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R oRNL 2551

Reactors—~Power

T1D-4500 (13th ed., Rev.)

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MOLTEN-SALT REACTOR PROGRAM
QUARTERLY PROGRESS REPORT
FOR PERIOD ENDING JUNE 30, 1958

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OAK RIDGE NATIONAL LABORATORY
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for the

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ORNL-2551
Reactors~Power
TiD-4500 (13th ed., Rev.)
February 15, 1958

Contract No. W-7405-eng-26

MOLTEN-SALT REACTOR PROGRAM
QUARTERLY PROGRESS REPORT

For Period Ending June 30, 1958

H. G. MacPherson, Program Director

DATE ISSUED

Fad

SEP 1 G958

 

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION

for the ARTIN MARIETTAENERCY S8

e A

3 wy5k 0350616 1

TEMS LIBRARIES
0, KA i M S

  
MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT

SUMMARY
PART 1. REACTOR DESIGN STUDIES critical inventories required for reactors fueled
Desi di with U235, In comparison with corresponding
1.1. Design Studies U237 reactors, the critical inventories of plutonium-

A conceptual design of a power reactor, desig-
nated the ‘‘interim design reactor,’”’ was prepared.
This two-region, homogeneous reactor uses molten
salt fuel which is circulated by a single fuel
pump located at the top of the core. The net power
output is 260 Mw, and the reactor has a thermal
capacity of 640 Mw, The total cost of the power
produced by this reactor is estimated to be 8.9
mills/kwhr, of which 2.5 mills/kwhr is fuel cost.

The core of the reactor was designed to have a
volume approximating that of an 8-ft-dia sphere,
and the shapes of sections taken through the
walls can be expressed as simple algebraic
functions. This latter criterion assures smooth
shapes, easily calculated volumes, and dimensional
reproducibility.

An analysis of the weights of various portions
of the system was made to aid in construction
planning and cost estimating. It was estimated
that the empty equipment would weigh 684,000 Ib,
the fuel salt would weigh 102,358 [b, the blanket
salt would weigh 194,769 |b, and the sodium
would weigh 100,188 |b.

A detailed evaluation of the fabricability of the
reactor was made, and it was established that
of the pressure-vessel
industry could be used. It is recommended that
the initial reactor be fabricated entirely in one
shop so that tolerances need not be so rigid for
fitup of parts as they would be if parts were
supplied by various vendors,

Design work was compieted on the fuel pump
proposed for the interim design reactor. The
motor and pump rotating assembly are completely
replaceable remotely as a cartridge unit. The
bearing is a salt-lubricated orifice-com-
hydrostatic bearing, and the upper
bearing is a hemispherical orifice-compensated
hydrostatic type that uses the pressurized helium
purge supply for lubrication and support.

conventional practices

lower
pensated

Nuclear calculations were continued, and some
were obtained for reactors fueled with
The initial regeneration
ratios 1.03 at critical
inventories of U233 that were much less than the

results
U233 agnd with plutonium,

obtained ranged up to

fueled reactors were about one-third as great.
The required plutonium concentration was found
to be well below the solubility limit in salts of
interest.

A comparative study of the various gases that
might be used as cover gases for molten-salt
reactor systems revealed that helium and argon
are, at present, the only suitable gases. Helium is
less expensive than argon, but it is not available
outside the United States,

Test rigs were designed for experimental
studies of three types of bearings. These bearings,
which are designed for molten-salt application,
are (1) a hydrodynamic bearing with conventional
and sleeve, (2) o hydrostatic, orifice-
compensated bearing mounted on the hub of the
impeller with stationary pockets supplied with
high-pressure molten salts, and (3) a hydrostatic
bearing with rotating pockets,

journal

1.2, Component Development and Testing

Development tests of salt-lubricated bearings
are under way with an INOR-8 journal and sleeve
being tested in the hydrodynamic bearing test rig.
This initial bearing failed when the thrust load
was increased from 100 to 125 Ib with the shaft
operating at 1200 rpm and the molten salt at a
temperature of 1200°F, Other journal and sleeve
materials are being studied, A rotating-pocket
hydrostatic bearing was fabricated for testing.

