoecnoccsscascsn
Off-Gas System Valves .seecsccsssoesosssscescssccosscscssocs

Preliminary Design of a 315-Mw (e) Graphite-Moderated Molten-
Salt PO’WGI‘ Reagtc‘?:o--o-tooooooooouoaoooooooooooooooooooooo.io

Nuclear Calculations .esseescoccoscscscesacecescsecococoosceenssse
Reactivity Effects in Experimental Molten-Salt Reactor c...
Neutron and Gamma Ray Activity in the Secondary Heat

- Exchanger of the Interim Design Reactor cccsecaecscsceoos
Effect of Ion-Exchange Processing of Rare-Earth Fission
Products on Performance of Interim Design Reactor cccesos
Multigroup Oracle Programs for Heterogeneous Reactor -
COmpu‘bationS P00 800000 GEOPE0EER00CO08 000000000 S0D0000C0O00D
Oracle Program GHIMSR for Gamma-Heating Calculations ccceoe
Initial and Long-Term Nuclear Performance Without Fuel
Processing of Low-Enrichment, One~Region, Graphite-
Moderated, Unreflected, Molten-Salt Reactor cececccscccss
Initial and Long-Term Nuclear Performance Without Fuel
Processing of One=-Region, Graphite-Moderated,
Unreflected, Th232-Conversion, Molten-Salt Reactor seece.
Initial Nuclear Performesnce of One-ge on, Graphite-
Moderated, Graphite~Reflected, Th23<-Conversion, Molten-
Salt Reactor .eescococcescessscsccoecosaosoecscocoososoanos

102. COMPONENT DEVELOPBEN.II.I AND lI’ESTING .UOUGOOOOO000000{)‘000.000000000

Salt-Lubricated Bearings for Fuel PumpsS coccccococcocceccsccccecocs
Hydrodynamic Journal Bearings coceccceccosccocccsoccasscasncs
Hydrodynamic Thrust Bearings ccccccoccococscscoscccosccescoccooo
Test Pump Equipped with a Salt-Lubricated Journal

Bearing sccococeescesccccococecesscccooonsooen0000006000000000
Hydrostatic Bearings ccoceesesscesccocoeccccccssccoesooccnnos
Bearing Mountings ccccocccoscoccceeessccococscocneascosccocncococaa

iii

(WS

30

31
1.3.

- Xiv -

Conventional Bearings for Fuel Pumps ...... cecsecoerssosscesasa
Mechanical Seals for Fuel PUmps ....eeceoess ceesssesansracnanas
Labyrinth and Split-Purge Arrangement .......oveeees. ceasos
Bellows-Mounted Seal ..ceceevn. cesossesseseassrannnns ceces
Pump Endurance TeSting .voevoererseereoonccasacsesoosaconcsccasse ,e
Frozen-Lead Pump S@8L . .icciooecooaesneeeasoosoonesasnacsonensss
Techniques for Remote Maintenance of the Reactor System .......
Mechanical .Joint Development ..... s ereeccsanssecannns ceoen
Remote Maintenance Demonstration Facility ...... cesecannns
Molten-Salt Heat-Transfer-Coefficient Measurement .....eeeese..
Triplex-Tubing Heat Exchanger Development ......... sevessencacs
Evaluation of Expansion Joints for Molten-Salt Reactor
Systems .eiviveennnvace cscscessessarse s cesacasretssessesnoa
Design, Construction, and Operation of Materials Testing
Loops ceveunn cocetrmrers seseceans i reesasesansensee teocesnuae
Forced-Circulation LOOPS tevevieceeeecnnsas ceenssossnsenesn
In-Pile LOOPS sevsneerssteroonnnnannsnnnnasanans et enscans
ENGINEERING RESEARCH v evennootoronnonensnsooecssssnnsonnessses
Physical Property Measurements ........... Sesesenssserasasassns
ViSCOSItY cvvieninniiieiineeetoonoeennssonnnsesonaaneesnes
Enthalpy and Heat Capacity ceceeeeeenennens s easssasannsas

Molten~Salt Heat Transfer StUdi€sS ceececcococsocscscoecooooososseos

PART 2. MATERIALS STUDIES

METALL[JRGY .....0“0&.00.00.00000..00000000000.0900.000GQQOOO.-O

Dynamic Corrosion Studi€s cccoccccossesococccoecssstoscoscsscsa
Thermal-Convection Loop TeStS cocoococcscosossosoesscccoosss
Forced=-Circulation Loop TeS8t8 scccsesccescocososcescoooocaoo

General Corrosion Studies sesececococoosocecnoooscnncoonoeossssas
Carburization of Inconel and INOR-8 in Systems Containing

Sodium and Graphite cceseecoccceccocceccosesososoeccosescsss
Carburization of Inconel and INOR-8 in Systems Containing

Fuel 130 and Graphite ccceoccoccscosscocecocccccoccosocsnso
Uranium Precipitation From Fused Fluoride Salts in

Contact with Graphite cccecececaccccaoscosccsoccoesncoao
Fused Fluoride Salt Penetration Into Graphite coeoocccoccecoe

Mechanical Properties of INOR=8 cccoococoocescassccooosscoocossos

Materials Fabrication Studies cecococcccsccccocoscococessessccsse
Triplex Tubing for Heat EXChangers scecceocccoccocoeececcsos
High-Temperature Stability of INOR=8 .ceo0co0cco0cscecoccocoss
Commercial Production of INOR=8 coocccococcocoscnsecosnoass
Evaluations of Composites of INOR-8 and Type 316

Stainless Steel cccoooccvecooecescnscsasassnonsonnoossoo
2.24;

2.3.

Welding and Brazing Studies cocoocosocococccccsescccscocescossesssa
Heat Exchanger Fabrication coccococcocococococceccosccocococossocoses
Welding Of INOR=8 ceeseecscscoccconocccoscossooscoononnsso
Welding of Dissimilar Metals eecocococcococococococovesssococoscssco
Pump Component Fabrication cecoceoccococcosococoecescocoscsoanscnos

CHEN[ISTR.Y AND RADIATION DAMA-GE 0900 00S#000DO0O00CO0LO00D0OB00000OCQ0D0CO0OESE

Phase Equilibrium Studies cecocesscceconcccccscosssocssccscossss
Systems Containing UF) and/or ThF) sccccosssccossscsosssos
The System LlF"PUF * 8006 CE0GE0D008000000008000000R0EDCOESD
The System KF'-LiF- F2 ©08000008€000000000000000060000000DCOCO
Compatibility of Fuel with Chloride Coolants ceoceococccecses
Fisgion=Product Behavior «ocsoceocoeccssesscsscooecococscsoscnoccses
Effect of UF) on Solubility of CeFy in LiF-BeF Solvents .
Removal of Traces of SmF3 by the Agdition of CeF3 to
LlF-BeF “UF 200009 SO0 0OSAO00DO0RSSICOVGO0BO00SEOARSEUOOOOD
Chemical Reactions of Oxides with Fluorides cccccscocooccses
Chemistry of the CorrOSion Process occecesecocooocscosesccecscons
Activity Coefficients of CrF, in NaF—Zth cacsccoosesensoo

Chromjum Diffusion in A110YS scsasvcscn seassoenas coasns
Sampling of Operating LOOPS ceccscsnceossccsssecssessscsas -
Effect of Fuel Composition on Corrosion Equilibria .......
Vapor Pressures of Molten Salts .ccoecoe soosocos ceosveessnansan

BeFp MiXtures .scceccccssocessccsoesessoccscosesoasoesonns
UEF'), soooscasooccsocsescosscoocnsoooasscnosonooescooonoooossosss
Aluminum Chloride Vapor as a Heat Transfer Medium and Turbine
Working Fluid :scccocscccesccscscassacnosasoscosocensnscossccsao
Estimation of Thermodynamic Properties cccceccceccccsccses
Corrosion of Nickel by AlCl. ccocscocecocoscacasecesosnons
Permeability of Graphite by Moltef Fluoride Mixtures .cceccecsoes
Effects of Thermal Cycling on Salt Stability ccsceocecccoeccacaance
Radiation Damage Studies .ccocccesoesccoesaccsescascocessssssaesse
INOR-8 Thermal-Convection Loop for Operation in the LITR ..
In-Pile Static Corrosion TesSt .ecccsescsseccsssscsccscssesse
Preparation of Purified Materials ccoesccoesessaocssccconcsssessas
Fluorides of Chromium coesecsesescecscsscosoesascoscscnnescace
Synthesis of Simple Fluorides by Reactions with Stannous
Fluoride cccocosescecccesnsessscncsscosocesccsccocsossssos
Experimental-Scale Purification Operations .cocseccosccososo
Transfer and Service Operations cceeeeccococscoccocosccncecse

FIEL PROCESSING 00..00.OQDOOQOAUQUOOOOOICOCO.0...0.0000.....9.0.

Solubility of Neptunium in Aqueous HF Solutions sesecccsscsccss-

Solubility of Corrosion-Product Fluorides in HF Solvent ccceces

Solubilities of Uranium and Thorium Tetrafluoride in HF

SOlvent 0 000 & 6 & 0000 0SS0 S S S 000 ES 8 E S SO0 DS SO 000 S8S S O0¢E BO0S SO0 OC & 6D A
Part 1.

Reactor Design Studies
- 3 =

1l.1. CONCEPTUAL DESIGN STUDIES AND
NUCLEAR CALCULATIONS

DESIGN OF A 30-Mw, ONE-REGION, EXPERIMENTAL, MOLTEN-SALT REACTOR

Conceptual designs of an experimental moltem-salt reactor and power
plant arrangement have been prepared. In these designs, consideration
has been given to fabricability and to the elimimation of awkward design
situations. An attempt has been made to take into account all features
of a finished power plant. Improvements in plant layouts and equipment
design will, of course, result from further study.

Reactor System. The proposed reactor consists of a 6-ft-dia

spherical core, a heat exchanger above the core, and a sump-type fuel
pump offset from the main vertical axis and located slightly above the
heat exchanger, as shown in Fig. 1l.1.1l. Liquid fuel leaving the pump
discharge is directed to the top of the heat exchanger, where it passes
downward. Flow through the heat exchanger is parallel and countercurrent,
with the fuel outside helical tubes containing an inert molten-salt heat
transfer fluid. The fuel leaves the heat exchanger through a pipe that
communicates with the top of the spherical core, passes downward through
the center of the sphere, and reverses its direction of flow at the
bottom. It leaves the core through an annulus at the top. This annulus
feeds the fuel directly to the sump bowl of the fuel pump.

A small portion (about 10%) of the total fuel flow is bypassed
through a gas-stripping system. Directly above the heat exchanger, at the
end of the pump-discharge diffuser, there is a plate perforated with small
nozzles which spray the bypassed fuel above the plate and cause i1t to be
intermixed with helium flowing to an off-gas system. This helium
originates as a purge gas in the seal region of the fuel pump. The
reactor fuel system may be filled with or drained of fuel through a dip

line that communicates with two valves in series and a storage tank
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A s a ' a’
. .'. 'A 7
s 2 . R
o « . -
Fs . N :A .
- - b
LAMINATED ot
.o .
SHlELDiNG—\.' ' .
. 4 : o & T ) ‘
< o . ) €, 4 re
. a o
a a a . B :-
' . a
sl Fa)
o " 4 .
* - a - .
'y A 3 s
° - aa .
- ‘, a‘. ‘\‘: a
4 ' AN 4 )'
a : = =
o \ ——— - -
ry “ a A; .
& a a "'
i
a4 A '
o P - *
= N - -
. -

Fig. 1.1.1. Conceptual Design of a One-Region, Experimental, Molten-Salt Reactor.

(not shown). A fuel enriching and sampling system, which is described in
a following section, is appended to the pump bowl.

There is an insulated gas passage surrounding the core vessel
which constitutes one leg of a heating or cooling loop. This gas loop
is provided to maintain the fuel above its melting point and below a fixed
maximum temperature during all off-design conditions. The remainder of
the gas loop consists of a blower, a heater, a cooler, and two dampers,
which operate to direct the flow through either the heater or the cooler.
The heater is described in a following section.

The reactor system is surrounded by a double=walled steel cell. The
space between the two walls is monitored for leaks. Space coolers within
the cell keep the cell walls below lSOOF.

The design of the reactor system provides for semiremote maintenance.
Those components which have the highest probability of requiring
maintenance or replacement are located so that replacement involves a
minimum of remote operations.

Power Plant Arrangement. Preliminary layouts of the power plant are
presented in Figs. 1l.1.2 and 1.1.3. It may be noted that the reactor is

at a higher elevation than the main condenser in the steam plant. This

arrangement avoids any possibility of flood water entering the reactor
cell. Obviously this design feature will raise the building cost, and
it may be moreheconomical to lower the reactor in relation to the condenser
and at the same time raise the whole plant to above the level: of the
cooling-water source. In this case the price to be paid would be extra
pumping costs. A detailed study of a particular site would be required
to establish the proper relations.

In contrast to most steam power plant designs, the main crane in
the turbine-generator room also provides services to many of the other
ma jor components of the plant, such as the evaporators, feedwater heaters,
and deaerator. A schematic flow diagram of the power plant is presented

in Fig. 1.1l.k.
UNCLASSIFIED
ORNL-LR-DWG 35664

14

- COOLANT
PUMP

REMOVABLE SHIELDING —.. . SUPERHEATER
|

H FUEL PUMP | }7“——':" T h ‘l Tflgfl‘)
. by

BLOWER ~

o |
|

%z»OFF*GAS
) ;LT HOLDUP TANKS

REAGTOR GAS HEATER —
{GOOLER BEHIND)

~ TURBINE-GENERATOR

L T ]

FEEDWATER HEATERS +===—

Fig. 1.1.2. Elevation Drawing of Power Plant for 30-Mw, Experimental, Molten-Salt Reactor.
FUEL DRAIN TANKS

(UNDER) AN

M

N
;

| -
|
ACCESS TO REACTOR FUEL
GAS HEATER, COOLER
AND BLOWER —— ——|_

|
SAMPLER, ENRICHER -

S CELL COOLER

- DRAIN TANK COOLER

] B I
ofey
HEATER. I

{UNDER)

DRAIN VALVES *

(UNDER)

-REACTOR
CENTER

(UNDER)

CONTROL ROOM

SUPERHEATER -

A

20

% Maccess To

ACCESS TO EQUlPMETI///
REMOVAL TUNNEL

\l

“~CHARGOAL TRAPS
{UNDER)

COOLANT; \»J
PUMP - |

I/ EVAPORATORS

S

-

d :i
DEAERATOR ™ ;)f
il .

UNCLASSIFIED
CRNL-LR-DWG 35665

L«— TURBINE-GENERATOR

I

Fig. 1.1.3. Plon Drawing of Power Plant for 30-Mw, Experimental, Molten-Salt Reactor.

FUEL PUMP

I

1210°F

REACTOR

96,3001b
1000°F 1450 psi

e

COOLANT PUMP

125°F

L

|- PRIMARY
HEAT EXCHANGER

/ "\_Fl
/
ATTEMPERATOR

UNCLASSIFIED
ORNL-LR-DWG 35666

- TUBINE-GENERATOR —D

SUPERHEATER —#
(COOLANT-STEAM)

OO
_F r -

4

1075°F
3.33cfs
1,460,000 Ib/hr

349,0001b
610°F 1470 psi

STEAM PUMP

593°F

228,700 1b

EVAPORATOR

97,260 Ib 450°F/ /
Y

FEEDWATER HEATERS
/

1% BLOWDOWN

260 1b FEEDWATER PUMP/

\\
j
- y
Y
[ CONDENSER
}
1% MAKE UP
960 Ib
/ Y ¥
namVAVAVA
]
— 1 -
/ . \CONDENSATE PUMP
FEEDWATER HEATER
DEARERATOR

Fig. 1.1.4, Steam System Flow Diagram for 30-Mw, Experimental, Molten-Salt Reactor.

-8-
- 9 =

Fuel Sampling and Enriching Mechanism. The fuel sampling and

enriching mechanism, shown schematically in Figs. 1l.1.5 and 1.1.6,
consists of a sampling-pot elevator; a horizontal conveyor; a sample-
depositing elevator; an enriching-capsule elevator; four, identical,
motorized worm-gear drives; several enriching-capsule magazines; several
shielded semple carriers; an enricher-magazine press; a sample-carrier
press; a solenoid-operated stop; a canned remotely controlled gate valve;
a number of canned remotely controlled ball valves; a supply of sampling
capsules; and a supply of enriching capsules. All parts of the system
are enclosed in vacuum-tight piping that is insulated and shielded where
required. Every operation is remotely controlled and electrically
interlocked so that accidental improper sequences of operation cannot
damage the equipment.

The sampling-pot elevator carries capsules from the horizontal
conveyor down into the sampling pot of the reacter, either to obtain
samples or to add enriching capsules. It is actuated by a motor-driven
bronze worm-gear. Between the horizontal conveyer and the reactor are
a ball valve and gate valve. On the down stroke of the elevator plunger,
these must be opened in turn as the plunger approaches them. On the up
stroke, they must be closed immediately behind it.

The capsules are propelled along the horizontal conveyor by a transfer
plate suspended from a track. There are breaks in the track where the
three elevators cross it. These breaks are filled by the plungers of the
elevators when they are in the up position. The transfer plate has a
hole in it through which the elevator plungers operate %o obtain or deposit
capsules. There is a solenoid-operated stop provided to center the
transfer plate hole over the sample-depositing elevator. End-~of -travel
stops center the hole over the other two elevators. A bolt-on blind
flange is provided for access at the reactor end of the conveyor, and a

bolt-on flange matching that of the drive unit is welded to the other end.
-10-

UNCLASSIFIED
ORNL-LR-DWG 35667

MCTORIZED WORM-GEAR DRIVES MCTCRIZED WORM-GEAR DRIVES

: SAMPLING-POT ELEVATOR ki ENRIGHING-CAPSULE
/ ! ELEVATOR—_ |

SAMPLE-DEPOSITING ELEVATOR —\

HORIZONTAL CONVEYOR

SAMPLING CAPSULES
ON GONVEYOR

CANNED, REMOTELY CON-
TROLLED BALL VALVES

SHIELDED SAMPLE N
45 CARRIER S

ENRICHER-MAGAZINE
PRESS

ENRIGHING-CAPSULE

X GANNED, REMOTELY CON- |
, MAGAZINE

TROLLED BALL VALVE

SAMPLE-CARRIER PRESS

— GANNED, REMOTELY CONTROLLED
P

GATE VALVE

REACTOR CELL TCP SHIELD SAMPLER ACCESS AREA

Fig. 1.1.5. Fuel Sampling end Enriching Mechanism,

UNCLASSIFIED
ORNL-LR-DWG 35668

\/Kgl {T\_/\/\
Nl ’
?'b I\si - é
7N Nz N7
NS A » N4 \\\
G i
@ Z

\\

//

NN R RN IR Y

(g) (b)

AN

AT IR

/
A\
SRR

Fig. 1.1.6. (a) Sampling-Pot Elevator Showing Plunger Under a Sampling Can. (b) Sample-Depositing Elevator
Showing Tongs.
- 12 -

At the bottom of the sample-depositing elevator plunger is a pair .
of remotely controlled tongs that grip and release capsules as desired.
In their up position, the tTongs bridge the gap in the horizontal
conveyor track. The tongs cannot open until the capsule is within the
shielded sample carrier. At the bottom of the elevator, the shielded
sample carrier rests upon a hydraulic press which forces it against a
seal in the elevator pipe flange to eliminate the necessity of bolting
it. The enriching capsule elevator is similar to the sample-depositing
elevator.

System Preheating Device, Two designs of the heater bank for

heating the gas used to preheat the reactor system are being considered.
Tubular heaters with an Inconel sheath are included in one design and .
flat ceramic heaters with an Inconel enclosure are utilized in the other
design. The main features of the units being studied are listed in -
Table 1.l.l. The chief advantages of the fiat ceramic heaters are: (1) the
vattage per unit surface area is less, (2) any number of heater dimensions
can be obtained, (3) the support for the ceramic plate is provided by
the enclosure, (4) fins can be added to the enclosure to provide
additional heat transfer surface if required, and the (5) vertical air
flow may simplify duct work.
A diagram of a heater unit with two, flat, ceramic heaters is shown
in Fig. 1.1.7. A group of these units would provide a heater bank
with a capacity of 6 to 50 kw in a 2-ft by 2-ft duct.

Off-Gas System Valves. Frozen-bismuth valves are being considered

for use in the off-gas system, because (1) conventional metal-to-metal

seated valves are not consigtently tight in the contaimment of fission- -
product gases, (2) valves employing elastomer closures are of limited
life because radiation destroys the elastomer, and (3) devices such as
bellows to replace stem glands tie the life of the valve to that

of the fatigue limit of the bellows. The use of frozen bismuth in a

trap to create a perfectly tight valve both processwise and environment-

Table 1l.l.l.

- 13 -

Features of Tubular end Flat

Ceramic Heaters

Variable

Tubular Heater

Flat Ceramic

Rating, w/in.2
Rating per heater, w
Voltage, v
Dimensions

Shape
Surface fins

Sheath temperature, °
Mounting

Air flow

Power source

Voltage control

4O
750-3500
230

18- to 8h=in. heated
length; 0.315 in. OD

Hairpin loop

Not available for
this temperature

Up to 1650

Vertical, tie wires
or supports may be
required which will
tend to cause hot
spots |

Horizontal

A group of T.5«kva,
0 to 270-v Power-
stats ganged and
operated by a common
shaft

Manual or automatic by
an air-operated motor

Up to 20
500-3500
230

Thickness, ~ 3/ in.;
width, up to 12=in.;
length, up to 18 in,

Flat plate

Enclosure can be made
with fins

Up to 1650

Vertical support for
ceramic plate is
provided by the
enclosure

Vertical, leads at end
of heater; horizontal,
leads at top of
heater

Same as for tubular
heaters

Same

-14-

UNCLASSIFIED
ORNL- LR—DWG 35669

HEATER LEADS

METAL CONTAINER

HEATERS

Fig. 1.1.7. Heater Unit With Two, Flat, Ceramic Heaters.
- 15 -

wise has been demonstrated.l Valves of this type have been considered

- in the past only for limited operation {once open and once closed), but
two schemes for repeat action valves, shown in Figs. 1.1.8 and 1.1.9,
are now being studied.

The valve shown in Fig. 1.1.8 is gravity and temperature operated.
In the valve open position the bismuth in the lower trap is in the frozen
state., Likewise, in the valve closed position, all the bismuth shown is
in the frozen state. Heating zones 2 and 3 are operated together only
when the valve is to be opened; the bismuth is melted for a sufficient
time to allow complete drainage of the trident-shaped upper trap.

Heating zones 1A or 1B are operated in conjunction with heating zone 2
when it is desired to close the valve. In this operation bismuth flows

) from one of the charge containers to the trident section and is
subsequently frozen in place. There is enough heat capacity in the
lower trap to prevent melting during the closure operation.

New charges of bismuth may be placed in either charge container
whenever the valve is in the closed position. If the valve is to have
a limited application, enough charge containers are built in place to
co@er the number of operations expected, and no attempt is made to add
charges after the original installation.

A gas-pressure and temperature-operated valve is shown in Fig. 1l.1l.9.
Although it is simpler in construction than the gravity-operated device,
it may-be argued that it is not foolproof. Strict control of the maximum
operating gas pressure while the sealant is in the molten state would be

highly necessary in this device.

lW. H. Kelley, Fuel Sampling Bismuth Valve Test, ORNL CF=57=7=36
(July 24, 1957).

-16-

UNCLASSIFIED
CRNL—LR—DWG 35670

HEATING ZONES 1A AND 18
SEALANT CHARGE CONTAINERS

HEATING ZONE 2

GAS IN

GAS OUT

VENT

c =

o
w
=<
o
i
(&)
=
T
<l
w
T

N

.

.

8
.
§ .
§\\\

FROZEN BISMUTH

FROZEN BISMUTH

7

-

La—— DUMP TANK

L DUMP TANK

VALVE CLOSED

VALVE OPEN

Sy stem,

Fig. 1.1.8. Gravity and Temperature Operated Frozen-Bismuth Valve for Off-Gas

-17-

SSSSSSSSSSSS
LLLLLLLLLLLLLLLL

SSSSSSS
> GA

SSSSSSS
GGGGGG

e
A
; S 5 b

_
=

AAAAAAAAAAAA

- 18 =

PRELIMINARY DESIGN OF A 315-Mw (e) GRAPHITE-MODERATED MOLTEN=-SALT POWER
REACTOR '

A preliminary studyg’3 was made of a one-region molten-salt power
reactor in which the fuel is in contact with the unclad graphite moderator.
The molten-salt fuel flows through holes in the  graphite, which is
contained in a cylindrical INOR-8 vessel. The graphite core is 12.25 ft
in diameter and 12.25 ft high, with 3,6-in.-dia holes on 8=in. centers.
The fuel, LiF=-BeF,-UF) (70-10=20 mole %), contains low-enrichment (1.30%)
uranium, and it occupies 16% of the core volume. A diagram of the reactor
system is shown in Fig. 1.1.10.

The choice of the power level for this design study was arbitrary,
since the core is capable of operation at 1500 Mw (th) without: exceeding
safe power densities. An electrical generator of 333-Mw (e) capacity
was chosen, which would give a station output of 315 Mw (e). The reactor
power rating would thus be 760 Mw (th).

A plan view of the reactor plant layout is presented in Fig. 1.l.12;
and an elevation view is shown in Fig. 1.1.12. The reactor and the
primary heat exchangers are contained in a large rectangular reactor cell,
which is sealed to provide double containment for any leakage of fission
gases. The rectangular configuration of the plant permits the grouping
of similar equipment with a minimum of floor space and piping. The

plant includes, in addition to the reactor and heat exchanger systems

2H° G. MacPherscn et al., A Preliminary Study of a Graphite Mcderated
Mplten Salt Power Reactor, ORNL CF=59-1-26 {(Jan. 13, 1959).

3¢, E. Guthrie, Fuel Cycle Costs in a Graphite-Mcderated Siightly
Enriched Fused Salt Reactor, ORNL CF-=59-1-13 {(Jan. G, 1959).

-19-

UNCLASSIFIED
ORNL~ LR-DWG 35088A

 SEAL WELD

- FUEL PUMP

. SALT CONTAINER
-

— GAS HEATING SHELL
GRAPHITE
CORE

FUEL PASSAGE
(TYPICAL)

Fig. 1.1.10, Diagram of Reactor System.
UNCL ASSIFIED
ORNL-LR-DWG 350864

COOLANT SALT PUMP

REHEATER COOLANT SALT CELL
| / EVAPORATORS

PRE-HEATING LOOP ( /

EQUIPMENT 1 T ~
REMOVAL HATCH y 4 S
\ WINDOWS | |& [):)
2\ - N
B
FUEL ENRICHERA & ~ |
T=—STEAM PUMP
fi\; 3 REHEAT LINE
CELL COOLINGA - —s= TURBINE - GENERATOR
\

3 HIGH PRESSURE LINE

/ & kl_

ACCESS LOCK —] L IA: :JJ

FUEL PUMPS (4) - { ]T

HEAT EXCHANGERS (4} i BD

COOLANT SALT LINES N
T~

AN

N

FUEL FILL-AND-DRAIN TANKS

O O )

SUPERHEATERS

Fig. 1.1.11. Plan View of 760-Mw (th) Graphite-Moderated Molten-Salt Power Reactor Plant,

-OZ-
UNCLASSIFIED
ORNL-LR-OWG 35087A

. :

COCLANT PUMP

R
1§t

SUPERHEATER

STEAM PUMP

J
.I.Z-

FUEL PUMP —]

REACTOR M HEAT EXCHANGER

FEET

)'/\
O
.

e

4

1
/ :
1
EVAPCRATORS

Fig. 1.1.12. Elevation View of 760-Mw (th) Graphite-Moderated Molten-5alt Power Reactor Plant.
- 22 -

and the electrical generation systems, the control room and the fill-and-
drain tanks for the reactor fluids. The plant characteristics are listed

below:
Fuel
Fuel carrier
Neutron energy
Active core
Fuel equivalent diameter
Moderator

Thermal shield

Primary coolant

Power

Electric (net)

Heat
Regeneration ratio

Clean (initial)
Estimated capital costs
Refueling cgele at full power
Shielding

Control

Plant efficiency
Fuel conditions, pump discharge
Steam -system (Loeffler)
Temperature
Preséure

Secondary loop fluid (coolant salt)

1.30% U235Fh (initially)
LiF-BeF,-UF ) (70-10-20 mole %)
Near thermal

1k £t

12.25-Ft-dia, 12.25-ft-high
graphite cylinder with 3.6-
in.-dia holes on 8=in.
centers

12 in, of iron

Fuel solution circulating at

35,470 gpm

315 Mw
T60 Mw

0.79
$79,250,000 or $252/kw
Semicontinuous

Concrete room wall,
9 £t thick

Temperature and fuel
concentration

41.5%
1225°F at 105 psia

1000°F with 1000°F reheat
2000 psia
LiF-BeF, (65-35 mole %)

Structural materials

Fuel circuit INOR-8

Secondary loop INOR=8

Steam generator Croloy steel (2.5% Cr, 1% Mo)

Steam superheater and reheater INOR-8
Temperature coefficient, (Ak/k)/°F Negative
Specific power 1770 kw/kg
Power density 117 kw/liter
Fnel inventory

Initial (clean) 700 kg of e
Critical mass, clean 178 kg of y°32
Burnup Unlimited

The off-gas system provides for the continuous removal of fission-
product gases and serves several purposes. The safety of the system in
the event of a fuel spill is considerably enhanced if the radicactive
gas concentration in the fuel is reduced by stripping the gas as it is
formed. Further, the nuclear stability of the reactor under changes of
power level is improved by keeping the high-cross-section Xe135
continuously at a low level. Finally, many of the fission-product poisons
are, in their decay chains, either noble gases for a period of time or
end their decay chains as stable noble gases, and therefore the buildup
of poisons is considerably reduced by gas removal.

Details of the nuclear calculations pertinent to this design study

are presented irn a subsequent section of this chapter.

NUCLEAR CALCULATIONS

Reactivity Effects in Experimental Molten-Salt Reactor. Various

reactivity effects in the experimental reactor described above were
calculated. The first calculation was an estimate of the reactivity

effect of draining the first layer of diphenyl coolant from the thermal
-2l -

shield. This draining resulted in a loss in reactivity, Sk/k, of 1.8%%.
Another reactivity calculation was performed to estimate the effect of
inserting 100 g of U235 in the form of UFh at the center of the reactor.
This resulted in a gain in reactivity, ¢5kafi of 0.39%. The Cornpone
program was used for these reactivity calculations.

A temperature coefficient was also calculated for the experimental
reactor by using the procedure described in the Molten-Salt Reactor

Program Status Reportoh The over-all temperature coefficient of reactivity
was estimated to be =6.75 x 1077 ( Sk/k;/°F,

Neutron and Gemma-Ray Activity in the Secondary Heat Exchanger of the
Interim Design Reactor. The present specifications for the Interim

Design Reactor call for the use of sodium as the heat transfer fiuid in
the first intermediate heat exchange loop that links the reactor to the
steam system. There are several disadvantages in the use of sodium,
among which are its flammability, incompatibility with the fuel salt
(uranium is reduced and may precipitate), ard a 15<hr half-life, highe
level, beta-and-gamma activity that is induced by exposure of the sodium
to delayed neutrons in the primary heat exchanger.

The use of the non-uranium-bearing salt mixture LiF=-BeF2 (6337
mole %) in the first intermediats loop as an alternative to the sodium
has accordingly been studied. This salt is an excellent heat transfer
medium; it is compatible with the fuel; and it is nonflammable. The
nature and amount of the induced radiocactivity comld therefore be the
decisive factor in the choice between the salt and the sodium.

A Cornpone calculation was made in which the neutron capture and
leakage parameters for the primary heat exchanger treated as a subceritical

reactor were estimated. This calculation was followed by a Sorghum

l"Mo'J.'t:en==E‘fia].'E; Reactor Program Status Report, p 179, ORNL=263%4

(Nov. 12, 19587,

- 95 .

calculation in which the spectra of the four main groups of delayed
neutrons were used separately to estimate the distribution of the captures
of these neutrons in fuel, metal, and coolant. With these data, an MIT
Practice School group estimated the neutron and gamma-ray activity to be
expected in the secondary heat exchanger.5 Their results are summarized
in Table 1.1.2. It may be seen that the gamma-ray dose from the molten-
salt coolant is only one-half that expected from the sodium. Further,
the decay rate in the salt is orders of magnitude greater. It was
estimated that following cessation of circulation of the coolant it would
require about two weeks for the activity of the sodium system to reach

a level of 7.5 mr/hr, whereas the salt system would reach this level in

5 to 10 min.

The possibility that neutrons might be generated in the secondary
heat exchanger by 7-n reactions was investigated by the MIT group. The
sodium would not produce any neutrons, but the possibility that the 1.63-
Mev gamma ray from F20 might lie sufficiently clese to the threshold of
the 7-n reaction in Be9 to produce a significant neutron activity was
studied carefuliye It was concluded that, on the basis of presently
avallable data, no appreciable neutron activity need be expected}

Effect of Ion-Exchange Processing of Rare-Earth Fission Products on

Performance of Interim Design Reactor. The present specifications of the

interim design reactor call for processing of the fuel salt (600 ft3) at
the rate of 1.6 ft3/day (600 ft3/yT) by the flucride=-volatility process
which would recover all isotopes of uranium from the salt. The base

salt, which would contain 62.5 mole % LiYF, 36.5 mole % BeF,, and 1 mole %

ThFh, together with the fluorides of fission products, corrosion products,

5D. J. McGoff et al., Activity of Primary Coolant in Molten Salt
Reactor, EPS-X-400, Oak Ridge, Tennessee, MIT Engineering Practice
School, UNCN-ORGDP (1958)

-2 -

Table 1l.l.2. Gamma-Ray Activity in the Secondary Heat Exchanger

Dose Rate
Gamma~-Ray Source (r/sec)
Sodium
Primary photons
1.37 Mev 1.19
2.75 Mev 3.23
Secondary photons
From l.37-Mev primary photons ,2.48
From 2.75=Mev primary photons 4.33
Total 11.2 )
Molten salt .
Primary photons, 1.63 Mev 1.51
Secondary photons from 1.63-Mev
primary photons 3,50
Total 5.0

and impurities, would be discarded. The high cost of Li7 (0.01% Li6)

prohibits the use of this process at the rate of more than about 1.6 £13

per day.
It has been proposed, -alternatively, to remove the rare-«earth

fission prcducts by ion exchange with CeF The fuel stream returning

3.
to the reactor system from the processing plant would contain approximately

0.15 mole % CeF Cross sections for cerium were computed from the

3° ;

6Molten-Salt Reactor Program Status Report, p 232, ORNL-263k4

(Nov. 12, 1958),

- 27 -

 resonance parameters, and the poisoning from 0.15 mole % CeF3 was
determined by means of a Cornpone calculation. The poisoning was
found to be negligible.

The long-term nuclear performance of the interim design reactor
with CeF3 processing was then studied by means of a Sorghum calculation.
A processing rate of 18 ft3/day (600 ft3/month) was selected, and it
was assumed that the fuel stream returning to the reactor system
contained negligible concentrations of high-cross-section fission
products. The results of the calculation are shown in Fig. 1.1.13,
where regeneration ratio, critical inventory, and cumulative net burnup
are plotted vs time of operation in years, and are compared with the
corresponding performance of the same system utilizing the fluoride=~
volatility processing method.

It may be seen that the ion-exchange processing method gives
substantial improvements in performance. The average critical inventory
is about 50% less, and the cumulative net burnup is about 30% less.
The calculation did not take into account the poisoning from samarium
isotopes and non-rare-earth fission products that would not be
removed by this process.

Multigroup Oracle Programs for Heterogeneous Reactor Computations.

The 3l-group Oracle program Cornpone was modified to permit evaluations
of heterogeneous reactors on a routine basis. A new boundary option
providing for a zero net current of neutrons was added, which permits
the calcuiation‘of an infinitely long unit cell in an infinite lattice

- in cylindrical geometry. A subprogram was added to compute, for each
neutron group, the ratio of the average flux in the fuel region to

the average flux in the cell, and the ratio of the average flux in the
moderator to the average flux in the cell. These ratios were designated

"group disadvantage factors'" for fuel and moderator regions, respectively.

The disadvantage factors were then applied in a calculation of

"finite, homogeneized" reactors. Fuel and moderator materials were

-28-

UNCLASSIFIED
ORNL-LR—-DWG 35672
FUEL SALT: LiF -BeF, (63-37 mole %) 4 mole % ThF, + UF,
BLANKET SALT: LiF -BeF,~ThF, (74-16-13 mole %)
POWER: 600 Mw (th)
PLANT FACTOR: 0.8
TOTAL FUEL VOLUME : ~600 ft°
CORE DIAMETER: 8 fi
CORE VESSEL: INOR-8, %3 -in. THICK
BLANKET THICKNESS: 2 ft

0.8
o
= o— w— N — — — -—-—-—---—.____
g 07 1= ION-EXCHANGE PROCESSING;
z s~ 18 £5/day
= /
= /
o
L /
w 0.6 .
o FLUORIDE —~VOLATILITY
i3 PROCESSING, 1.66 f1/day
0.5
1000 T
, @ ° /-\-_____——"/FLUORiDE—VOLATILITY
@ 3
x o PROCESSING, 4.66 ft~/da
S 2 800 /day
=z @
l-I>J @
Z g 700
3 5
g u
IS L4 \
= 4 600 2
E oo \
E O \ ON-EXCHANGE PROCESSING, Jp—
= 500 —\tffls/day — o o
400
1600 J i
-5 / —=
Y 3 4200 [ FLUORIDE-VOLATILITY y -
o5 PROCESSING, 1.66 f1~/day -
> o 4—’
E < 800 i
4 % -+~ TION-EXCHANGE PROCESSING,
g = - 3
= % 400 :’/ 18 ft /de
Q @ ’/
0
0 4 8 12 16 20

TIME OF OPERATION (year)

Fig. 1.1.13, Long-Term Performonce of Spherical, Homogeneous, Two-Region, Molten-
Fluoride-Salt Reactor Fueled With U235.

- 29 -

mixed in proportions corresponding to those used in the unit cell,

The diameter of the equivalent spherical core and the thicknesses of
core vessel, blanket, reflector, etec., were specified, as usual. A
subprogram was added by means of which the cross sections of the fuel
and moderator materials were then multiplied groupwise by the appropriate
disadvantage factors. From that point the calculation proceeds in

the usual manner.

The modified program was tested by computlng the multlplicatlon
constant for the ORNL Graphite Reactor. Cross sections for U23 metal
were estimated by taking into account resonance saturation and Doppler
broadening. Neutron energies ranged down to 0.079 ev, the thermal energy
(11800F) in the molten-salt reactors. Cross sections for aluminum were
also added. With the use of specifications given by Ramsey and Cagle,7
k was estimated to be 0.967 for the reactor at 1180°F. This value appears
to be reasonably correct. Further calculations toc permit éstimation
of the multiplication at the operating temperature (3500F) are planned.
Cross sections for U238 in a mixture containing 70 mole % LiF, 10 mole %
BeFE, and 20 mole % UFh were computed and added to the cross-section
library.