Tests of conventional organic-liquid-lubricated
A sump pump in which
the lubricant is Dowtherm A has operated satis-
factorily for 1620 hr with a molten sailt as the
pumped fluid at a temperature of 1200°F. The
shaft is 2600 rpm. The oil-lubricated
pump rotary assembly that is being operated in a

bearings were continued.

speed

gamma-irradiation field at the MTR has accumu-
lated a gamma-ray dose of 5.9 x 107 r. The bulk
of the Gulfcrest 34 lubricant is external to the
radiation field, and thus the exposure of the oil
has been only 108 r,

Calculations were made of the conditions re-
quired for operation of a gas-lubricated bearing.

ihi
A small hemispherical bearing is to be used to
test the validity of the calculations.,

Test operation of a NaK pump with a labyrinth
and split-purge seal arrangement was terminated
when the passages to the labyrinth became con-
stricted. Material that contained carbon was found
to have blocked the passage. Face surfaces of
the lower seal, which had operated 3386 hr, were
free of NaK and were in good operating condition.
A bellows-mounted seal being subjected to an
endurance test in a NaK pump continued to operate

Over 5100 hr of operation has
Since there are no elastomers in

satisfactorily.
accumulated.
this type of seal, it may be suitable for operation
in a radiation field.

Components of electric motors are being tested
for service at high temperatures in a radiation
field, since the successful deveiopment of salt-
lubricated bearings would make totally submerged
canned-rotor pumps applicable to molten-salt
systems. Investigations are under way of high
permeability magnet steels, insulating materials,
and current-carrying conductors.

Tests were made to determine the length of
time required to bring a typical fused-salt piping
system to a temperature of 900°F by preheating
a portion of the system with electric-furnace
elements and transmitting the heat to the remainder
of the system by forced circulation of helium
With about 49% of the surface
heated, ¢ hr and a minimum helium pressure of 98
psig were required to reach 900°F.

Screening tests of mechanical joints for remote
separation of system components were continued.

within the piping.

None of the joints tested have given any evidence
of leakage during tests with molten salts, Joints
are being assembled for tests with sodium,
Experimental remote maintenance work on a
NaK pump was completed, and the results indicated
the feasibility of remote maintenance work on
accessible components of a molten-salt system,
A  three-dimensional viewing system is being
developed. High-quality welds were produced in
preliminary
emphasized the need for a good viewing system,
An engineering layout was prepared for a remote

remote welding experiments, which

maintenance demonstration facility, and detailing
and fabrication of the components are under way.
Combination heater and insulation units designed
for remote application and removal were fabricated
for testing.

Commercially available expansion joints were
ordered for testing. If expansion joints can be
used in fuel and coolant circuit piping, the extra
space and fluid inventories required for thermal
expansion loops could be avoided.

Processing of the data obtained from a heat
transfer coefficient test in which fuel 130 (LiF-
Ber-UF4, 62-37-1 mole %) was circulated in
an available heat exchanger test facility was
completed. The data are presented in comparison
with similar data for other fluids. A molten-
salt-to-air radiator was designed and is being
fabricated for a molten-salt-to-molten-salt heat
transfer test,

Operation of forced-circulation and thermal-
convection corrosion-testing loops was continued,
Improvements are being made to the test stands to
assure operational reliability.

Out-of-pile tests of components of the in-pile
loop being prepared for insertion in the MTR
were continved, Difficulties with gravity-filling
of the system because of the high surface tension
of fuel 130 have delayed the work, New filling

techniques are being developed.

1.3. Engineering Research

Preliminary values were obtained for the vis-
cosities of the salt mixtures NaF-Ber-UF4
(53-46-1 mole %), LiF-BeF -UF , (53-46-1 mole %),
and LiF-Ber-UF4 (62-37-1 mole %) over the
temperature range 500 to 900°C. The enthalpy and
heat capacity of the mixture LiF-BeF,-UF,
(62-37-1 mole %) were established in the temper-
ature range 100 to 800°C. Studies were initiated
of the thermal conductivity, surface tension, and
thermal expansion of the beryllium-containing
fluoride salts, Initial measurements of the heat
transfer coefficient for LiF-Ber-UF4 (53-46-1
mole %) flowing through a heated Inconel tube have
indicated that this salt behaves, with respect to
heat transfer, in the same manner as ordinary fluids,

Hydrodynamic studies were continued with small-
scale glass models of proposed reactor cores,
In the straight-through flow model it was found
that the inlet high-velocity flow essentially short-
circuited the core and passed directly from the
entrance to the exit without appreciable spreading.
The remainder of the core was filled with slowly
rotating fluid that had extremely low velocities

along the sphere wall. The concentric system
with annular inlet flow exhibited a number of
peculiarities which can be associated with the
shortness of the annulus. The main fiow was
down along the sphere surface located 180 deg
from the inlet elbow and up along the back surface
at the 0-deg position.
in the equatorial plane at the center of this fiow.
Extension of the central pipe in the concentric

A tapering vortex existed

pipe-entrance system showed increased velocities
at the bottom of the sphere.