With the use of the new code and the previously calculated cross
sections for various ThFh melts, several uranium and thorium cycle
reactors were computed, as described in subsequent sections of this
chapter.,

In order to study the long-term behavior of heterogeneous reactors,
the Oracle program Sorghum was modifiéd. Provisien for absorption of
neutrons by graphite was made, and group disadvantage factors derived
from the modified Cornpone program were applied separately to fuel and

graphite. The fission product Smlhg was treated separately from the

TM. E. Ramsey and C. D. Cagle, Research Program and Operating Experience
on ORNL Reactors, Geneva Conference Paper 450 11958).

- 30 -

other figsion products. For evaluating plutonium converters, the

240 241
u  , Pu 7,

existing program was modified to compute ingrowth of P and °

Puahz, in place of Pa233, U233, and U23h.

The modified program works well with bare and blanketed reactors,
but the critical calculation was found to be unstable in the case of
graphite-reflected reactors. In the reactor model embodied in Sorghum,
the net leakage in each group is assumed to be proportional to the
space-averaged flux in the corresponding group. This is a reasonably
good approximation in bare and blanketed reactors in which the sloving
down in the second region is small, and the net leakage in any given
group is nearly independent of the leakage in groups of higher energy.
In graphite-reflected reactors, however, this approximation is not
good. Two cases of graphite-moderated reactors have been studied for
periods of up to ten years by means of the modified program described
in subsequent sections of this chapter.

Oracle Program GHIMSR for Gamma-Heating Calculations. A code was
written to expedite gamma-heating calculations, in which an integral-

spectrum method is used to calculate the heating in a spherical reactor
system. Gamma-energy absorption coefficients for 12 groups are stored
and used to obtain average absorption coefficients for the reactor and
core vessel,

The necessary inputs for the calculations include the radius of
the reactor, the core and core vessel compositions, and gamma source
distributions in space and energy. The time of computation for one
source is approximately 3 min. The program was checked against hand
computations and was found to be performing satisfactorily.

Initial and Long-Term Nuclear Performance Without Fuel Processing

of Low-Enriéhmenfi, One-Region, Graphite-Moderated, Unreflected, Molten-

Salt Reactor. The modified Cornpone program described above was used

to estimate the initial nuclear performance of the graphite-moderated,

molten-salt reactor described above. As stated, the reactor has a

..31..

spherical core 14 ft in diameter enclosed in an INOR-8 pressure
vessel 1 3/L4 in. thick surrounded by an iron thermal shield 12 in.
thick. Fuel channels 3.6 in. in diameter penetrate the core and

are arranged in a square array measuring 8 in. between centers. The
fuel solution is LiF-BeF,-UF, (70-10-20 mole %), and the calculation
indicated that a fuel enrichment of 1.3 mole % is required to make
the system critical.

The slowing=-down, leakage, and disadvantage parameters obtained
from the Cornpone calculation were used in a Sorghum calculation to
study the long-term nuclear performance of the system without fuel
processing. An extract from the results is given in Table 1l.l.3,
and key parameters are graphed in Fig. 1.1l.1lk.

The accumulation of fission products progressively poisons the
reactor. The regeneration ratio, initially 0.79, falls steadily to
0.5 after 7200 megawatt years (Mwy) of cumulative power generation.
The critical inventory falls, at first, because of the ingrowth of
plutonium isotopes, but it then rises steadily from an initial value
of about 1000 kg to about 4500 kg. The cumulative net burnup amounts
to about 1500 kg in 7200 Mwy; the enrichment of the fuel added varies,
but averages 60% during the first five years.

The cost figures given in ref 2 were calculated on the basis of
a one-group nuclear calcuhation3 in which progressive hardening of
the neutron spectrum was ignored. Revised cost estimates based on
the multigroup calculations described here are expected to show a fuel

cost of about 2.5 mills/kwhr.
Initial and Long-Term Nuclear Performance Without Fuel Processing

of One-Region, Graphite-Moderated, Unreflected Th 232 -Conversion, Molten-

Salt Reactor. The nuclear performance of the reactor described in the
preceding section fueled with the mixture LiF-BeF,-ThF, (67-16=13 mole %),
together with sufficient U235F U238Fu to make the system critical,

was studied. The initial critlcal inventory was found to be 829 kg,

and the initial regeneration ratio was 0.79. The long-term performance
Table 1.1,3. Initial and Long=Term Nuclear Performance Without Fuel Processing of Low=-Enrichment, One-Region,
Graphite-Moderated, Unreflected, Molten-Salt Reactor

Core diameter: 14 ft
Power: 760 Mw (th)

Fuel volume: 900 i3

Plant factor: 0.8

After Cumulative Power

Initial State Generation of 480 Mwy

After Cumulative Power

Generation of 2400 Mwy

After Cumulative Power

Generation of 4800 Mwy

After Cumulative Power

Generation of 7200 Mwy

Neutron
Inventory ' Neutron Neutron Neutron Neutron
Absorption Inventory . Inventory . Inventory ) Inventory .
{kg) .. . Absorption i Absorption . Absorption Absorption
Ratio (kg) Ratio* (ko) Ratio* (kg) Ratio* (ko) Ratio*
Fissionable isotopes
u23s 788 1.000 652 0.578 1,145 0.486 2,275 0.561 3,715 0.630
py239 106 0.409 248 0.426 384 0.346 528 0.290
pu24l 4 0.013 63 0.088 12 0.093 143 0.080
Fertile isotopes
y23s 55,150 0.790 54,992 0.672 54,443 0.533 53,870 0.455 53,386 0.403
py240 16 0.056 55 0.140 62 0.122 66 0.102
Fuel carrier
Li/ 5,760 0.052 5,760 0.036 5,760 0.016 5,760 0.009 5,760 0.006
g9 37,900 0.026 37,900 0.022 37,900 0.018 37,900 0.017 37,900 0.016
Moderator
Be? 1,060 1 0.001 1,060 0.001 1,057 0.001 1,057 0.000 1,057 0.000
cl? 0.051 0.036 0.017 0.011 0.008
Fission products
sm14? 0.029 0,010 0.078 0.009 0.192 0.009 0.415 0.00%
Other 781 0.026 908 0.085 1,815 0.125 2,723 0.150
Parasitic isotopes
U236 27 0.002 109 0.009 228 0.016 379 0.024
Np237 5 0,001 18 0.004 37 0.006
py242 5 0.002 n3 0.041 235 0.071 314 0.078
Core vessel and leakage 0.158 0.133 0.103 0.089 0.080
Neutron yield, 7 2.078 1.996 1,973 1.928 1.882
Total fuel inventory, kg 788 762 1457 277 4386
Cumulative net burnup, kg 58 347 817 1400
Net fuel requirement, kg of U235 788 820 1804 3588 5786
Regeneration ratio 0.790 0.728 0.673 0.577 0.505

*Neutrons absorbed per neutron absorbed by fissionable isotopes.

ce
_33.

UNCLASSIFIED
ORNL-LR-DWG 35673

0.8
=
2 o7
- ~N
O
b
o
=
w 0.6 ™~
G}
w
5 \
FUEL SALT: 1_iF~-BeF2—UF4
{(70-10-20 mole %)
0.5 L CORE DIAMETER: 12.25 ft

FUEL-TO~-MODERATOR VOLUME

TOTAL FUEL VOLUME: 900 ft°
TOTAL POWER: 760 Mw {th)
PLANT FACTOR: 0.8

4000 //

CRITICAL INVENTORY
Fd

2000 CUMULATIVE

/ NET BURNUP

0 2000 4000 6000 8000
TIME OF OPERATION (Mwyr)

CRITICAL INVENTORY AND BURNUP
(kg of fissionable isotopes)

Fig. 1.1.14. Long-Term Nuclear Performance Without Processing of a One«Region,
Unreflected, Heterogeneous, Graphite-Moderated, Molten-Fluoride-Salt Reactor Fueled
With 1.30% Enriched Uranium,
- 3L -

of the system was studied by means of the Oracle program Sorghum.
An extract from the calculations is given in Table 1.l1.4.

For a cumulative power generation of 7200 Mwy, the critical
inventory rose to 1416 kg, and the cumulative net burnup amounted to
1083 kg. This performance is substantially better than that exhibited
by the corresponding low-enrichment system, which required an inventory
of 4386 kg and had a cumulative net burnup of 1400 kg. The disparity
between the two systems is expected to increase with time.

Initial Nuclear Performance of One-Region, Graphite-Moderated,
Graphite-Reflected, Th?32-Conversion, Molten-8alt Reactor. The initial

nuclear characteristics of a heterogeneous, graphite-moderated

and -reflected, one-region, molten-salt reactor were studied. The
reactor considered in the calculations consists of a cylindrical core,
15 ft in diameter and 15 ft high; surrounded by a 2.5-ft-thick graphite
reflector contained in a l-in.-thick INOR-8 pressure shell. The core
is penetrated by cylindrical fuel passages arranged on an 8-in.
triangular lattice paraliel to the core axis. The resulting unit cells
are hexagonal and 15 ft long. This system has not been optimized,

and a much better design might conceivably be evolved. The initial
nuclear characteristics of the system are compared in Tables 1.1.5 and
1.1.6 with the characteristics of the interim design reactor.

The modified Oracle program Cornpone was used to calculate k and
group disadvantage factors for the fuel and graphite. These results
were then used for complete reactor calculations in spherical geometry
of a core having the same volume as the actual cylindrical core. Mean
"homogeneized" densities (atoms/cm3) were used for each element.

It may be seen that the regensration ratio of the heterogeneous
reactor is more favorable than that of the interim design reactor. The
initial critical inventory of U2359 on the other hand, is 60% higher than
in the interim design reactor. It is estimated that a power level of
760 Mw (th) can be attained in the heterogeneous system compared with
600 Mw (th) in the homogeneous reactor.

35

Table 1.1.4. Initial and Long=-Term Nuclear Performance Without Fuel Processing of One-Region, Graophite-Moderated, Unreflected,
Thzaz-Conversion, Molten-Salt Reactor

Fuel volume: 900 ft°3
Plant factor: 0.8

Core diameter: 14 ft
Power: 760 Mw (th)

After Cumulative Power
Generation of 7200 Mwy

After Cumulative Powaer
Generatian of 4800 Mwy

After Cumulative Power

Generation of 2200 Mwy

After Cumulative Power

Initial
nitial State Generation of 480 Mwy

Neutron
Inventory Absoroti Invent Neutron : , Neutron | Neutron Neutron
(kg) . r-p 1:t:m vekn ory Absorption nventory Absorprion nventory AI::SOrpfion lnvenfory Absorpfion
atio (kg) Ratio* (ke) Rotio* tkg) Ratie* (kg) Ratio*
Fissioncble isctopes
y233 129 0.175 424 0.508 541 0.529 593 0.470
y23s 829 1.000 702 0.819 473 0.480 570 0.457 817 0.513
pu239 1 0.006 3 0.012 4 0.014 7 0.017
Fertile isotopes
Th432 38,438 0.783 38,438 0.767 38,438 0.700 38,437 0.597 38,437 0.511
Y234 4 0.001 42 0.009 97 0.016 143 0,019
Y238 64 0.008 78 0,008 108 0.013 141 0.016 187 0.020
Fuel carrier
Li’ 6,328 0.054 6,328 0.052 6,328 0.045 6,325 0.035 5,328 0.027
- g1? 37,570 0.025 37,570 0.025 37,570 0.024 37,555 0.022 37,570 0.021
Moderator
Be? 1,840 0.001 1,840 0.001 1,840 0.001 1,840 0.001 1,840 0.001
cl2 0.051 0.049 0.043 0.034 0.027
Fission products 181 0.034 908 0.154 1,815 0,263 2,723 0.336
Parasitic isotopes
y23s 3 0.004 10 0.013 174 0.020 240 0.026
Np237 7 0.004 19 0.008 32 0.012
Miscellaneous
Pa233 18 0.009 16 0.008 14 0.006 12 0.004
Core vessel and 0,149 0,148 0.141 0.128 0.116
leakage
Neutron yield, 1 2.071 2.099 2.155 2.146 2,120
Total fuel inventory, kg 829 832 900 1,115 1,417
Cumulative net burnup, kg 69 287 638 1,083
Net fuel requirement, kg 829 ¢01 1,187 1,753 2,500
of U235
* Regeneration ratio 0.791 0.768 0.714 0.623 0.546

*Neutrons absorbed per neutron absorbed by fissionable isotopes,
Table 1.1.5.

- 36 -

Initial Nuclear Performance of One-Region,

Graphite=-Moderated Graphite-Refiected, Th232-Conversion, Molten-Salt
Reactor Compared With That of the Two-Region, Homogeneous Interim Design

Reactor

760-Mw (th) Th>3°-

Conversion Reactor

600-Mw (ti) Interim
Design Reactor

Core Blanket
Thorium content 13 mole % 1 mole % 13 mole %
Core size 15 x 15 ft 8 ft sphere 2=in.=thick
cylinder layer

Fuel channel

diameter 5 in.
Fuel volumetric

fraction in core 0.3543 All Rlanket
Total fuel volume 1480 £t3 607 £t
Volume of fuel 3 3 3

in core 9o ft 268 ft 913 ft
Volume of fuel 3 3

external to core 540 ft 339 ft
Volume of graphite 5340 ft3
Weight of graphite 623,000 1b
Thermal fissions 2% 11%
Thermal absorptions 5%
Neutron yield, YZ 2.073 1.80
Regeneration ratio 0.79 0.63

Table 1l.1l.6. Initial-State Inventory and Neutron Absorptions

760-Mw (th) Tho3° - 600-Mw (th) Interim Design Reactor
Conversion Reactor Core Blanket
Inventory Neutron Inventory Neutron Inventory Neutron
(kg) Absorption (kg) Absorption (kg) Absorption
Ratio¥* Ratio¥* Ratio¥
Figsionable isotope
ye3? 970 1.000 604 1.000
Fertile isotopes
U238 81 0.010 45.3 0.039
THhe 32 41,000 0.782 2,100 0.36h 30, 500 0.228
Fuel carrier
Li7 5,250 0.038 3,920 5,030
Fld 4, 000 0.031 24,000 25,100
Moderator
Be9 3,890 0,050 3,008 0.102 1, 460 0.011
cte 283,000
Core vessel and leakage 0.162 0.052

¥*
Neutrons absorbed per neutron absorbed by fissionable isotopes.
- 38 -

l.2. COMPONENT DEVELOFPMENT AND TESTING

SALT-LUBRICATED BEARINGS FOR FUEL PUMPS

Hydrodynsmic Journal Bearings. The fourth test of a salt-lubricated
hydrodynamic bearing was completed and terminated on schedule after 1000
hr of operation. The test was performed with salt No. 130 (LiF-BeFE-UFh,
62-37-1 mole %) under steady-state conditiens: temperature, 1200°F;
Journal speed, 1200 rpm; bearing radial. load, 200 1lb. Postrun examination

revealed that slight rubbing had occurred between the bearing and journal.
Since this test bearing was started and stopped only a very few times

as compared with the bearing which was tested previously and started and
stopped 87 times (rather than 94 times, as reported erroneously in the
previous reportl) during a total operating time of 784 hr, it appears
that the rubbing marks may not be attributed entirely to starting and
stopping operations. Wear on the journal bemering used in the fourth test
appears to be slightly less than on the Jjournal bearing which was

sub jected to 87 starts and stops.

A fifth test of 360 hr duration, for which new bearing and journal
parts fabricated of INOR-8 were used, was also completed. Salt No. 130
was used as the lubricant and the system was maintained at 1200°F. The
Journal speed and bearing radial load were varied over wide ranges (600
to 2700 rpm end O to 500 1b, respectively). Thirty combinationg of speed
and load were applied for periode of 2 hr each. There were no difficulties
observed during these tests. Postrun examination of the bearing revealed

that slight rubbing occurred between the bearing and journal.

1P. G. Smith, W, E. Thomas, and H. E. Gilkey, MSR Quar. Prog. Rep.
Octo 31’ 1958, ORNL-2626, P 1.90

- 3"1..

A sixth test with the bearing and journal used in the fifth test
is under way to investigate bearing performance at steady-state conditions
of journal speed and bearing radial load (1200 rpm and 200 1b) and at
temperatures varying in 5OOF intervals from 1200 to 15OOOF. Bearing
performance is being observed for 4 hr at each temperature level.

All these tests have been performed with a radial clearance of 0.005
in. between the bearing and journal surfaces, as measured at room tempera-
ture. The continuing analytical studies and surveys of recent literature
indicate that it may be possible to improve the performance of this
bearing by increasing the clearance to 0.007 in.

Hydrodynamic Thrust Bearings. Construction of a thrust-bearing

tester is approximately 80% complete. The tester shown schematically
in Pig. 1.2.1, utilizes a PK type of centrifugal pump (without impeller)
that was modified in the impeller region to permit installation of the
test thrust bearing. The tester includes provisions for varying bearing
load, journal speed, and test temperature. The thrust runner is mounted
on the lower end of the pump shaft and the stationary member of the thrust
bearing is supported by a load actuator. The load actuator consists of
two concentric bellows with common end flanges. Pressure applied
internally to the bellows assembly resultis in an axial force {design
maximum, 500 1b at 1200°F) applied to the thrust bearing. The first
bearing to be investigated will be of the step type.

Test Pump Equipped with a Salt-Lubricated Journal Bearing. A PK

type of centrifugal pump is being modified to include a molten-salt-
lubricated journal bearing near the impeller, an upper bearing (both
radial and thrust) lubricated either with oil or grease, and an overhung
drive motor. The molten-salt-lubricated bearing will be tested through
the full scope of loadings and clearance problems existing in a pump

operating at high temperatures.
- 40-

UNCLASSIFIED
ORNL-LR-DWG 35674

IEMARAERS HIS NN
oL INmmmefi | |

EQUALIZER

[0
SN L
PRESSURIZING LINE
T !‘I T
LOAD ACTUATOR N Wi /
T
N

it

THRUST-BEARING
STATOR

THRUST-BEARING
RUNNER

Fig. 1.2.1. Diagram of Thrust-Bearing Tester,

")"'l"

Hydrostatic Bearings. The previously described2 water tests of

hydrostatic bearings were completed and a report describing the tests
and giving the results is being prepared.3 The final test, for which
the bearing radial clearance was 0.005 in., was made at various speeds
with various system flow resistances. Postrun examination revealed that
slight rubbing occurred between the bearing and journal. Further work
with this type of bearing (and also with rotating-pocket hydrostatic
bearings) has been deferred because of the favorable results being
obtained with the less complicated hydrodynamic type of journal bearings
and the increased emphasis on thruste-bearing testing.

Bearing Mountings. Investigations of bearing mounts to accommodate

dimensional changes resulting from differences in coefficients of
expansion of the bearing and mounting materials have been de-emphasized
because of the favorable results being obtained with the use of INOR-8
for Jjournal, sleeve, and mount. Promising techniques have not yet been
found.; The "Thermoflex" journal bearing mounts designed by Westinghouse
Electric Manufacturing Company are not yet available in the sizes of

interest.

CONVENTIONAL BEARINGS FOR FUEL PUMPS

Tests with Dowtherm "A" as the pump lubricant were completed, as
stated previously,h and a final report5 on the tests and results has been

issued.

2P. G, Smith, W. E. Thomas, and H. E. Gilkey, MSR Quar. Prog. Rep.
Oct. 31, 1958, ORNL=2626, p 19.

3P. G. Smith and H. E. Gilkey, Hydrostatic Journal Bearing Water
Tests « Modified PK~-A Pump, ORNL CF~59~2-1 (to be issued).

hD. L. Gray, W. E. Thomas, MSR Quar. Prog. Rep. Oct, 31, 1958,

ORNL'2626, p 20.

5w. E. Thomas, Dowtherm "A" Journal Bearing Test, ORNL CF=58-10=40
(Nov. 6, 1958).

- 43 -

MECHANICAL SEALS FOR FUEL PUMPS

Labyrinth and Split-Purge Arrangement. The previously described
results of tests of a PK type of pump with a labyrinth seal and split-

purge gas arrangement were correlated and a test report was issued.
Bellows«Mounfied Seal., The modified Fulton-Sylphon bellows-mounted
seal7 being subjected to an endurance test in a PK~P type of centrifugal
pump has accumulated an additional 2520 hr of operation since the previous
report period, for a total of 10,630 hr. The pump has been operating
at a temperature of 1200°F, a shaft speed of 2500 rpm, and a NaK flow rate
of 1200 gpm. The test=seal leakage continued to be negligible throughout
the peridd. Three pump stoppages occurred; one that was caused by a
power failure and two that were for replacing brushes in the drive motor.

PUMP ENDURANCE TESTING

An MF type of centrifugal pump has been circulating fuel 30 (NaF'-
ZrF) -UF, , 50=46=l mole %) in continuous operation for 13,500 hr since
June 26, 1957, without maintenance. For the past 11,500 hr the pump has
been operated under cavitation damage conditions. The fuel 30 flow rate
is 645 gpm; the shaft speed is 2700 rpm; the pump tank cover gas is at
a pressure of 2,5 psig; and the temperature of the circulating salt mixture
is 1200°F, A recently pu‘blish.ed8 operational resumé,presents an account
of various pump starts and stops and changes in operating conditions.
During this quarter the pump was stopped three times. One stop was

- Gb L. Gray, Test of PK Pump Split Purge Gas Labyrinth Seal, ORNL
CF-59-1-5 (Jan. 17, 1959).

7b. L. Gray, MSR Quar. Prog. Rep., Oct. 31, 1958, ORNL-2626, p 21.

8(}° G. Smith, Resumé’MFnj Pump Test Operation in MFF Loop, ORNL
CF-58-11-38 (Nov. 17, 1958). |

...],’_3-

momentary because of a power outage. It was stopped the second time
for 5 min to replace the tachometer generator on the drive motor. The

third stop was of 8 min duration to replace drive-motor brushes.

FROZEN-LEAD PUMP SEAL

The small frozen-lead pump seal being tested on a 3/16-in.-dia shaft,
as described previously,9 has operated continuously since it started
operstion on June 13, 1958. As of January 1, 1959 the accumulated
operating time was L4L84LO hr. Following slight leakage of lead during the
first 100 hr of operation, there has been no furthef leakage.

A large frozen-lead seal on a 3 l/h-in.-dia rotating shaft has been
placed in operation. The frozen-lead seal retains molten lead at a
temperature of approximately lOOOOF in a tank. The test equipment is
shown in Figs. 1.2.2 and 1.2.3. The three separate cooling coils used to
maintain the lead below its melting point in the bottom of the sealing
gland are shown in Fig. 1.2.4. Thermocouples, located between the cooling
coils, may also be seen in Fig. 1.2.k.

The ghaft is mounted in two self-aligning, pillow=-block bearings
attached to the support frame. A 10-hp Louis Allis motor and magnetic
clutch are conmected to the bottom end of the shaft through an adjustable
coupling.

As may be seen in Fig. 1l.2.2, the sealing gland incorporates a long,
tapered amnulus, which narrows to provide a clearance between the shaftl
and the sealing gland of a few thousandths of an inch. The upper end of
the shaft rotates in the lead tank. An inert atmosphere of argon at
atmospheric pressure or higher can be provided over the surface of the

molten lead in the tank. Either water or air can be circulated through

9. B. McDonald, E. Storto, and J. L. Crowley, MSR Quar. Prog. Rep.
Oct. 31, 1958, ORNL-2626, p 23.

-44 -

UNCLASSIFIED
ORNL—LR— DWG 35675

Te i LEAD TANK
l RN l
SNy
S \
% Tc9
}
3
N
N
Tc3 % T 10

— 3Y,-in-DIA SHAFT

AR
[ )

im fi\\\ § th
z:l
Te Tc2
TeS Tc8
Tc 4 Te7
Tct7 Tce
‘l = |
) \ UPPER BEARING
-

COOLING COILS ON
SEALING GLAND

LOWER BEARING

C_—
SUPPORT P E T_\l
LATE &/ U
— ]
L COUPLING TO MOTOR

Fig. 1.2.2. Sketch of Frozen-Lead-Seal Test Equipment Showing Thermocouple Locations.
UNCLASSIFIED
PHOTO 32824

=28

L

n-Lead-Seal Test Equipment.

.45.

Fig. 1.2.3. Froze

-46-

" UNCLASSIFIED
| PHOTO 32826

Fig. 1.2.4. Sealing Gland, Shaft, Cooling Coils and Thermocouples for Frozen-Lead Seal Test.

..)_l_T.-

the cooling coils attached to the outside of the sealing gland. The flow
in each of the three cooling coils can be controlled separately. The
supply pressures of watef and alr are maintained at constant levels.

In preparation for filling with lead, the equipment was heated to
operating temperature. With the shaft rotating at 100 rpm, the tank was
filled with 120 1b of lead at 1000°F. At the present time, the shaft
is being run at 1300 rpm. The corresponding peripheral speed of the shaft
is 1100 fpm, which is about 10 times the peripheral speed of the 3/16-in.-
dia shaft operating in the small frozen-lead seal test. Water at a flow
rate of approximately 2 gpm is being used as the coolant for the seal
gland. Smell amounts of lead have leaked sporadically from the sealing
glafld.

Representative operating temperatures for the sealing gland and tank

at the thermocouple locations indicated in Fig. l.2.2 are listed below:

Thermocouple Temperature Thermocouple Temperature
Number (OF) Number (°r )
1 880 T 590
2 875 8 810
3 935 9 925
L 590 10 985
5 815 il 860
6 275 17 290

TECHNJQUES FOR REMOTE MAINTENANCE OF THE REACTOR SYSTEM

Mechanical Joint Development. Detailed descriptions of the three

types of mechanical joints being considered for use in a remotely maintainabie

reactor system were presented previously,lo‘as well as the results of

lQA. S. Olson, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-247k, p 20.

- L8 -

screening tests of the three joints, which were conducted under thermal
cycling conditions in a high-temperature loop that circulated fuel 30
(NaFnZ_thaUFh, 50=46-4 mole %) at temperatures up o 1_500° Rl
Additional tests were made in a high-temperature loop that circulated
sodium at temperatures up to 13OO°F°13

Tests were continued during the quarter on the two, previously
described,ll large freeze-flange mechanical joints for use in a L=in,
pipe line. The tests were conducted in a high-temperature loop that
circulated fuel 30 at temperatures up to 1300°F under thermal cycling
conditions. |

13 the freeze=flange clamps

Subsequent to the first thermal cycle,
were removed, and additional braces were welded onto the clamps to permit
greater clamping loads. The clamps were cleaned and given an oxidation-
resistant coating with the use of the "Black Magic" process. The coated
gsurfaces of the clamps and the bolts were then covered with a special
lubricant. Seal rings of annealed 28 aluminum were installed in each
Joint, and the seal was effected with a bolt loading of 200 ft=-1b per
bolt.

The cold leakage rates for these two large freeze-flange jointé
before preheating of the loop were 6 x 10='11 cm3 of helium per seEond
or less. The leakage rate after preheating of the loop to 13OOOF was
2 X 10“"8 cm3 of helium per second or less. During the loop test,
temperatures were measured at the points indicated in Fig. 1l.2.5. The

molten salt was circulated at a flow rate of approximately 4O gpm. The

1, S. Olson, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-24Th, p 20.

12W. B, McDonald, E. Storto, A. S. Olson, Screening Tests of Mechanical
Pipe Jointg for a Fused Salt Reactor System, ORNL CF=58=-8-33 (Aug. 13, 1950}.

13A. S. Olson, MSR Quar. Prog. Rep. Oct., 31, 1958, ORNL=-2626, p 25,

-49.

UNCLASSIFIED
ORNL—LR—-DWG 3454

(O FLANGE NO.1 THERMOCOUPLES

[ | FLANGE NO.2 THERMOCOUPLES
DOUBLE LINES INDICATE WELLS

Fig. 1.2.5. Diagram of Large Freeze-Flange Mechanical Joint Showing
Location of Thermocouples.
salt temperature was cycled 12times between 1100 and l300°F; each cycle
was of 2 hr duration. Representative temperatures measured during the
test are listed in Table 1l.2.1.

Table 1.2.1. Temperatures Measured During Thermal Cycling of
Freeze~-Flange Mechanical Joints in Development Test No. 2

Thermocouple Minimum Cycle Maximum - Cycle
Number Temperature Number Temperature Number
(°F) (°F)
Freeze«Flange Joint No. 1
1 450 8 540 2=3=1]
2 875 3-4-5-8-11~12 1025 9
3 980 3el=5=11=12 1150 2=23=526=T=3=10
L 980 3elje5-11-12 1145 9
5 880 = 3-4-5-809-11-12 1020 9
6 420 8 495 2-3
7 300 lei=5=829 335 3
8 245 8 290 a3
Freeze~Flange Jolint No. 2
9 510 h=5 585 2
10 900 -3 1055 9
11 995 3 1175 9
12 960 3=5-11-12 1135 9
13 910 3 1080 9
1Y 465 8«3 550 T=10
15 315 5=8 350 2-4=10=11

16 300 4=8 330 1=2«3=b-T=10-11

N

There was no indication of salt leakage during the test, but gas

leakage tests after the high-temperature salt test yielded leakage rates
3

in excess of the maximum alloweble leakage rate of 10_7 cm” of helium

per second. After the clamp bolts on both flanges were tightened to
again provide a bolt loading of 200 ft-1b per bolt, the flanges were
leak tested a fourth time. The leakage rates were 9.4 x 10-801113 of helium
per second or less and were thus again below the specified maximum
allowable rate,

Both joints were easily disassembled in about 15 min. The frozen
salt in each Jjoint was caked to the stainless steel insert screen, as
shown in Fig. 1.2.6. One fiange is shown with the screen and salt seal
removed in Fié. 1.2.7. The flange faces were sufficiently clean for
reassembly of the Joint.

Leakage rates were again measured after reassembly of the joints with
the same aluminum sealing rings. The leakage rate for joint No.l was
in excess of lO-7 cm3 of helium per second. The leakage rate for joint
No., 2 was 6 x 10"'8 em> of helium per second.

Remote Maintenance Demonstration Facility. Construction of the

remote maintenance demonstration facility iz expected to be essentially
complete by Juily 1. The General Mills remotely operated manipulator for
this facility will be delivered in February.

MOLTEN-SALT FEAT-TRANSFER.COEFFICIENT MEASUREMENT

The modifications reguired for salt-to-salt heat transfer tests in
the test facility for molten-salt heat-transfer-coefficient measurements

were completed, and 52 data runs were made. The heat transfer data
-52.

UNCLASSIFIED
PHOTO 45806

Fig. 1.2.6. Disassembled Large Freeze-Flange Joint With Screen and Sealing Ring in Place.

2535

UNCL ASSIFIED
PHOTO 45807

Fig. 1.2.7. Disassembled Large Freeze-Flange Joint After Removal of Screen.

-5)_|_-

obtained are currently being analyzed. A description of this test
facility and the results of a salt-to-liquid metal test were presented

previously;lh’15

TRIPLEX-TUBING HEAT EXCHANGER DEVELOPMENT

A seven-tube test heat exchanger was designed (Fig. 1.2.8) with
which to evaluate the triplex heat exchanger tubing being developed
by the Metallurgy Division for molten-salt or liquid metal=-to-steam
superheater applications. The triplex tubing consists of inner and
outer tubes with sintered metal powder or wire mesh in the annulus between
them. This tubing was designed to provide positive separation of fluids-
witfi minimum thermal resistance between the inner and outer tubes and to
provide a means of leak detection in that the pressure of a gas occupying
the voids in the sintered metal in the annulus could be monitored.
Details of the tube development program are included in Chap. 2.1 of
this report.

The design of a small heat exchanger test stand has been revised
to accommodate this heat exchanger for heat trénsfer and reliability

e#aluation tests.

EVALUATION OF EXPANSION JOINTS FOR MOLTEN-SALT REACTOR SYSTEMS

A total of six commercially available expansion joints has been
evaluated in the expansion joint test facility. A description of the

th, C. Amos, R. E. MacPherson, and R. L. Senn, MSR Quar. Prog.
Rep. Jan. 31, 1958, ORNL-24Th, p 27.

15J. C. Amos, R. E, MacPherson, and R. L. Senn, MSR Quar. Prog.
Rep. June 30, 1958, ORNL-2551, p 31. ‘

MOLTEN

SALT
[P S—

(B mam]

UNCLASSIFIED
ORNL-LR-DWG 35676

SPACERS

i .
/g-in. RODS ——
AN

‘ OUTSIDE TUBE : 1-in. GD, 0.065-in. WALL
“__J INSIDE TUBE: -in.OD, 0.065-in. WALL
ANNULUS: SINTERED NICKEL POWDER
FINAL DRAWN TUBE: 0.875-in. OD

TS HEAT EXCHANGER SHELL
3Y5-in. SCHED-40 PIPE

-gg-

Lo idml  moLTeN
| g ) SALT

_~GAS HEADER

i - e —- - - -

“HELIUM GAS CONNECTIONS FOR LEAK-DETECTION SYSTEM —

- = = — — — 83— — - = = ]

~ ~— NaK HEADER

Fig. 1.2.8. Triplex Tubing Heat Exchanger.
-56-
test facility and results of the first three tests were presented
previously.16 The results of the last three tests are presented in

Table l.2.2. The points of failure are shown in Fig. 1.2.9.

Table 1.2.2. Descriptions of Expansion Joints and Results of Tests

Vendor Cook Electric Co. Flexonics Corp. Flexonics Corp.
Material Inconel Inconel Stainless steel
Test fluid Molten salt Molten salt Sodium
Rated maximum

traverse 2 1/k in, 2 5/8 in. 2 5/8 in.
Test traverse 2 1/4 in. 2 5/8 in. 2 in.
Number of cycles

to failure 34 29 31

All the units were cycled over their maximum allowable traverse at
rated temperature and pressure of-l3OOOF and 75 psig, respectively,
except the Flexonics Corp. stainless steel unit, which was cycled over
approximately 75% of its rated maximum traverse. This reduction in the
cycle did not appreciably increase the life of the unit. The early
failure of all the units indicates that this type of expansion joint
design would not be satisfactory for molten-salt reactor application at
13OOOF and 75 psig without design modifications and further development
testing. The typical failure, shown in Fig. 1.2.10, indicates a high-
stress point that could be eliminated by redesign.

163. C. Amos, MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL~-2626, p 32.

UNCLASSIFIED
PHOTO 32923

FAI LUR s

-/~

COOK ELECTRIC CO.
INCONEL

FLEXONICS CORP
STAINLESS STEEL
FLEXONICS CORP
INCONEL

Fig. 1.2.9. Bellows Removed from Expansion Joints Tested in a Molten Salt or Sodium at 1300°F Showing Location of Fallure
UNCL ASSIFIED
PHOTO 32921

Fig. 1.2.10. Cracked Area of Flexonics Corp. Inconel Expansion Joint Bellows That Was Tested in a Molten Salt ot 1300°F.

—59-

DESIGN, CONSTRUCTION, AND OPERATION OF MATERIALS TESTING LOOPS

Forced-Circulation Loops. The operation of forced-circulation

corrosion-testing loops was continued. Three new loops were started and
one loop was terminated, on schedule, during the quarter. By the end

of October, the 15 forced-circulation loop test stands were operating
at full capacity. Eleven of the facilities occupying five column bays
of Bldg. 9201-3 may be seen in Fig. 1.2.11. Thé remaining four loops
are located to the right and across the aisle from the area shown,

Two of the three new loops started during the quarter were constructed
of INOR-8 and the other was constructed of Inconel. One INOR-8 loop and
one Inconel loop (designated MSRP-12 and 9377~5, respectively, in the
operations summary, Table 1.2.3) were equipped with a molten-salt
sampling device,lT which makes possible the removal of samples during
operation. The results of analyses of the 57 samples that have been
removed from these two loops are presented in Chap. 2.2 of this report.

The loop terminated was the Inconel loop listed as 93hhi~l in
Table 1.2,3, which had operated a full year with a temperature difference
across the loop. The facility from which this loop was removed was
modified according to the new design18 for improved operation before
loop 9377~5 was installed. Of the 15 facilities in ofieration at the
present time, nine are of the new design and the remaining six are in
various stages of improvement.

In November the entire facility, with 1k operating loops, survived
a mamentary power failure without a freeze-up of the salts. All loops

regained flow as soon as power was restored. Those loops which had been

Y5, 1. Crowley, A Sampling Device for Molten-Salt Systems, ORNL-2688
(to be issued).

18J. L. Crowley, MSR Quar. Prog. Rep. June 30, 1958, ORNL-2551, p 36.

Fig. 1.2.11. View of Operating Area for Forced-Circulation Corrosion-Testing Loops.

-09-

1 . v v '] *
Table 1,2,3. Forced-Circulation Loop Operations Summary as of December 31, 1958
Composition A ] A . Maximum Temperature Hours of
Loop L M ol and S; Number of plprox;Qmute Ppmx';:me Wall Difference Operation at o
Designation oop Material and Size Circulated Flow Rate Reynolds Temperature Across Loop  Conditions Special Features Comments
Fluid® (gpm} Number (°F) (°F) Given
9344-1 Inconel, ]/2 in, OD, 123 2 3250 1300 200 8800 Loop terminated Nov, 5, 1958,
0.045 in. wall after one year of operation
9354-3 INOR-8 84 2,75 1200 100 8374 Contains a Hastelloy B Loop fluid dumped temporarily
Hot leg,“';/B in, sched 40, 4500 pump on Oct. 25, 1958, when small
Cold leg, ]/2 in. OD, 5400 oil fire broke out near pump
0.045 in, wall
9344.2 Inconel, 1/2 in. OD, 0.045 12 2.5 8200 1200 200 8067 Normal operation
in, wall
9377-3 Inconel, ‘/2 in, OD, 0.045 131 2 3400 1300 200 6980 Normal operation
in. wall
9354-1 INOR-8, ]/2 in, OD, 0.045 126 2.5 2000 1300 200 6503 Normal operation
in, wall
9354-5 {NOR-8, 3/8 in. OD, 0,035 130 1 2200 1300 200 5590 Contains box of graphite Normal cperation
in, wall rods in hot fluid
sire-cu'nh
9354-4 INOR-8 130 2.5 1300 200 3852 Contains three machined Normal operation
Hot leg,s/a in. sched 40, 3000 sample inserts in re-
Cold leg, ‘/2 in, OD, 3500 sistance-heated
0.045 in, watl section®
MSRP-7 INOR-8, ]/2 in. OD, 0.045 133 Not available 1300 190 3418 This loop and all others  Loop fluid dumped Oct. 13,
in. wall listed below contain 1958, when pump stopped
redesigned safety after loss of power to both
features® magnetic clutchs
9377-4 Inconel, 1/2 in. OD, 0.045 130 1.75 2600 1300 200 3408 LFB pump rotary element,
in. wall changed Dec. 9, 1958, be-
cause of bad bearing
MSRP-6 INOR-8, ]/2 in. OD, 0.045 134 Not available 1300 200 3028 Normal operation
in, wall
MSRP.8 INOR-8, 1/2 in, OD, 0.045 124 2.0 4000 1300 200 280¢% Normal operation

in. wall

L9
Table 1.2.3 {continued)

Composition A ) A ) Maximum Temperoture Hours of
Loop L M ‘ol and Si Number of :Iprommote pproxu;r;ote Wall Ditference Operotion at )
Designotion oop Materiol and Size Circulated ow Rote ReYn: s Temperature Across Loop  Conditions Special Features Comments
Fluid® (gpm) Number (°F) (°F) Given

MSRP-9 INOR-8, ]/2 in, OD, 0.045 134 Not available 1300 190 2636 Normal operation
in., woll

MSRP-10 INOR-8, l/2 in. OD, 0,045 135 Not available 1300 200 2469 Normal operation
in, wall

MSRP-11 INOR-8, ‘/2 in. OD, 0.045 123 2.0 3200 1300 190 2128 Started Oct, 3, 1958; normal
in. wall operation

MSRP-12 INOR-8, ]/2 in. OD, 0.045 134 Not ovailable 1300 200 1521 Contains o molten-solt Storted Oct. 29, 1958; normal
in. wall sampling device operation

9377-5 Inconel, 1/2 in. OD, 0,045 134 Mot available 1300 190 873 Contains a molten-salt Started Nov. 25, 1958; narmal
in. wall sampling device operation

z9

“Composition 12:  NaF-KF-LiF (11.5-42-46.5 mole %)
Composition 84: NaF-LiF-Be F2 (27-35-38 mole %)
Composition 123: NaF-BeF -UF‘1 {53-46-1 mole %)
Composition 124: NaF-BeF —ThF4 (58-35-7 mole %)
Composition 126: LiF-BeF --UF‘1 (53-46-1 mole %)
Composition 130: LiF-BeF ‘UFA (62-37-1 mole %)
Composition 131: LiF-BeF -UF4 (60+36-4 mole %)
Composition 133: LiF-BeF -ThF4 (71-16-13 mole %)
Composition 134: LiF-BeF 'ThFd'UFd (62-36.5-1-0.5 mole %)
Composition 135: NaF-Ber-ThFd-UF4 (53-45.5-1-0.5 mole %}

I:’J. L. Crowley, MSR Quar, Prog, Rep. Jan. 31, 1958, ORNL-2474, p 31.