1.4. Instrumentation and Controls

Inconel-sheathed Chromel-Alumel thermocouples
with magnesium oxide insulation and hot-junction
closure welds made by the Heliarc welding process
are being tested for endurance and stability., In
10,000 hr of exposure to sodium at 1500°F, only
two of 38 thermocouples have failed because of
weld closure deficiencies. Drifts from initial
temperature readings are within £0.75%.

Test facilities were prepared for investigating the
suitability for molten-salt reactor service of the
resistance-type fuel level indicator, Design work
is under way on modifications required to improve
a commercial mechanism for use in switching low-
level transducer signals. Various types of pressure
transducer are being evaluated for molten-salt
service,

1.5. Advanced Reactor Studies

A conceptual design was prepared of a 5-Mw
experimental reactor in which molten salt fuel
would be circulated by thermal convection, This
simple, reliable system could be converted to a
50-Mw pilot plant by adding a fuel pump and in-
creasing the capacity of the heat dump. The
5-Mw reactor could be constructed of components
already developed. It would demonstrate the
feasibility of continuous operation of a molten-
salt reactor, provide in-pile corrosion data, and
serve as a mockup to develop and demonstrate
maintenance procedures, The 50-Mw
system would be sufficiently similar to a large-
scale power-producing plant to lead directly to

design and construction of a large power plant,

remote

The possibility of a thermal-convection reactor
of approximately 600-Mw thermal output was also
investigated. |t was found that a fuel inventory
of 1775 13 would be required, which is to be
compared with the 530 ft° estimated for a reactor
system in which the fuel is circulated by a pump.

The heat exchange equipment that would be
required for gas cooling of a molten-sait reactor
was studied, Helium, steam, and hydrogen were
the gases considered. For a given set of con-
ditions, hydrogen was the most effective, but
helium and steam were reasonably comparable. An
optimization of the size of the tubing to be used in
the heat exchanger gave a value of 0,5 in. Larger
tube diameters led to excessive tube lengths, and
smaller diameters led to large numbers of tubes in
the matrix,

PART 2, MATERIALS STUDIES

2.1, Metallurgy

Metallurgical examinations and corrosion evalu-
ations were made of specimens from several
Inconel and INOR-8 thermal-convection loops in
which various fluoride mixtures were circulated.
One of the Inconel loops, which were operated at a
maximum hot-leg temperature of 1250°F in order to
determine the corrosive effects of various fluorides
under MSR temperature conditions, gave results
which contradict the corrosion postulate that ThF4
should effecta lower corrosion rate than comparable
additions of UF ,. No explanation for the increased
attack by the ThF ,-containing mixture is readily
available, INOR-8 thermal-convection
loops that were operated under MSR temperature
conditions for 1000 hr with various fluoride mixtures
showed no attack, and one INOR-8 loop that was
operated for more than 6300 hr was found to have
widely that ranged
to a maximum depth of 0.75 mil at the hottest
point. The initial results for INOR-8 loops are
favorable.

Seven

scattered subsurface voids

Specimens for studies of the effect of carbu-
rization on the mechanical properties of Inconel
and INOR-8 were prepared by exposure in a sodium-
graphite system, since carburization takes place
slowly if at all in a molten-salt system, Control
specimens were given the same heat treatment
in an argon atmosphere, INOR-8 was found to
have been more heavily carburized than Inconel
by the sodium-graphite system. Tensile tests
showed that carburization increased the tensile
strength and yield strength of Inconel and reduced
its ductility, The [NOR-8 specimens were found
to have a lowered tensile strength, slightly in-
creased and greatly reduced

yield strength,
ductility, Studies of INOR-8 and Inconel! in salt-
graphite systems are being planned.

Preliminary tests have shown brazing alloys with
high gold and silver contents to have promising
Therefore
long-term corrosion data are to be obtained by
insertion of samples in the hot legs of thermal-
convection loops.

corrosion resistance to molten salts.

Good correlation has been found among tensile
property data for INOR-8 obtained at ORNL, Haynes
Stellite Company, and Battelle Memorial Institute,
Data are now available on yield strength, tensile
strength, ductility, relaxation, and Young’s modulus
as functions of temperature. Preliminary creep
data have been obtained, and extensive creep
tests are under way.

An investigation of the influence of composition
variations on the creep-rupture strength and the
microstructure of INOR-type alloys was completed,
A general consideration of all the data obtained
favorably supports the composition selected for
the alloy INOR-8, Embrittlement studies have
indicated that aging in the temperature range from
1000 to 1400°F has no significant effect on the
properties of INOR-8,

Five air-melted heats of INOR-8 were prepared
by Westinghouse, and about 20,000 [b of finished
products will be supplied from these heats. The
first shipment of seamless jubing was received,
and it was found to be of excellent quality.