8]

NN R

€J. L. Crowley, MSR Quar, Prog. Rep. June 30, 1958, ORNL-2551, p 36, Fig. 1.2.16.

- 63 -

provided with the necessary equipment assumed isothermal operation
automatically, and steps were taken to prevent freeze-ups of the
remaining semiprotected loops. All loops were returned to test
conditions with no adverse effects and with a minimum of lost time.

The preventive maintenance program on the corrosion-testing
facilities requires that the loops be operated isothermally when it is
necessary to repair or replace worn or defective equipment. Such events
are listed in Table 1.2.3 under comments only if it was necessary to
drain the operating fluid into the system sump or to change the operating
fluid when a system was opened.

In-Pile Loops. Operation of the first in-pile loop was started in

the MTR December 3, 1958. The loop operated satisfactorily at design
conditions until Januvary 8, 1959 (nearly two MIR cycles) when a high
radiation‘field accompanied by fission-gas release developed in the
vicinity of the cubicle containing the loop. Loop operation was
terminated. The radiation release has been traced to a partially
plugged pump-sump purge-outlet line which caused fission gases to back-
diffuse up the pump-sump purge-inlet line to a point outside the cubicle,
where they escaped ithrough a leak. The loop had accumuldted a total of
860 hr of operation, of which approximately 700 hr were with the reactor
at power.

The operation of the second prototype in-pile pump was terminated
after 2500 hr of satisfactory operation at the design conditions of the
first loop. Examination showed the pump to be in the as-installed
condition, except for an accumulation of salt in the intermediate region
between the pump sump and bearing=housing region. The accumulation had
been obgerved by X-ray examinmation during operation of the pump, but it
was considered to be of insufficient quantity to cause difficulty.

The third prototype in-pile pump, which is still being tested, has
accumnulated over 2000 hr of satisfagtory operation. This pump is identical

to that installed in the second in-pile loop and incorporates means to
- 64 .

eliminate salt accumulation in the intermediate region. X~ray
examinations of the operating pump have verified the effectiveness of
the modification.

The assembly of the second in~pile loop is approximately 60%
complete. This loop will incorporate modifications to the purge system
which should prevent a recurrence of the difficulty encountered with

the first loop.

- 65 .

1.3. ENGINEERING RESEARCH

PHYSICAL PROPERTY MEASUREMENTS

Viscogity. Viscometers of the gravity-flow capillary-efflux type;

which have been employed almost routinely in determinations of the
viscosities of molten-salt mixturesyl have recently given unexpectedly
irregular results2 for beryllium-containing fluoride mixtures. Efforts
to increase the experimental precision of the data have led to modifications
both in the efflux-cup design and in the cup calibration technique.

The design of the modified efflux-cup viscometer is illustrated in
Fig. 1.3.1. It differs from previous designs in two significant respects:
(1) the length of the cup was increased to provide either 120 or 240%
more fluid volume, and (2) a serrated "skirt" was added to the lower
outside portion of the cup. The first of these changes allowed longer
flow times and diminished the effect of initial surface "humping," and the
gecond change assured that drainage along the outside surface did not
interfere with the discharge from the capillary tube.

The new cups were calibrated with a NaNOE-NaN03-KNO (LO=T7=53 wt %)
mixture of known v1scosity.3 In the temperature range 150 to 400° C,
this salt has a kinematic viscosity that is in the same range as
the kinematic viscosities of the fluoride salts at 500 %o 800°¢C.
Comparison of the NaNO.-NaNO_-KNO_ calibration with room-temperature

2 3 3
calibrations with glycerine-water solutions of similar kinematic viscogities

lS, I. Cohen and T. N. Jones, Viscosity Measurements on Molten Fluoride

Mixtures, ORNL-2278, (June 28, 1957}.
W. D. Powers, MSR Quar. Prog. Rep. June 30, 1958, ORNL-2551, p 38.

3W. E. Kirst, W. M. Nagle, and J. B. Castner, Trans. Am. Inst. Chem.
Engrs. 36, 371 (1940).

266 .

UNCLASSIFIED
ORNL-LR-DWG 35677

CUP, 1.0 in.0D, 0.881 in. ID
% OR 2% in. LONG.

T

i
M,

i
[ ppates

WEEP HOLE — |

"SKIRT" EIGHT EQUALLY

CAPILLARY TUBE, 0.039 OR
SPACED POINTS

0.046 in. ID, 1.0 in. LONG

Fig. 1.3.1. Modifled Efflux Cup Viscometer.

- 67 -

indicated a significant temperature dependence. The room-temperature
calibration leads to high and inconsistent results for the fluoride
salts.

New viscosity values were obtained for the mixtures LiF-BeFE-UFh
(62-37-1 mole %) and LiF-BeFe--UFrThFh (62-36.5-0.5-1 mole %} with the
modified and recalibrated efflux cups, and comparison of the new values
with the data obtained previously confirmed the calibration effects
discugsed above. Over the temperature range of 500 to SOOOC, the new
viscosity values for the two mixtures may be closely correlated by the

equation
Ho=h eB/T ,

where‘/I is the viscosity in centipoise, T is the temperature in OK, and
A and B are experimentally determined constants. The values obtained

for A and B for LiF-BeF,-UF) (62-37-1 mole %) are A = 0.0580, B = L4550,
and for LiF-BeF,-UF) -ThF) (62-36.5-0.5-1 mole %) are A = 0.0680, B = L49T.
The new viscosity values for the mixture LiF-BeF,-UF) (62-37-1 mole %)
are presented in Fig. 1.3.2, along with the values obtained previously

for comparison.
Enthalpy and Heat Capacity. Measurements of the enthalpies in the

solid state of three mixtures of lithium chloride and potassium chloride
were completed. The results are given in Table 1l.3.1, along with the

previously reported values for these mixtures in the liquid phase.

MOLTEN-SALT HEAT TRANSFER STUDIES

The "continuous-operation" apparatus designed for studying surface
film formation in molten-salt systems by heat transfer coefficient
measurements is indicated schematically in Fig. 1.3.3. The mixing chambers
nave been redesigned so that the thermocouples can be placed along the

centerline of a chamber to obtain better fluid mixed-mean temperatures.
H, VISCOSITY (centipoise)

N
AN ' ’
\ o) | j —
A |
Q
C,
20 \\0 ;
\ 41 =0.0580 e%5%/7
‘\A\/o/-
@]
. o NOTE: IN THE VISCOSITY EQUATION,
a L 7 1S IN °K
10 | o &
. \"; Py
N (] e .J
l \ . %e
6 —— SYMBOL  TYPE OF CUP CALIBRATION MEDIUM —| & o
A \tr o @ ® e
A WITH SKIRT NaNO, - NaNOz = KNO5 D . 2N
A WITH SKIRT NoNO, - NaNOz ~KNO, - -
P WITH SKIRT NGNO,~NaNO5 ~KNO5 o
o WITHOUT SKIRT GLYCERINE - WATER SOLUTION o~
° WITHOUT SKIRT 'GLYCERINE - WATER SOLUTION _
2 R -
1
400 450 500 550 600 650 700 750 800

-68-

UNCLASSIFIED

ORNL-LR-DWG 35678

TEMPERATURE (°C)

Fig. 1.3.2. Viscosity of Molten Salt Mixture LiF-BeF,-UF , (62-37-1 mole %).

B50O

Table 1.3.1. Enthalpy Equation Coefficients and Heats of Fusion of Several LiCl=-KC1l Mixtures

Salt Mixture Phase Temperature Enthalpy Equation Coefficients* Heat of Fusion (cal/g)
Range (°C) a b c Temperature H - Hg
- (c)
x 1077
LiCl-KC1l Solid 100-350 -6.8 +0.212 +6.35 L43 T0
(70-30 molie %) Liquid 450-800 +15.3 +0.361 -3.12
LiCl-KCl Solid 100-350 -7.1 +0.204 +6.26 370 61.
(60-40 mole %) Liquid 400-800 +21.2 +0.315 -0,18
LiCl-KC1 Solid 100-350 -5.9 +0.192 +5.40 450 6l
(50=50 mole %) Liquid 500800 +38.5 +0.239 +1 .62

*
The enthalpy is given as HT - H3O°C = a + bT+—cT2; the heat capacity is then evaluated as cp = b + 2cT.

-69-
.70-

UNCLASSIFIED
ORNL-LR-DWG 35679

FLOWMETER

ELECTRODES

: \ HEAT SINK

TEST

SECTIONS MIXING CHAMBER (LOCATED IN PIPE

PUMP AT EACH END OF TEST SECTION)

Fig. 1.3.3. System for Heat Transfer Coefficient Measurements of BeF,-Containing Molten Salts.

- 71 =

Further, the thermally indeterminate regions between the test-section
electrodes and the mixing chambers have been eliminated by incdrporating
the electrodes as the end plates of the mixing chambers, The components
for this system are being fabricated and assembled. |
Initial calibration of the turbine-type flowmeter to be used in the
apparatus has been completed at room temperature with water as the
éalibrating medium. No deterioration of the calibration with time was
observed. The essential components of the apparatus are being mocked
up with the use of Tygon-coupled sections in order to determine both the
pump speed vs flow rate characterigtics and the over-all system pressure

Arop.
Part 2. Materials Studies
- 75 «

2.1. METALLUGRY

DYNAMIC CORROSION STUDIES

Thermal-Convection Loop Tests. Examinations of two Inconel thermal-

convection loops that operated with fused fluoride mixtures for one year
at 13500F vere completed during the quarter. One of the loops (1181) which
circulated a blanket salt of the composition LiF-ThFh (71-29 mole %),
showed attack in the form of subsurface voids to a depth of 6.5 mils along
hot-leg sections. No attack or deposits were observed in cold-leg sections.
The other loop (1184) which circulated the fuel mixture NaF--Zth-UFh
(55.3=40.7=4 mole %), was similarly attacked to a depth of 7 mils and
showed no attack or deposits in cold-leg sections.

A third Inconel loop (1223) was examined after completing 1000 hr
of operation at 1350°F with the fuel mixture LiF-BeF,-UF . Attack in
thig loop occurred as light intergranular void formation to a maximum
depth of 1 mil. The fuel mixture circulated in this loop was obtained
from the same batch as a mixture circulated in loop 1222, described
previously.l However, prior to operatiocn in loop 1223, an analysis of the
fuel was made to detect sulfur contamination. In the course of this
analysis, the fuel, while molten, was purged extensively with HF and
hydrogen gas. No measurable sulfur contamination was found, and, in
general, the impurity concentrations appeared to be unaffected by this
treatment. However, the as-received fuel caused heavy attack to a depth
of 3 mils in loop 1222 under conditions otherwise identical to those of

loop 1223. An effect of the purging treatment on the corrosiveness of

1
P 53.

J. H. DeVan et al., MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL-2626,

- 76 -

the fuel is therefore evident, but the mechanism through which this effect
occurred has not been identified.

Operation of an Inconel thermal-convection loop with the salt
mixture NaF-LiF-KF (1l.5-46.5=42 mole %) was terminated after 4673 hr in
order to provide additional test space for salts of currently greater
program interest. The maximum operating temperature of the loop was
1250°F. Examination of this loop (121k4) revealed extensive attack along
hot-leg sections to a depth of 13 mils, but the cold-leg sections showed
no evidence of mass transfer. Chemical analysis of the salt after the
test showed a significant increase in the chromium content compared with
that found by analysis prior to the test.

The occurrence of such gross attack in the absence of observable
mass transfer strongly ifidicated that the fuel became contaminated either
prior to or subsequent to introduction into the loop. The high chromium
concentration was unexpected because there was no UFh in the salt. A
subsequent check of the salt by petrographic and x-ray analyses revealed
the presence of approximately 15% KF°2H20. The presence of this
constituent suggests that water, a highly oxidizing impurity in tpg
fluoride systems, was introduced into the salt at some point befafe or
during the test and was responsible for the extensive attack.

Forced-Circulation Loop Tests. The status of the forced-circulation

loops now in operation with fluoride salts is presented in Chap, 1.2; no

tests were completed during the quarter.

GENERAL CORROSION STUDIES

Carburization of Inconel and INOR-8 in Systems Containing Scdium

and Graphite. A system containing sodium and graphite, which forms an

—77-

effective carburizing medium for nickel-base alloys,2 was used to
carburize Inconel and INOR-8 by exposure for 4OOO hr at 1200°F. The
test procedures and apparatus were described previouslyo3 Room-
temperature mechanical tests show that the tensile strength and elongation
of the carburized INOR-8 decreased (~6.5% and 50%, respectively) in
comparison with the tensile strength and elongation of the control
specimens. The control specimens were subjected to an argon atmosphere
for the same time and at the same temperature that the carburized
specimens were exposed to the sodium-graphite system. The tensile and
yield strengths of the carburized Inconel specimens increased (~24% and
9%, respectively), while the elongation decreased, in comparison with the
same properties of Inconel control specimens. Carburization was found
metallographically to a depth of about 7 mils on both metals. Chemical
analyses of the carburized specimens of both TInconel and INOR-8 showed
decreasging carbon conteht with depth as indicated in Table 2.l.1l. The
values obtained from the chemical analyses were used to prepare the
curvesh shown in Fig. 2.1.1. From these curves, a carbon content vs
distance relationship for certain times was determined, as indicated in

Fige. 2.1.2,

2E. E. Hofffian, W. H. Cook, and D, H. Jansen, MSR Quar. Prog, Rep,
Jano 311 1958, 0RNL"’2’-|'7,+) D 5""0

3E. E. Hoffman, W. H. Cook, and D. H. Jansen, MSR Quar. Prog. Rep.
June 30, 1958, ORNL=-2551, p 59. o

“w. H. Cook and D. H. Jansen, A Preliminary Summary of Studies of
INOR-8, Inconel, Graphite, and Fluoride Systems for the MSRP for the Period
from May 1, 1958 to Dec. 31, 1958, ORNL CF=59-l=h (to be published)..

- 78 -

Table 2.,1.1. Carbon Content of Millings Taken from Inconel and %NOR-B
After Exposure to Sodium-Graphite System for LOOO hr at 1200 F

Carbon Content (wt %)

Material .

At 0 =3 mils At 3-6 mils At 6~9 mils At 9=12 mils
Inconel 0.21 0.12 0,055 0.030
INOR=8 0.27 0.10 0.048 0.026

*
Surface of specimen.

A second carburization test was completed for which the test
conditions were identical to those described above, except that the
duration of the exposure was LOO hr, rather than L4000 hr. No
carburization .of the specimens could be detected metallographically after
this test. Mechanical property tests showed no reductions in elongation
of the Inconel or INOR-8 specimens.

Carburization of Inconel and INOR-8 in Systems Containing Fuel 130
and Graphite. Two tests at l300°F in which Inconel'and INOR-8 were
exposed to fuel 130 (LiF-BeFE-UFh, 62-37~1 mole %) in systems containing
graphite were terminated after 2000 and LOOO hr, respectively. The
Inconel specimens were attaéked to depths of 3 and 7 mils, respectively,
in the 2000- and 400O~hr tests, and there were reductions in mechanical
strength as a result of the attack,h The INOR-8 specimens showed no
carburization in either case. Surface roughening, which was found only
on the specimens from the L400O-hr test, was to a depth of less than 0.5 mil.
No significant changes in mechanical properties were foufid for the INOR-8
specimens from either the 2000- or LOOO=hr tests in argon or in fuel
130 exposed to graphite. The mechanical test results gave indications that

INOR~8 was not carburized under the conditions of this test.h

-79-

UNCLASSIFIED
ORNL-LR-DWG 35449R

4.0 T [ I 1
‘ ALLOY O (cm?/sec)  C, (wi%)
| INCONEL 7X10™%2 0.26
os | | _INOR-8 3.5%40° "% 0.40 _
i ‘ .
i |
| | |
by
— 7
AN
G o6 a ‘7’0 9 - - —
b NS
g‘ :_Z‘ —i + X 7,
~ o} B 2 AN
- 3] R < CIESAY?
o V e (g AS -~
O ® m (@] O¢ A q_
o4 2\ A RN\S 2,
e oloz\5 ¢ - 5y,
Pl O\Z @N\'p '9(‘\
i “‘l AR AN
2\e P\ LN\& \
o\~ |
0.2 PO —
NS N
; -\é.. Tl«,_ \ \
¢ 10 20 30 40 50
DISTANCE FROM INNER SURFACE {mils}
C=WEIGHT %, CARBON AT POINT WITHIN TUBE WALL
C;.=WE|GHT %> CARBON AT SOLID—-LIQUID INTERFACE
(INIT1AL TUBE SURFACE)
Co= INITIAL WEIGHT 7. CARBON IN ALLOY
O =DIFFUSION CONSTANT
Fig. 2.1.1. Approximate Carbon Distribution in Walls
of 50-mil Inconel and INOR-8 Tubing Exposed for

Various Times to Liquid Sodium and Graphite at 1300°F
on Inner Surface and Argon on Outer Surface.

UNCLASSIFIED
ORNL-LR-DWG 25120R

040
| W
|
‘ | |
030 } F | T—
88
3 .
Z 020 —— - - }— —
& oy
S ey@Or
i ; 1
J’eors
<9
‘ €arg
. | |
0 10 20 30 40 50

DISTANCE FROM INNER SURFACE (mils)

Fig. 2.1.2. Amount of Carbon (wt %) at Points Within
a 50-mil Thick INOR-8 Tube Wall After Yarious Exposure

Times

Surface and Argon on Outer Surface.

at 1300°F to Liquid Sodium-Graphite on Inner
- 8 -

Uranium Precipitation From Fused Fluoride Salts in Contact with

Graphite, Tests were continued in the study of the compatibility of
p,

graphite and molten-salt fuel mixtures. As described previously,
the test systems are radiographed periodically to check for UO2
precipitation from the fuel mixture and penetration of the fuel into the
graphite.

Uranium oxide precipitation has been found in all teste in which
the fuel mixture 130 has been used in contact with graphite at 1300 h
in these tests the ratio of the calculated projected surface area of the
graphite to the volume of the fuel was 3.6:1. The fraction of the
uranium that precipitates from fuel 130 appears to be finite and
eomplete within a 5-hr exposure to graphite at 13000F under a vacuum of
less than Oal/u. The basis for the conclusion that the precipitation is
finite is that no detectable changes in the initial quantities of
precipitate have'been seen radiographically in any of the test systems,
even after éxposures as long as 3000 hr.

The quantity of uranium precipitated appears to be an indirect
function of the purity of the graphite; that is, the greatest quantities
of precipitated uranium are found in the test systems containing the
most impure graphite. The impurities in the graphite are traces of'
metals, and since the adsorptive qualities of graphite are increased and
degassing is made more difficult by metallic contaminants, it is believed
that the oxygen source for the uranium precipit;tion was adsorbed to a
greater extent on the less pure graphiteo

The graphites used in the test systems in which uranium precipitation
has been found were degassed at 2350 F for 5 hr under a vacuum of less
than h/l, flooded with argon, and then exposed to air for 3 min while they

. H. Cook, MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL-2626, p 61.
-8l -

were being weighed. Apparently the brief exposure to air was sufficient
for the graphite to pick up enough contamination to cause the uranium
precipitation. In contrast to the results being obtained for fuel 130,
no segregation or precipitation has been found radiographically when fuel
30 (NeF-2rF)-UF), 50-L6-4 mole %) is in contact with graphite at 1300°F
under a vacuum of less than 0.14 for 3000 hr or longer.

In the tests under way,.steps are being taken to determine definitely
whethef graphite is directly or indirectly responsible for uraniup
precipitation from fuel 130 under these,con@itions. Since, on the basis
of the preliminary.results,,graphite does appear to be responsible,
attempts are being made to determine whether the precipiiation can be
eliminated by using purer graphite and/or by purging the graphite with gas
or liquid prior to exposure to fuel 130. '

Fused Floride Salt Penetration Into Graphite. Fuel 30 penetration

into graphite pore spaces has been erratic in tests of 1000 hr or less
at 13OOOF under a vacuum of €0.1 Pl The gi'aphite specimens have

2, b The penetration is probably

shown no penetratibn to heavy penetration.
due to contamination of the graphite; reaction products are being sought

to verify this supposition. The effect of pressure on the penetration of-
fuel 30 into graphite pore spaces at 13OOOF is being investigated. The
pressure being used in a test now in progress is 150 psia.

In examinations just started, weight-change measurements of TSF-grade
graphite specimens that had been exposed for 2000 and 4000 hr to fuel 130
at 13OOOF under a vacuum of approximately O.;}( have indicated that fuel
130 did not penetrate the graphite during the 2000-hr exposure. A weight
gein resulting from the LOOO-hr exposure is being investigated to
determine whether there was penetration or whether reaction products were
formed. In two tests of 100-hr duration at 1300°F, fuel 30 under a.pressfire

of 150 psia has not penetrated evacuated (<112/4) graphite.h
- 82 -

MECHANICAL PROPERTIES OF "INOR-8

Fabricability and corrosion resistance to fused salts are perhaps
the most important requiremehts for a structural material for the
molten-gsalt reactor, but the high-temperature'strength of the material
is also important from a design standpoint. The INOR-8 alloy presently
being considered as the structural material for the molten-salt reactor
system was developed principally to possess corrosion resistance; it |
is presently being studied from the standpoint of its strength properties.
Tests are under way in which tensile strength, creep, relaxation, and
fatigue effects are being evaluated. This tésting program has two
objectives: first, to obtain reliable data suitable for use in design
engineering, and,‘second, To explore many of the factors, other than
stress and temperature, fihich may affect the strength properties.

Tensile and creep data obtained in air were presqnted previously6’7
which indicated that the strength of INOR-8 approaches thatrof the
strongest stainless steels. Tests to determine the creep strength of
INOR-8 in molten salts are under fiay. Times to 1% strain as a function
of temperature and stress in short-time tests were presented previously,7
and long-time tests are in progress at 1100 and 1200°F . Specimens
stressed at 25,000 psi at 1100°F have mnot yet reached 1% creep strain.
even after 6000 hr.

The results obtained by ORNL in molten salts were compared with
results obtained by the Haynes Stellite Company in air in the previous
report,7 but differences in the heat treatment of the specimens prevented

any conclusions from being drawn with respect to the relative effects

6b A. DOuglas, MSR Quar. Prog. Rep. June 30, 1958, ORNL=-2551, p 6k4.

7]E{ W. Swindeman and D. A. Douglas, MSR Quar. Prog. Rep. Oct. 31,
958 ORNL—2626, p 64.

- 8% -

of the environments. Subsequently, tests in air were.initiated at ORNL
on material identical to that tested in molten salts. The data thus
obtained in air are compared with the moltenesalt data in Table 2.1l.2.

Table 2.1.2. Comparison of Creep Data for INOR-8 Sheet Specimens
Tested in Fuel Salt 107 and in Air

Temperature Stress FEnvironment Time (hr) to Specified Strain Time to

(°F) (psi) 0.5%  1.0%  2.0% 5.0% Rupture
_ (hr)
1800 3,000 Salt 4.5 17 48 115 130
Air 8 17 35 75
1700 5,000 Salt 4.5  12.5 31 110 250
Air 5.4 12.5 25 58 90
1500 8, 000 Salt 19 52 125 350 886
Air 29 49 86 175 260
1250 20,000 “Salt* 110 265 680 1600 2400
Air 180 320 720 1500 2180

- .
Interpolated data.

On the basis of a 1% creep strain, the data indicate that environment
does nqt‘appear to seriously affect the creep properties. At 1500, 1700,
and lBOOoF, however, the times to greater strains and to rupture are much
longer in molten salts than in air. At 12500F and a stress of 20,000
psi, the molten salt and air data agree fairly we%l at all strain levels.
A correlation of data from tests on sheet and rod specimens in air
is presented in Table 2.l.3. Rod specimens appear to be the strongér'afi
20,000 psi, and the sheet specimens are slightly stronger at 10;000 psi.
- 8k -

Table 2.1.3. Comparison of Creep Data for INOR-8 Sheet and Rod
Specimens Tested in Air at 1250°F

Stress Specimen Time (hr) to Specified Strain Time to
(psi) 0.1% 0.2% 0.5% 1.06 2.0% 5.0% Rupture
- (hr)
20,000  Sheet 30 80 190 350 730 1700 1786
Sheet 30 6L 180 320 720 1500 2177
Rod 20 125 300 510 920 1800 3535
Rod 15 100 300 510 920 2200
15,000  Sheet 190 350 750 1400
Sheet 120 330 780 1500
Rod 130 300 40 1500
12,000 Sheet 600 931 2275
Rod 320 TLO 2082
10,000 Sheet 1650 2600 4270
Rod 760 1600 3596

Tensile data available for INOR-8 show that the ductility decreases
rapidly with increasing temperature between 1000 and lSOOOFo For example,
elongations ranging from 15 to 30% have been found at lhOOOF; whereas,
at 1000°F, the ductility may be as high as 60%. Since low ductilities
are sometimes associated with notch senéitivity, several tensile tests
were conducted in order to establish the notched-tosunnotched strength
ratio. A ratio greater than unity is considered to indicate notch
strengthening, while a ratio less than unity signifies notch sensitivity.
Results from these tests are presented in Table 2.1.4. On the basis of

- 85 -

Table 2.1.4. Effect of Notches on the Tensile Strength and
Ductility of INOR-8

Tensile Reduction Notch
Temperature Geometry Strength in Area Strength
(psi) (%) Ratio
Room Unnotched 116,000 48 1.08
Notched 125,000 20
1500°F Unnotched 148, 500 15 1.37
Notched 66, 500 8

(0.0045<in. radius)

the definition given above, INOR-8 is notch strengthened at room
temperature and at 1500°F. Further studies have been planned in order
to ascertain the effect of surface carburization dh the influence of
notches.

Rotating-beam fatigue studies on INOR-8 are in progress under a
subcontract at the Battelle Memorial Institute. Tests have been programmed
at 1200 and 1500°F. At 1500°F, under an alternating stress of +30,000
psi at 100 rpm, fatigue liveé for two specimens were found to be 730,000
and 930,000 cycles. A comparable life for Inconel tested under similar
conditions would be produced by a stress of approximately 18,000 psi.

MATERIALS FABRICATION STUDIES

Triplex Tubing for Heat Exchangers. The feasibility of fabricating
heat exchanger-tubing that involves two concentric tubes with a porous
metal in the annulus between them through which a gas can be passed for
leak detection is being studied. Initially, two samples, each approximately

~-86--

18 in. long, were prepared by tamping and vibrating Inconel powder into
the 0.125-in. annulus between two Inconel tubes and sintering for 2 hr
at 2250°F in a hydrogen atmosphere;. A density of about 50% of theoretical
wasg achieved in the sintered material in the annuli of the tubes, and
X=ray examination showed the bonding to be unsatisfactory. One sample
was subsequently drawn from 1.000 in. OD to 0.875 in. OD and resintered.
This tube was found by x=-ray and metallographic inspettions to have
‘good bonding; a density of 75% wag achieved in the sintered material
in the annulus. Evaluation tests of these tubes by the Reactbr Projects
Divigion have shown the tube with 50% dense sintered material in the
annulus to have a low permeability for helium and the permeability of
the 75% dense material to be a factor of 100 less.

In order to obtain good thermal conductivity across the annulus,
attempts have been made to fabricate tubes with porous nickel rather
than Inconel in the amnulus. Preliminary studies of the sintering
characterigstics of =325 mesh and =100 +200 mesh nickel powder after
tamping and sintering over the temperature range of 1800 to 2300°F showed
the latter particle-size powder to be the more promising for use in the
core. The finer powder, =325 mesh, was subject to gross shrinkage when
sintered within the indicated temperature range. The first triplex
tube assembled with =100 +200 mesh nickel powder in the annulus was
sintered 1 hr at 2100°F, drawn from 1,000 in. OD to 0.875 in. 0D, and
resintered 1 hr at 2100°F., In order to maintain reducing conditions
within the annulus, the tube was continually purged with hydrogen during
sintering. Upon metallographic examination of the tube, however, oxide
layers were found at the interfaces between the sintered material and
tubing. A second triplex tube assembled in the same manner as the first
was evacuated to 0.1l /, sintered for 1 hr at 2100°F, drawn from 1.000
in. OD to 0,938 in. OD, and resintered for 1 hr at 2100°F. Again it
was found that poor bonds were formed between the tubing and the sintered
material and that oxides were present at the interfaces. The difficulties

- 87 -

experienced in achieving an interface bond have been attributed to
oxygen contamination of the nickel powder.
Prefabricated porous materials available from commercial vendors
are to be used in future attempts to prepare triplex tubing.
High-Temperature Stability of INOR-8. Embrittlement studies of
INOR-8 in the temperature range 1000 and 1L4L00®F are in progress. In
these studies the changes in the tensile properties of annealed sheet

are being determined at various aging times and temperatures. In the

tensile tests completed thus far, specimens aged for 5000 hr have shown
no significant differences from the annealed specimens. Specimens being
aged for 10,000 hr have compiéted 9000 hr of the heat treatment. |
Cormercial Production of INOR-8., The fabrication qualities of
INOR-8 melted by the arc-cast method are being evaluated. The starting
metal was a standard air-melted heat which was forged into electrodes,

subsequently vacuum arc-melted, and again forged into extrusion billets.
The billets are scheduled to be pierced, extruded, and redrawn into
tubing. INOR-8 prepared by vacuum-induction melting produced extrusion
billets which were unsuitable for further processing because of large
center defects. Upon completion of the present study, INOR-8 will
have been produced by three processes, and a basis will have been established
for placing future orders.
Evaluations of Composites of INOR-8 and Type 316 Stainless Steel.
The compatibility of INOR-8 and type 316 stainless steel at high

temperatures is being investigated. The evaluation methbds include

measurements of the diffusion penetration of the alloys into each other,
the identification of reaction layers, and measurements of the mechani-
cal properties. The data accumulated thus far indicate that the alloys
are compatible at temperatures up to 1800°F. Composites of INOR-8 and
type 316 stainless steel therefore appear to be promising for use in

fused salt-to-sodium heat exchangers.
--88 -

WELDING AND BRAZING STUDIES

Heat Exchanger Fabrication. The welding and brazing problems

encountered in constfucting a triplex-tube heat exchanger are being
studied, and suitable fabrication procedures are being developed. The
basic design of such a heat exchanger consists of two tube sheets at
each end, with the outer tube wall of each tube welded to one tube
sheet and the inner tube wall of each tube welded to the other tube
sheet. The space between the two tube sheets can thus be used for

leak detection. A difficult access problem is encountered in the
welding of the outer tube walls to the first header, since conventional
welding torches are too bulky to fit between adjacent tubes.

For experimental welds, an extension was attached to a small
welding toréh and a shielding-gas nozzle and a trailer shield were
manuf&ctfired from pyrex glass. The tungsten electrode was bent at
an angle of 10 deg to permit placement of the electrode tip and, hence,
the cathode spot, over the tube-to-tube sheet interface.

These torch modifications required that the torch move around its
own axis, as well as around the tube axis, with identical periods of
rotation. A sketch showing the rotational movements required to make
the welds is presented in Fig. 2.1.3. A mechanism was designed and
fabricated to permit this complex rotational motion and a rotating
shielding-gas seél and current contact were enclosed in a suitable
assembly.

A mockup heat exchanger was constructed to assist in the develop-
ment of a welding procedure and to provide information and experience
applicable to the construction of an actual unit. Simulated tubes
vere constructed by separating individual stainless steel and Inconel
tubes with wire spacers. The assembly, which was welded and back

brazed successfully, is shown in Fig. 2.1.k.

s UNCLASSIFIED
N v-27712

UNCLASSIFIED
ORNL— LR~ DWG 35680

ROTATION OF ELECTRODE AND SHIELD

ROTATION OF TORCH AROUND TUBE
e —
H AROUND ITS OWN AXIS

— NOZZLE AND TRAILER SHIELD

WELD
— ELECTRODE

/7 INNER TUBE WALL
——— POROUS ANNULUS

OUTER TUBE WALL

-68-

POSITION A

POSITION B

Mockup Heat Exchanger Assembly After

Fig. 2.1.4.
Welding of Inner and Outer Tubes to Headers.

Fig. 2.1.3. Rotational Movements of Special Welding Torch Assembly.
- 90 -

Welding of INOR-8. A summary report8 is being prepared in which the
weldability of three nickel-molybdenum-alloys -- Hastelloy B, Hastelloy W,
and INOR-8 -- is discussed. The report will correlate the information
avellable on the microstructures of weld metal in the as-welded condition,
the microstructures after aging at 1200, 1300, and 1500°F3 the influence
of'aging at these temperatures on the room-temperature hardness, and
the effect of aging at 1200°F on the room- and,elevated;temperature strength
and ductility. | ‘

Studies are continuing in an effort to improve the high-temperature
ductility of INOR-8 weld metal. It was reported previously that all-weld-
metal tensile test specimens of INOR-8 heats from Haynes Stellite Company,
Westinghouse Electric Corporation, and ORNL exhibited essentially the
éame mechanical properties in the as-welded condition. The ductility
at 1300°F and above is considered to be marginal. |

The effects of wvarious heat treatments of the specimens were studied,
and the results, summarized in Tabie 2.1.5, indicate that the ductility

Table 2.1.5. Effects of Heat Treatments on the Mechanical
Propertieg of INOR-8 Weld Metal
)

Tensile Strengtfi Elongation

Treatment (psi) _ ($in 1 in. gage)

' At Room At At At. Room At At

Temperature 1200CF 1500°F Temperature 1200°F 1500°F

As welded 117,000 72,000 53,000 37 17 7
Aged 500 hr =

at 1200°F 115,000 81,000 35 27
Aged 500 hr

at 1500°F 111,000 55,000 29 28
Held 4 hr

at 1800°F 115, 000 54,000 31 43

o

G. M. Slaughter, P. Patriarca, and R. E. Clausing, Welding of Nickel-
Molybdenum Alloys (to be published in the Welding Journal).

- 91 -

at elevated temperatures was substantially increased. A metallographic
examination indicated that considerable carbide spheroidization occurred
as a result of heat treating at 1500°F and above. The carbide spheroi-
dization may be seen in Fig. 2.1.5. The carbides in the as~-welded INOR-8
weld metal appeared to have precipitated extensively along the grain
boundaries, as shown in Fig. 2.1.6. This spheroidization is thought to
be at least partially responsible for the improvement jn ductility at
lSOOOF after aging at 1500°F. The 1800°F treatment resulted in slightly
more spheroidization than is shown in Fig. 2.1.5. The microstructural
changes resulting from aging at 1200 and 1500°F will be studied further
with the aid of the electron microscore.

These results indicated that the marginal elevated-temperature
ductility of INOR-8 weld metal in the as-welded condition might result
from an unfavorable carbide distribution, as well as from the infiuence
of trace elements in the 0.009 to 0.001 wt % range, which are known to
cause "hot shortness'" in nickel-base alloys. Special heats of INOR-8
were therefore prepared, as described below, in order to study each of
the above conditions.

Four different methods have been utilized thus far to deoxidize
and purify the weld filler metal during the casting of the original
ingot. These are: (1) use of Inconel weld wire (Inco. No. 62) as the
source of most of the nickel, chromium, and iron, (2) use of nickel
weld wire (Inco. No. 61) as the source of all the nickel, (3) addition
of an aluminum-titanium-manganese master alloy to purify the melt
(process recommended by Electro Metallurgical Company), and (h) additions
of aluminum, titanium, manganese, silicon, boron, and magnesium to
purify the melt (as recommended by International Nickel Company). The
results obtained to date on as-deposited INOR-8 are shown in Table 2.1.6.

- . Wl % = L S S et
14 Loty (.\k‘« < Xy N . 2.2 ‘) y '.’ UNCVEQ;:;?ED §*
. <l R N rs
s s > =4
LYY Byt O Xy SNIEERSy V-
L TR ¥ AR PRA . ™
I P { ) «\‘ e 1 [ coe/
iy . 2 = N\ " 5 oF . N 003
‘. 3 P SN . .+
P £nd 010
e R 2 >
- L= o« o 2 fa
) LA Tt Y o P \‘ g fore |
- ~ LB o3
S " o AR oL foue
- J‘\-’ Lol & \ . - LILE
' R T o Los|
¢ - » -

SO 4 . R D At 3 Jd P
~af.- < - P or g ¥ ' e
. - - . abin® - . L

3 = yu e &
"““\ b &'n\':' \ S
-
-

Fig. 2.1.5. Carbide Spheroldization Occurring In INOR-8 Weld Metal Aged for 500 hr at 1500°F. 200X. Etchant:

copper regia.

1000X

Fig. 2.1.6. Carbide Precipitation Along Grain Boundaries of As-Welded INOR-8 Weld Metal. 1000X. Etchant:

copper regia.

Table 2.1.6. Effect of Melting Practice on Mechanical Properties of INOR- 8
Weld Metal in AsAWelded Condition

Tensile Strength Elongation
Heat No. - Welding Practice (psi) (% in 1 in. eage)
’ At Room At At At Room At At

Temperature 1200°F 1500°F Temperature 1200°F 1500°F

MP-1 Inconel weld wire used to obtain 113,000 77,000 59,000 Ll 30 15
50% of the nickel, 100% of the
chromium, and 90% of the Fe; 1%
Cb wag also obtained from the
weld-wire addition

MP -2 Master alloy obtained from 108, 000 69,000 57,000 43 22 15
Electro Metallurgical Company
used to add 2% Al + Ti

MP-3 0.5% Mn, 0.10% S8i, 0.2% Al, 0.2% 107,000 67,000 48,000 46 25 13
Ti, 0.025% Mg, and 0.005% B added
as recommended by International
Nickel Company

MP-4 Same as MP-3 except boron
omitted Tests in progress
MP-5 100% of the nickel added as nickel

weld wire Tests in progress

"g6."