Experimental studies of bearing materials are
under way. Flame-sprayed INOR-8 coatings were
successfully bonded to INOR-8 journals, Similar
molybdenum coatings cracked severely and sepa-
rated from the INOR-8 wupon thermal cycling.

Molybdenum rods sprayed with molybdenum are
now being tested.

Additional welding studies have further indi-
cated that the weldability of INOR-8 is satis-

Sound welds can be made, and the weld
characteristics of the material are

factory.
deposition
comparable with those of Inconel or the stainless
steels, Tests of all-weld-metal specimens of
INOR-8 and Inconel have shown that INOR-8 has
a slightly higher ultimate tensile strength than
that of Inconel, a significantly higher yield strength,
and a markedly lower high-temperature ductility,
A weld made in a 10-in.-dia Inconel pipe with a
5/B-in. wall by a semiremote welding process being
developed for the PAR project by Westinghouse

was examined. Radiographic examination showed

vi

the weld to be completely sound; there was no
evidence of porosity.

Examinations of the cast-metal seals of two
flanged joints that were tested with molten salts
It was found that slight oxidation
had impeded wetting. The joint with a silver-copper
alloy seal appeared to be less subject to non-
wetting than the joint with a pure silver seal.

Studies are under way of an internal tube weld-
ing procedure being developed by the Griscom-
Russell Company that would be applicable to the
attachment of tubes to tube sheets in heat ex-
changers if back-brazing were impractical, Such
a procedure may be applicable to the fabrication
of the large heat exchangers that will be required
for molten-salt reactors.

were made,

2.2. Radiation Damage

Apparatus is being assembled for in-pile tests
of the corrosion of INOR-8 by the molten salts of
interest.  An electrically heated mockup of a
loop for operation in the LITR is nearing com-
pletion. Parts for the in-pile model have been
fabricated.

Final examination of an Inconel loop that circu-
lated o molten salt in a vertical hole in the LITR
showed the corrosion to be the same as that which
would have been expected in the absence of
radiation. A new fuel salt sampling method in
which the salt is melted out was found to be as
satisfactory as the previous method of drilling
into the salt to obtain a sample.

Further preparations were made for the instal-
lation and operation of a forced-circulation loop
in an ORR facility. Studies of bearings for use in
the loop pump are under way, and the motor of the
used in previous in-pile loops is being

pump
redesigned.

Inconel capsules for testing the stability of
graphite in contact with molten-salt fuel were

shipped to the MTR for irradiation ot 1250°F.

INOR-8 capsules were fabricated for similar tests,

2,3, Chemistry

Phase equilibrium studies of LiF-BeF, systems
and/or ThF, were continued. A
mol ten-salt breeder-reactor fuel with a liquidus
temperature of 440 + 5°C and with no more than
36 mole % BeF, is available in the LiF-BeF,-
ThF -UF , system. Studies of the LiF-BeF ,-ThF,

containing UF,
system have shown the existence of three eutectics
with temperatures in the range 360 to 429°C,
Studies of the LiF-ThF ,-UF, system have shown,
as was expected, extensive formation of solid
solution,

Additional data were obtained on the solubility
of PuF, in alkali fluoride=beryllium fluoride
mixtures, The data indicate that the solubility
of PuF, in LiF-BeF, mixtures is at a minimum
for mixtures containing about 63 mole % LiF and
that it is at a minimum in the NaF-BeF, system
for mixtures containing about 57 mole % NaF.
Data for the solubility of PuF, in an LiF-BeF,
(63-37 mole %) mixture containing 1 mole % ThF,
indicate that the addition of ThF, does not
appreciably affect the solubility of PuF, in this
solvent,

Further experimental measurements were made
of the solubilities of the noble gases in molten
salt mixtures, Data were obtained for the solubility
of argon in NaF-KF-LiF (11.5-42-46.5 mole %) at
600, 700, and 800°C and of helium in LiF--Bel:2
(64-36 mole %) at 500 to 800°C., The trends of
the data were the same as those previously
observed with mixtures containing ZrF,. [t has
been demonstrated that the solubilities of HF in
LiF-BeF, and in NaF-ZrF, mixtures are about
the same when the alkali fluoride content is low,
As the alkali fluoride content is increased, how-
ever, the solubilities of HF in the two mixtures
differ markedly.