- o4 -

As may be seen in Table 2.1.6, all the methods described above for
deoxidizing and purifying the ingot have shown promising results. The
ductility at 15000F has been increasged from an average value of 7% to
13 to lB%. Other modificafions in melting practice will be invesgtigated
in an effort to obtain a ductility at 1500°F of 20%.

In order to determine the influence of carbon content on the duc-
tility, vacuum-induction melts containing 0.00+%, 0.03%, and 0.06%
carbon were made, fabricated into weld wire, and deposited as weld
metal. The results of the all-weld-metal tensile tests of these speci-

mens are summarized in Table 2.1.7. The data show that no increase in

Table 2.1.7. Influence of Carbon Content on As-Welded Mechanical
Properties of INOR-8 Weld Metal

Carbon Tengile Strength Elongation
Heat Content (psi) (% in 1 in. gage)
No. (%) At Room At At At Room At At
0 0
Temperature 1200°F 1500°F Temperature 1200°F 1500 F
0.06 117,000 72,000 53,000 37 17 7
MP-7  0.003 112,000 70,000 52,000 35 16 7
MP=6  0.00+ 76,000 47,000 40,000 9 9 7

high-temperature ductility was attained by decreasing the carbon content
of the INOR-8 weld deposit. Also, when no carbon addition to the ingot
was made, the ductility values at room temperature and at 1200°F were
seriously diminished. These poor ductilities are assumed to result
from a preponderance of grain-boundary films (probably oxides). These
films were of such a magnitude that grain-boundary fissures were numer-
ous. The properties of lower-carbon-content melts should also be

investigated when melted with improved practices of the type discussed

in the preceding paragraphs.
Welding of Dissimilar Metals. The problem of selecting & suitable filler

material for welding dissimilar metals has often been an obstacle in the
fabrication of various components. A dissimilar metal weld is composed
not only of filler metal, but substantial quantities of both of the base
materials being joined. The filler metal should thus have the following
properties: (1) high tolerance for dilution by elements such as iron,
nickel, chromium, and copper without forming brittle or crack-sensitive
alloys, (2) a moderate coefficient of thermal expansion, (3) ability to
withstand high service temperatures over long periods of time without
suffering from harmful effects such as sigma-phase formation and carbon
migration, (4) good oxidation and corrosion resistance, and (5) good
gstrength and ductility over the temperature ranges of interest.

The International Nickel Company has developed such a filler wire
for inert-arc welding of dissimilar metals, and the mechanical properties
have been studied and reported.9 The high titanium content of the wire
makes it highly susceptible to age hardening at the molten-salt reactor
service temperature of lQOOOF, and its application may, therefore, be
limited.

A niobium-containing coated electrode, designated Inco Weld "A"
electrode, was also developed for the metallic-arc welding of dissimilar
metals, and the mechanical properties were determined in both the as-
welded and as-aged conditions. The results of this study are shown
in Table 2,1.8. As would be expected from the composition,; the metallic-
arc weld deposits are not subject-to appreciable aging. The tensile
strength at room and elevated temperatures is not appreciably altered,
and the ductility is not decreased. It thus appears that this electrode
might be useful for Jjoining dissimilar metals for high-temperature

9%. M. Slaughter, MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL-2626,
p T0.

- 96 -

Table 2.1.8. Mechanical Properties of Inco-Weld "A" Electrode
in As-Welded and As-Aged Conditions

Test Temperature Tensile Strength Elongation
(°F) (psi) (9 in 1 in. gage)

Room 93, LoO L1
1200 68, 500 28
1300 61,700 19
1500 37,000 32
Room (after aging 500 hr

at 1200°F) 97, 400 38
1200 (after aging 500 hr

at 1200°F) 65,000 W2
Room (after aging 500 hr

at 1500°F) 92,600 40
1500 {after aging 500 hr

at “1500°F) 32, 300 49

service where the metallic-arc process is permissible.

Pump Component Fabrication. A metallographic study of numerous

molybdenum~to-Inconel brazed Jjoints has been conducted in connection
with the fabrication of a pump-shaft extension,9 Molybdenum sheets
1/8 in. thick and of varying sizes were brazed to 1/2-in.-thick
Inconel plates with Coast Metals No. 52 alloy, copper, and an 82 wt %
Au-18 wt % Ni alloy. Examinations of the samples were conducted
visually and metallographically after brazing and after thermally
cycling 15 times from 1200°F.

It was found that extreme stresses are built up in the molyhbdenum-
to~-Inconel brazed Jjoints from differential thermal expansion (molybdenum,
2.9 x 1070 in./in./°F; Inconel, 8.7 x 1070 in./in./°F). The stresses

may cause severe cracking alpng the center of the brazed joint, along

...97..

the brazing alloy-molybdenum interface, or in the molybdenum base
material. As would be expected, the cracking tendencies are decreased
as the area of contact decreases.

Joints brazed with copper exhibited no cracks after brazing, but:
numerous transverse and longitudinal cracks in the Jjoint were noted on
all specimens which were cycled 15 times from 1200°F td room temper-
ature. It appears that the copper lacks adequate strength and ductility
at the elevated temperatures to accommodate the stresses bullt up in
thermal cycling.

The stronger,lless ductile alloy, Au-Ni, was characterized by
longitudinal cracks along the Jjoint. In many cases, the molybdenum
base metal near the joint was also cracked.

 Joints brazed with Coast Metals No. 52 alloy, a strong, brittle
material, showed varied properties. In most samples, only slight
cracking was observed after thermal cycling, while severe cracking
was noted in one sample. The fillets were subject to cracking in most
cases.

Since Coast Metals No. 52 alloy appeared to be the most promising
of the three brazing alloys tested for this application, a prototype
pump shaft and extension were constructed. The conditions of the
test unit will be limited, however, since some of the molybdenum fingers

were cracked during machining.
2.2. CHEMISTRY AND RADIATION DAMAGE

PHASE EQUILIBRIUM STUDIES

Systems Containing UF) and/or ThF). Phase studies are being conducted

to deFermine whetper the N’aF-BeFE-ThFh-UFh system has any advantages over
the LiF-BeFe-ThFh-UFh system as a fuel for a graphite-moderated one-region
reactor. Investigations of the phase relationships in the system NaF-
BeFE-ThF,4 have shown that they are grossly similar to those derived for
the system NaF-BeFE---UFh by the Mound Laboratory,l The choice of NaFaBngm
ThFh mixtures for use in nuclear reactor systems is limited to compositions
having liquidus temperatures of 550°C or lower, and thus the ThFh concens
tration is restricted to the range 10 to 15 mole %. All observations of
the effect of UFh substitution for ThFh in solid solutions (LiFeThFhm
UFEQLiF-Ber-ThFh-UFh, NaF-ThFhJUFh) have indicated that UF)4 substi-
tution causes the liquidus temperature of the resultant solid solution

to be lowered. It would appear, therefore, that the liquid temperatures
encountered in the system Na.F-BeFE--ThFl+ represent maximum temperatures

to be encountered for the liquidus as UFh is substituted for part of

the ThF, in NaF-BeF -ThFh mixtures.

L 2

The System LiF—PuF3. Chemical analyses of filtrates from saturated
solutions of PuF3 in LiF—BeF2 and N‘aF-BeF2

atures and compositions have shown that Put o

melts at a variety of temper-
3 is sufficiently soluble
in such melts to form a fuel mixture for a high-temperature plutonium-

burning reactor.2’3 A sensitive apparatus capable of detecting thermal

1C. J. Barton et al., MSR Quar. Prog. Rep. Oct. 31, 1957,

ORNL-2431, p 26, Fig. 2,22,

2Co J. Barton, W. R, Grimes, and R. A. Strehlow, Solubility and
Stability of PuF3 in Fused Alkali Fluoride-Beryllium Fluoride Mixtures,

ORNL-2530 (June 11, 1958).

3C. J. Barton and R. A. Strehlow, MSR Quar. Prog. Rep. Oct, 31,
1958, ORNL-2626, p 80.

..99..

effects on cooling curves from l-g samples has been developed to study
phase relationships in these and similar systems. Thermal analyses of
LiF-PuF3 mixtures containing 5 to 38 mole % PuF3 show a single eutectic
at 19.5 mole % PuF., which melts at 743 + 2°C. Examinations of slowly-
cooled mixtures by means of a polarizing microscope mounted in a glove
box aided in the location of the eutectic composition and showed no
evidence of compounds or solid solutions in this system.

The System KF-LiF-BeFe. The possibility of using low-melting
point mixtures in the system KF-LiF-BeF2 ag coolants for nuclear

reactors was studied. S8Such mixtures would be attractive because of

the lower gamma activity they would have in comparison with similar
NaF-base mixtures after irradiation in a reactor.

Examinations of 12 melts from thermal analysis experiments by
the use of optical and x-ray techniques showed that the system KF-LiF-
BeF, contains two termary compounds which have the probable formulas

2
KF+LiF*BeF. and KF.LiF-2BeF.. Phases present in the cooled melts

indicate tiat the system isedivided into the following assignable
compatibility triangles: KF-LiF-3KF*BeF,, 3KF*BeF,-LiF-2KF-BeF,,
2KF'BeF2-LiF-KF'LiF‘BeFE, KF-LiF-BeFecLiF-aLiF-BeFe, and KF-LiF-BeFQ—
2LiF-BeF2~KF-LiFe2BeF2v The first two of the triangles will each contain
a eutectic. Neither of these eutectics will have a melting temperature
lower than h7500u Mixtures displaying liquidus temperatures as low as
450°C will probably contain as little as 37 mole % BeF,. Mixtures

having liquidus temperatures of 100°C or lower will probably not be

found that will contain less than 40 mole % BeF,.

Compatibility of Fuel with Chloride Coolants. Some consideration
has been given to use of a mixture of LiCl and RbCl (58.3 mole % LiCl)
as an unreactive coolant for a molten fluoride reactor. Some of the
consequences of a leak between the fuel (69.5 mole % LiF, 29;5 mole %

BeF,, 1 mole % UFh) and this coolant mixture have been examined

2,
experimentally.

- 100 -

The vapor pressure at 900 to 988°C of an equimolar mixture of fuel
and coolant is intermediate between the vapor pressures expected for
the ummixed fluids. The material distilled from the mixture consists
of fluorides and chlorides of Li, Rb, and Be and about 0.2 at. % U,

It is obvious that no compounds of low boiling point result.

Mixtures of coolant and fuel contained so little uranium that
examinations of quenched specimens did not demonstrate whether a uranium
compound was the primary phase. Quenching of mixtures of this coolant
with an LiF-BeFe-UFh (67-28-5 mole %) fuel mixture containing more
uranium showed that Li2Bth is the primary phase from pure fuel to
below 50 mole % fuel in the mixture. The liquidus temperatures dropped
from 440°C for the pure fuel to below 360°C for a mixture with 25 mole
% fuel. ‘It is obvious that no uranium compound would deposit as the

primary phase at reactor temperatures.

FISSION-PRODUCT BEHAVIOR

Effect of UFh on Solubility of CeF3 in LiF-BeF2 Solvents. The
solubility of CeF3 was determined in LiF-BeF2 (62-38 mole %) containing
no UFh' Two additions of UFh were then made to this solvent to give

mixtures containing 1.2 and 1.9 mole % UFh’ respectively, and the

solubility of CeF, was determined in each mixture at several tempera-

3

tures. No effect of UFh on the solubility of CelF., was observed.

3

Removal of Traces of SmF. by the Addition of CeF

3 3
An experiment was completed to determine whether the method used to

f i - F - 20 bl 6. hid
5 from LiF-BeF,, UFh (62.8-36.4

0.8 mole %) would be satisfactory for the removal of trace amounts.

remove relatively large amounts of SmF

It was also desired to determine whether the concentration of SmF

3
remaining in the liquid could be calculated a priori from a knowledge

of the individual solubilities of SmF3 and of CeF3 and a material

- 101 -

balance of the system. The method of calculation was based on relation-

ships described ];;reviously.)4 The results are summarized in Table 2.2.1.

Table 2.2.1. Removal of Traces of SmF3 from LiF-BeF -UFh

2
(62.8-36.4-0.8 mole %) by the Addition of CeF3

SmF3 added: 578 ppm as Sm

SmF,_ in Filtrate (as ppm Sm)

CeF3 Added Temperature 3
(vt %) (Oq) Calculated Observed
2.1 487 342 394
10.1 736 480 480
10.1 587 218 288
10.1 492 95 190

The tabulated results show that the method is effective in

removing traces of SmF., from this liquid. Agreement between the

3
calculated and observed concentrations is considered good in view
of the extensive extrapolations involved when dealing with trace
quantities.

Chemical Reactions of Oxides With Fluorides. The behavior of

oxides in molten fluorides is being studied as part of an effort

to explore chemical reactions which can be adapted to the reprocessing

hW. T. Ward, R. A. Strehlow, W, R. Grimes, and G. M. Watson,
Solubility Relations Among Some Fission Product Fluorides in NaF-ZrF)-

UFy (50-56-5 mole %), ORNL-2L2L (Jan. 15, 19587.

- 102 -

of molten-fluoride-salt reactor fuels. Attempts to precipitate U0,
from solutions of UF) (6.5 wt % U) in LiF-BeF,, (63-37 mole %) by
treatment with excess BeO have shown that the UFh content of the melt
decreases rapidly at first but that the precipitation rate becomes
very slow after about 30 min; less than 40% of the UFh precipitates
in 3 hr. It is likely that the BeO becomes coated with U02 and that
the reaction is effectively prevented. The surface area of the BeO
should, accordingly, prove to be an important variable.

Sparging such an incompletely reacted system with HF in an
attempt to redissolve the precipitated UO, appeared to accelerate

2

the precipitation of U0 In a typical experiment the UFh content

of the melt was reducedgfrom 6.5% to 4.4% U by 3 hr of contact with
excess BeO at 600°C and was further reduced to 2.9% U by 3 hr of
sparging with HF at this temperature. A possible explanation for

this behavior is that the HF is capable of exposing additional surface
area of the BeO pellets previously coated with insoluble UO2 and
thereby accelerating the reaction. An alternate and equally likely
explanation is that HF serves as an intermediate reactant which

reacts with BeO to produce H20 and bring about precipitation of UO2°

CHEMISTRY OF THE CORROSION PROCESS

Activity Coefficients of CrF2 in NaF~Zth° The activity coef-

ficlents of the fluorides of structural metals such as chromium,

nickel, and iron dissolved in dilute solutions of molten fluorides

are of interest to the understanding and prediction of corrosion
reactions taking place in systems in which molten fluorides are in
contact with alloys of these metals. The activity coefficients of

FeEe and NiF2 dissolved in molten NaF-Zth (53-L47 mole %) were determined
- 103 -

>

and reported previously.” The activity coefficients of CrF, in the

seme solvent were determined at 850°C (ref 6) and at 750°C %ref ),
and the activity coefficients determined at 800°C are reported here.
The study of the activity coefficienfis of CrF2 in this solvent, as
originally planned, was concluded with these measurements.

The average equilibrium quotients and activity coefficients for
CrF, in NaF-ZrF) (53-47 mole %) at 850, 800, and 750°C are summarized
in Table 2.2.2. The equilibrium constants listed for reaction 2
in Table 2.2.2 represent simply the arithmetic averages of the
experimentally determined equilibrium guotients. Since no significant
effect was noted #ith changes of CrF2 concentration, thg arithmetic
aversges may be consldered as equilibrium constants obtain€d by
extrapolation to ififinite dilution.

Chromium;Diffusion in Alloys. For the ultimate understanding

of the corrosion mechanism and prediction of corrosion rates occurring

in systems containing molten fluoride solutions in contact with
chromium~bearing alloys, it is necessary to determine the various
factors that affect the diffusion rates of chromium in the alloys.
As a first step of a systematic study of the fundamental corrosion
processes, technidues have been developed aqd describeds to determine
self-diffusion coefficients of chromium in chromium-nickel alloys
with Cr51
As was reported previously,8 the magnitudes of the diffusion

ag a radiotracer.

coefficients obtained from experiments conducted with specimens
which had been polished but not hydrogen fired were one to two orders

5C° M. Blood, W. R. Grimes, and G. M. Watson, Activity Coefficients
of Ferrous Fluoride and of Nickel Fluoride in Molten Sodium Fluoride—
Zirconium Fluoride Solutions, paper No. 75 Division of Physical and
Inorganic Chemistry, 132nd Meeting of the American Chemical Society,
New York, Sept. 8=12, 1957.

_6C° M. Blood, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-24Thk, p 105..
7. M. Blood, MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL-2626, p 96.

8R° B. Evans, R. J. Sheil, and W. M. Johnson, MSR Quar. Prog. Rep.
Oct. 31, 1958, ORNL-2626, p 99. '

- 104 -

Table 2.2.2. Equilibrium Constants for the Reduction of CrF2
by Hydrogen Gas and Activity Coefficients of .
- CrF, in NaF=Zth (53=47 mole %)

Reactions:® (1) CrF, (1) + H, (g)=cCr (s) + 20F (g)
(2) CrF, (d) + H, (8)=>Cr (s) + 28F (g)
Heat of fusion of CrFE:b 5500 cal/mole

Melting point of Cr'F2:b 1375°K

. Activity
Equilibrium Constant, Kfi Coefficients
Temperature - m
(°c) Reaction 1 Reaction 2 y (1) y (d)
850 8.0 x 1073 (1.45 + 0.17) x 1073 0,18 1.0 '
800 2.8 x 1073 (6.77 + 1.12) x 1074 0.2l 1.0
750 8.6 x 107" (2.21 + 0.52) x 1074 0.26 1.0 )

a'Su.bscrip‘t.s s, & 1, and d refer to solid, gaseous, supercooled liquid
and dissolved states, respectively.

bL. Brewer et al., Natl. Nuclear Energy Ser. Div, IV (1950).

Ccalculated equilibrium constants obtained from thermodynamic properties
listed by Brewer et al. (footnote b).

of magnitude higher than those obtained with specimens which had been

polished and then hydrogen fired at 1100 to 1200°C for 2 hr. Photo-

micrographs of the fired and unfired specimens showed that considerabie

grain=growth had occurred in the fired specimens. The differences

in the alloy structures of the annealed and unannealed specimens may

be seen in Fig. 2.2.1. .
From the experiments during this quarter with specimens annealed

in helium, diffusion coefficients were obtained that were very

similar to those obtained with hydrogen-fired specimens. The numerical

Fig. 2.2.1.
2 hr ot 1100 to

-105-

UNCLASSIFIED
T-16052

UNCLASSIFIED
T-15972

e ‘L ois
oA
R g g R
_,’ x
778 (1Led
\ |/ |r@
Y X <
Py

Effect of Annealing on Grain Size of Inconel. (a) Annealed for

1200°C. (b) Unannealed.

magnitudes of the over-all diffusion coefficients obtained by contacting
the specimens with NaF-ZrF (53-L47 mole %) at 675°C were 1.3 x 10 =15

-15 cm /sec for the

cm /sec for the annealed spe01men and 63 x 10
unannesled specimen.

The results of these experiments have shown that the grain
structure has a marked influence on the self-=diffusion rates of
chromium in Inconel. Grain size appears to be a controlling factor.
Hydrogen firing and annealing in helium had the samé effects on the
over=all 4iffusion rate.

Sampling of Operating Loops. Two forced=circulation loops for

corrosion testing of materials in circulating fluoride salts were
equipped during the quarter with devices which provide for frequent
sampling of the circulating salt during loop operation. One loop
equipped with a sampling device was fabricated from INOR-8 and the
other from Inconel. Both loops were charged with a salt composed
of LJ'_F‘=-BeF2-ThFh=UFh (62-36,5-1-G.5 mole %).

Salt samples are being teken from these loops, while they are
operating, with gradually decreasing frequency. The samples taken
from the INOR-8 loop during the first 1000 hr of operation showed
a slow but steady increase of chromium in the salt from about 420
to 530 ppm. For the Inconel loop, the increase in the chromium
content of the salt was somewhat more rapid, with the chromium content
in the salt having increased from about 350 to about L50 ppm in 500
hr of operation. Operation of these loops is continuing. In neither
loop has the chromium content of the salt reached a steady-state value.

Effect of Frel Composition on Corrosion Equilibria., Equilie-

bration studies of the irdividual reactions involved in the corrosion
of structural metals by molten fluorides have been continued. Most
of the recent experiments have been designed to refine previous

results by obtaining better material balances and to corroborate pre-

- JCT .-

viously described9 equilibrium behavior.

Attention has also been given to some instances of anomalous
results. The repeated experiments have shown the anomalous results
to be fallacious, but the puzzle of their origin has not been com-
pletely solved. There are preliminary indications of an effect of
the pretreatment of pure Fe on the extent of the reaction

O_A
2UFh + Fe ir——.Fee + 2UF3

in typical fuel mixtures, which might mean that thermodynamic
equilibrium concentrations for this reaction are not yet definitely

known.

VAPOR PRESSURES OF MOLTEN SALTS

BeF2 Mixtures. An apparatus for obtaining vapor pressures by

weighing the salt saturating a known volume of transpired inert

gas is being used for measurements of the CsF-BeF, system. The
objective of the experiments is to determine the thermodynamic
activities as a function of composition; a combination of total
vapor pressure and transpiration results is necessary in order to
ascertain the molecular species in the vapor. The LiF«-BeF2 systen,
because of complex association in the vapor phase, is not amenable
to a determination of activities from vapor pressures, but the
results for the CsF-BeF2
good estimates for LiF--BeF2 mixtures.

UFh’ The vapor pressure of liquid UFh was measured between

b and 180 mu Hg (1030 to 1300°C), and the following relationship

system are expected to provide a basis for

9J. D, Redman, MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL=2626,
P 9.

-..108. -

was obtained:

202.44 £ 0,15

82,280 + 220 = 19
2.303 R ’

2.303 RT R

log p (mm Hg) = - log T +
where T is in %K. At the normal boiling point, 1729°K, the heat
of vaporization is L49.4 kcal/mole. A re-evaluation of available
literature data for the sublimation pressure of the solid led to
the re}ationship

16,500 X 450
T

log p (m Hg) = - + 13,294 + 0,420

and to a heat of fusion estimate of approximately 20 kcal/moleo

ALUMINUM CHLORIDE VAPOR AS A HEAT TRANSFER MFEDIUM AND TURBINE WORKING
FLUID

Estimation of Thermodynamic Properties. Gaseous aluminum chlor;de
under a pressure of 1 atm consists almost entirely of A12016 at
h50 ¥, whereas at 1200 OF the fraction dissociated is O. 7. The

resulting increases in specific heat and thermal conductivity make

this gas attractive as a working fluid for a turbine. In the pre-
vious report,10 results of calculations of dissociation equilibria,
effective specific heat, thermal conductivity, and viscosity of the
gas at various temperatures and pressures were presented. These

calculations were extended considerably during the quarter and are

being incorporated in a topical report.ll The results of a study of

lQM Blander and R. F. Newton, MSR Quar. Prog. Rep. Oct. 31,
1958, ORNL-2626, p 103.

;M Blander et al., Aluminum Chloride as a Thermodynamic
Working Fluid and Heat Transfer Medium, ORNL-2677 (to be issued).

- JUJ -

the equilibria of possible mechanisms for the corrosion of metals by
gaseous AlCl3 are also presented in the reporto11 The velocity of
sound (co) in the gaseous medium, which would be an important parame-
ter in turbine design, was calculated from the expression12

2 . /X
o ""’/(ap)s ?

where v is the specific volume of the gas in cm3/g and p is the
pressure of the gas. The term (3v/5 p)S was evaluated at constant
entropy (S) from standard thermodynamic relationships and the
equation of state for an ideal associating gas.

The calculations indicate that aluminum chloride is probably
an attractive gaseous heat transfer medium and that it would require
very low pumping power. As a working fluid for a turbine, the rela=
tively large difference Qf the specific volume of the gas at the
high and low temperatures of the cycle, as compared with a working
fluid that does not associate, minimizes the effect of inefficiencies
of the pump and turbine system.

Corrosion of Nickel by.AlCl3°
the compatibility of aluminum chloride and structural metals were

Experimental investigations of

continued. Aluminum chloride gas was held at a pressure of about
1 atm for 850 hr in a capsule of metallic nickel whose ends were
maintained at temperatures of 670 and hSOOC, respectively. A very
small metallic deposit, which was identified spectrographically as
nickel, was found on the capsule wall in the 450°C zone. It is
likely that this deposit was formed by the temperature-sensitive

reversible reaction

12J. O. Hirschfelder, C. F. Curtiss, R. B. Bird, Molecular
Theory of Gases and Liquids, Wiley (1954).

-- 110 -

AlCl3 + Ni ==~ AlCl + N:LC]..2 9
which was described previously.13 The quantity of material deposited
could account for very little attack in the hot zone. Photomicrographs
of the capsule showed general and intergranular attack in the 67000
zone, with the maximum penetration of the intergranular attack being

to a depth of about 1 mil. No subsurface voids were observed.

Further examination of the capsule after completion of the experi=-
ment indicated that a cold spbt was present on the capsule near the
67000 zone during the run. It is possible that condensation of Nicl2
on this cold spot could have resulted in increased corrosion of the
wall.

PERMEABILITY OF GRAFHITE BY MOLTEN FLUORIDE MIXTURES

Previous reports in this serieslu have described tests in which

reactor-grade graphite was impregnated by exposing out-gassed specimens
to molten LiF-MgF, or NaF-LiF-KF mixtures at 1700°F for several hours
under inert gas at a pressure of about 20 psig. In order to ascertain
whether such treatment protects the graphite from penetration by reactor
fuel mixtures, 0.5~ and l.0-in.-dia specimens impregnated with the

salt mixture Lin-MgF2 (67.5-32.5 mole %; mp, 1350°F) were exposed at
1250°F for 800 hr to a typical fuel mixture, LiF-BeF,~UF) (62-37-1

mole %).

- No evidence of gains or losses in weight of the graphite specimens
resulted from exposure to the fuel mixture. However, chemical analysés
df successive 1/32-in. layers machined from the specimens indicate
that penetration of the fuel mixture had occurred. The beryllifim concen=

13R. E. Moore, MSR Quar. Prog. Rep. Oct. 31, 1058, ORNL-2626, p 102.
1L

G. J. Nessle and J. E. Eorgan, MSR Quar. Prog. Rep. June 30, 1958,
ORNL=-2551, p 99; MSR Quar. Prog. Rep. Uct. 3L, 1950, URNE-QEEE, P fU?.

-.111 -

tration of the successive layers decreased from 5000 ppm in the outer
layer tq 2000 ppm in the center of the rod; the uranium concentration
similarly decreased from 3500 to 1000 ppm. These figures indicate that
nearly two=-thirds of the LiFuMgF2 in the outer layer of the specimen

was replaced by the fuel. The ratio of beryllium to uranium was sub=-
stantially lower in the center of the graphite than in the-ffiel mixtures,
but it is evident that impregnating graphite with LiFnMgFg will not
satisfactorily prevent the subsequent penetration by LiF-sBeF2-UFh mixtures.

EFFECTS OF THERMAL CYCLING ON SALT STABILITY

A geries of tests was run with sealed nickel capsules to determine
the effects on beryllium-containing salt mixtures of freezing and
remelting, without agitation. The mixtures LiF-BeF,-UF) (62-37=1 mole %),
LiF-BeFQ-ThFh (71-16-13 mole %), and LiF-BeF,-ThF) -UF, (62+36.5=1~0.5
mole %) were tested. In one set of experiments, the salts were held
at‘12509F'under gtatic conditions for 25, 100, and 500 hr. In another
set, the salts were thermally cycled five times from 400 to 850, 1000,
1200, and 1400°F, respectively. The maximum temperature of the cycle
was held for 30 min before cooling was started. All tests were run in
duplicate.

Analyses of sections from the salts held at 1250°F for varying
periods of time showed no detectable change in composition. The salts
which were subjected to thermal cycling showed, however, definite
increases in uranium and/or thorium content in the bottom sections of
the capsules, with a corresponding decrease in lithium and beryllium
content. The degree of concentration of‘UFh or ThFLl in the bottom sections
was greater at the lower maximum cycle temperatures. The samples taken
to the higher temperatures were apparently mixed by convection.

It is evident that when beryilium-based fuels are handled care
must be taken to ensure compiete melting of the batch before any portion
- 110

of it is transferred to another container or test rig. Thermal cycling
of these salts under static conditions must be avoided.

RADIATION DAMAGE STUDIES

INOR-8 Thermal-Convection Loop for Operation in the LITR. The ine
pile thermal-convection loop comstructed for testing fused-salt fuels
in an INOR-8 System in the LITR was operated in preliminary out=of-pile
tests, and satisfactory circulation was obtained. The loop was filled
with solid fuel, which was melted in Place., Removal of gas bubbles
from the fuel was found to be a problem but was accomplished by pressure

changes brought about by alternately refrigerating and warming the
charcoal-adsorber bed of the loop. During radiographic examinations
of the loop to determine the extent of gas removal, a dense deposit
was noted at the bottom of the lower bend, which is presumably U0, .
"The amount of UO2 observed would generate an intoierable amount of heat
at that point if the loop were operated in-pile. Examination of the
batch of fuel from which this loop filling was obtained revealed UO2
deposits on the surfaces of some pieces. Analyses indicate that the
fuel circulating in the loop is free of U02° A second in-pile loop iz
being assembled that will be filled with selected pieces of Uogafree
fuel.

In-Pile Static Corrosion Test. Two INOR-8 capsules that had been
filled with LiF-BeF,-UF, (62-37-1 mole %) and shipped to the MIR for

2
irradiation were found upon arrival to have suffered mechanical damage

to the thermocouples that prevented their insertion., These capsules

are being repaired.

PREPARATION OF PURIFIED MATERIALS

Fluorides of Chromium. Chromous fluoride of good purity results

-.113 -

15

from the reaction of molten stannous fluoride with chromium metal,
The container material needed for the reaction is "graphitite," an
expensive, low-porosity grade of graphite; stannous fluoride melts
leak through containers of ordinary graphite. If cadmium fluoride or
bismuth trifluoride is substituted for the stannous fluoride in the
reaction, comparable yields (over 90%) of chromous fluoride are obtained
in crucibles of ordinary graphite. Molten cupric fluoride and lead
fluoride also react with chromium metal to form chromous fluoride, but
the yields are lower, probably as a result of loss of material by leak-
age through graphite. Cadmium metal volatilizes from the charge when
CdF2 is used as the oxidant. When the other oxidants are used, the
salt and metal separate into well-defined layers.

Batches of less than 4O g of chromous fluoride are conveniently
prepared by maintaining chromic fluoride at 1100°C in a platinum or
graphite container for 4 hr. The yield of chromous fluoride is con-

sistent with the equation

3CfiF3———> 2Cr.F2 + CrF5 °

Attempts to prepare large batches of CrF2 by the disproportionation
of CrF3 at 1100°C for a limited time, however, yield a product whose
composition is intermediate between CrF2 and CrF3° This intermediate
fluoride has also resulted from (1) partial reduction of CrF3 by E,
and (2) partial oxidation of CrF, by HF. It can apparently be prepared
in nearly pure form by treatment of CrF2 with Sn‘F2 or by reaction of
Cr, CrClQ, or CrCl3 with an excess of Sano The y=llowish green crystals
of this compound are biaxial negative and show a distinctive X-ray
pattern by which the compound can be identified readily. Chemical

analysis shows the F/Cr atom ratio to be 2.4. The compound may plausibly

5. J. Sturm, MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL-2626;, p 9.

- 11h--

be represented as chromous hexafluochromate, that is, CrII3(CrIIIF6)20

The reaction, presumably, is

5CrF +-SnF2——€> Cr3(CfiF6)2 + Sn

2

The oxidation does not procezd further, even if a 400% excess of SnF,
is provided.

Synthesis of Simple Fluorides by Reactions with Stannous Fluoride.
Stannous fluoride is also useful for synthesizing fluorides of other

metals in a low valence state. Molten stannous fluoride was reacted
with manganese, zinc, aluminum, iron, vanadium, and uranium to yield,
s ZnF,, AlF.,, FeF,, VF_, and Uic The equation for

regspectively, MnF 09 3 59 3
a typical reaction is

Mn + San———>-MnF2 + Sn .

The manganous and zinc fluorides prepared by this procedure were pure,
since the molten salt and the tin resulting in the reaction formed
well-defined layers which were easily separated. The reaction of the
uranium metal with stannous fluoride was so vigorous that the molten
product was spattered about.

Experimental-Scale Purification Operations. Small-scale proceszing

of various molten fluoride salts for use in corrosion tests, physical
property studies, and small=gcale component testing continued in the
experimental facilities located in Building 9201-3 and Building 9928.

A total of 265 kg of mixtures without beryllium and 73 kg of beryllium-
containing compositions was procesged during the guarter,

Trangfer and Service Operations. Service cperations were about b,

normal during the guarter, with approximately 98 kg of liquid metals
and 1123 kg of fluoride salt mixtures being dispensed. Approximately -
112 operations wers involved in the normal procegses of filling, draining,

and sampling various test equipment,

- 115 -

2.3. FUEL PROCESSING

A process for the recovery of the fuel and blanket salts of a molten-
salt reactor is being developed that is based on the volatilization of
uranium as UF6 and on the appreciable solubility of LiFnBeF2 salts in
nearly anhydrous HF. Previously reported work indicated this procedure
to be generally applicable to molten-salt reactor systems employing
LiF-BeFe-UFh or LiF--BeFe-UFh-ThFh salts. The process is particularly
promising in the degree of separation of the recovered LiF-=BeF2 salt
from important rare-earth neutron poisons as a result of the general
insolubility of polyvalent-element fluorides in anhydrous HF. A
tentative process flowsheet was presented previously.l

Studies of the behavior of neptunium, uranium, thorium, and
corrosion=product fluorides have been made in further development work on
the HF salt-diszolution process. The results appear to be consistent
with the proposed process flowsheet in that the recovered LiF--BeF2 salt
would effectively be decontaminated from neptunium. Much further investi-
gation of some of the variables involved will be necessary, however,

before specific flowsheets for specific reactor designs can be prepared.

SOLUBILITY OF NEPTUNIUM IN AQUEQUS HF SOLUTIONS

The rate of neptunium buildup in a U235 burner reactor increases
with time as a result of buildup of the parent U236 isotope. After 20
years, the neptunium concentration will be approximately 0.015 mole %
if there is continuous removal on a once-per=year cycle. For such a

reactor, the neptunium would contribute about 40% as much poisoning as

lM R. Bennett, D. O, Campbell, and G. I. Cathers, MSR Quar. Prog.
Rep, Oct. 31, 1958, ORNL-2626, p 111, Fig. 2.4.1.

- 116

contributed by all fission fragments at that 'bime.2

The solubility of neptunium in 80 to 100% HF solvents saturated
with LiF and BeF2 was found experimentally to be sufficiently low to
permit its removal along with the rare earths. In the experiments,
small quantities of neptunium (agueous nitrate solution) were added
incrementally to HF splutions saturated with LiF and BeFe. The concen-
tration of nitrate added was approximately 0.02% of the fluoride salt
concentration. The observed solubilities of neptunium are reported

in Table 2.3.1 as milligrams of neptunium per gram of solution and as

Table 2.3.1 Solubility of Neptunium (IV) in LiF-BeF

Saturated Aqueous HF e

Approximate Solubility in Solvent Np Solubility

HF Concentration - (mg/g of solution) Relative to Salt
In Solvent (wt %) TiF Béféfii Np Solubility (mole %)
80 29 111 0,026 0.0031
90 64 70 0.011 0.0012
94 96 60 0.0086 0.00072
100 112 40 0.0029 0,000214

mole % of neptunium relative to the mole % of dissolved LiF-BeF. salt.

The neptunium was determined by alpha counting and pulse-height

2

analyses in order to distinguish neptunium from the plutonium present

as an impurity. The plutonium incidentally, appeared to be carried

2Molten-Salt Reactor Program Status Report, ORNL-2634 (Nov. 12, 1958).

- 117 -

with the neptunium to a considerable extent; the ratio of the amount of
plutonium in solution to the amount undissolved was within a factor of
2 of the ratio for neptunium although the plutonium concentration was
500 times smaller. |
Additions of iron and nickel metal to the HF solutions resulted
in a significant reduction in the gross alpha activity, probably as the
result of reduction to neptunium (III) and to plutonium (III). The
reduction has not yet been verified by neptunium pulse-height analyses.
In the trivalent state, behavior somewhat similar to that of the rare
earths might be expected; the observed solubilities are in the same
range as those reported previously for rare earths. It is éxpected
that the rare earths, neptunium, plutonium, and, possibly, uranium will
behave as a single group and will therefore exhibit lower solubilities
when present together than the values reported here for the separate

components.

SOLUBILITY OF CORROSION-PRODUCT FLUORIDES IN HF SOLVENT

Measurements were made of the solubilities of corrosion=-product
fluorides (iron, chromium, and nickel) in HF solutions saturated with
LiF. Chromium fluoride was found to be relatively soluble, with the
solubilities being about 8, 12, and 18 mg of chromium per gram of
solution in 100, 95, and 90 wt % HF solvents, respectively. Iron end
nickel fluorides were found to be less soluble. The iron fluoride
solubility values varied from 0.08 to 2 mg of iron per gram of solution;
the average deviation of the values, neglecting extreme values, was
0.1l i 0.10. The nickel fluoride solubility varied from 0.0l to 1.3
mg of nickel per gram of solution, the average being 0.15 i.0°06° The
regults were so inconsistent that no trend with HF concentration could
be established. When iron, nickel, and chromium fluorides were presént

together, the solubilities of all appeared to be somewhat lower.
< 118 -

Similar measurements in 90 and 100 wt % HF solvents with no dissolved
salt gave values in the same range, but in the 80 wt % HF solvent the
iron and nickel solubilities appeared to be sgignificantly higher.

SOLUBILITIES OF URANIUM AND THORIUM TETRAFLUORIDE IN HF SOLVENT

Measurements of the solubilities of thorium and uranium fluorides
in the HF solvents indicated that these fluorides are relatively insolu-
ble. All ThFh determinations gave less than 0.03 mg of thorium per
gram of solution, which is the limit of detection. Better analytical
methods are available for UFh’ and the solubility values were, in general,
in the range 0.005 to 0.010 mg of uranium per gram of solution. In the
presence of fresh iron or nickel metal, somewhat lower solubilities
(0.002 to 0.005) were observed, indicating, perhaps, a higher solubility
for higper oxidation states. |

»

WO O Ovan =W o =
L] »

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- 119 -

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ORNL-268L4
Reactors-Power
TID-4500 (14 th ed.)

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- 120 -

89. G. C. Williams
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96-135. Laboratory Records Department
136. Laboratory Records, ORNL R.C.
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Reports previously issued in this series are as follows:

ORNL~-2378 Period Ending September 1, 1957
ORNL~2431 Period Ending October 31, 1957
ORNL~247k Period Ending Januvery 31, 1958

ORNL~2551 Period Ending Junz 30, 1958
CRNL~2626 Period Ending October 31, 1958