Studies of the solubilities of fission-product
fluorides in molten alkali fluoride—~beryllium
fluoride systems were continued. Data were

obtained for the solubility of Ce F, over the temper-
ature range of 450 to 700°C for Lil:-BeF2 and
NaoF-BeF, mixtures containing 50 to 70 mole %
alkali fluoride. It was found that the solubility
passed through a minimum at about 62 to 63
mole % alkali fluoride in both solvents,

The chemical reactions of oxides with fluorides
in LiF-KF are being studied in an investigation
of the chemical separation of solutes in fluoride

mixtures by selective precipitation as oxides,
The characteristics of BeO as a precipitating
agent are being studied.

The activity coefficients of NiF, dissolved in a
molten mixture of LiF-BeF. (62-38 mole %) are
being determined. Data obtained at 600°C gave
calculated activity coefficients of 2347 and 515
with respect to the solid and liquid standard
states, respectively. It is known, however, that
the assumed melting point for NiF, is in doubt.
The solubility of NiF, in LiF-BeF, (61-39 mole %)
was measured and was found to be independent of
the amount of NiF, added.

A series of experiments for rechecking data
obtained at high temperatures on the diffusion co-
efficients for chromium in nickel-base alloys
and to extend the data to temperatures below
600°C is being planned. A depletion method is
to be used to check the high-temperature data, and
a constant-potential method will be used for the
low-temperature experiments,

A study of the vapor pressures of the CsF-BeF,
system was made in order to obtain information on
the effect of composition on the thermodynamic
activities in fuel mixtures containing BeF.,.
Deviations from ideal behavior were observed that
were strongly dependent on the size of the alkali
cation,

Studies of fused chlorides as heat transfer fluids
were initiated. The mixtures KCl-ZnCl,, LiCl-
ZnCl,, ond LiCl-RbCl are being investigated.

Experiments are under way to study the satu-
ration of graphite with an inert salt whose melting
point is somewhat higher than proposed reactor
fuel temperatures as a possible method for pre-
venting the graphite from absorbing molten-salt
fuel.,  Graphite rods that were soaked in and
completely penetrated by an LiF-MgF., mixture
arenow soaking in LiI=-Bel:2-Ul:4 (62-37-1 mole %)
at 1200°F,

Alteration of a production facility to provide
for the large-scale processing
containing materials was

of beryllium-

nearly completed.

vii
CONTENTS

SUMMARY ettt e e s et s r et b et et se e et e em e et be bR b r s e bbbt b

1.1.

1.2.

1.3.

PART 1. REACTOR DESIGN STUDIES
DESIGN STUDIES ..ottt s et bt et o sd e ne e nb b e e eee

Interim Design Reaetor. .ot e s st e be e s sb e et sa bbb e s
Reactor Core Configurafion ........ccovcerieriirerieiercercnireceeceee e ettt bbb easser et s b ebasneas
Weight Analysis of Reactor System .......ccoociieiiiriiiiin et s e

Evaluation of Fabricability of Reactor Vessel ..o
Outer Blanket Shell ...t ettt ettt
INNEr €ore Shell ... ettt sttt et st s bbb er e et sas bbb
Blanket System Pump Housing ..o

Fuel Pump Design ...t et et ettt ettt et st aaas

NUClEar CalcUltions ....oieii et e et ereee e st ee e e aea s e snaseese et e gessensan smseneentanns
Modifications of Oracle Program Sorghum for Calculational Analyses of
Molten-Salt REACIOrs ........cciiiiiiiiiiiece et e et rans sreaa et earesan e et eebanean
Analyses of Reactors Fueled with U233 oot
Analyses of Reactors Fueled with Plutonium ..o

Argon as a Protective Atmosphere for Molten Salts .. .....cccooiniiiiiiciii e

Bearing Tester Designs ...ttt e cme e e st e b e s e
Hydrodynamic Bearing Tester ...t e
Hydrostatic Rotating-Pocket Bearing Tester ..o e
Hydrostatic Stationary-Pocket Bearing Tester. ...

COMPONENT DEVELOPMENT AND TESTING ....cocooiieiiiiriirecrecrect et et

Fuel Pump Development ...t sttt eae s sbe st e n e et ene e
Development Tests of Salt-Lubricated Bearings .......cccooiiimimineciniie e
Development Tests of Conventional Bearings .........cc.coooiiiiiiiiiiiiic e
Gas-Lubricated Bearing Studies ..........cooooiiiiiic et sttt
MEChaniEal Seals oo et ertr b e e a s a et eresresenaaene snereenen
Radiation-Resistant Electric Motors for Use at High Temperatures...........c.oooeiiiiiiiiiiiiienn.
Piping Preheating Tests ..ot st bbb bbb

Development of Techniques for Remote Maintenance of the Reactor System.......coccoviinivirennn,
Mechanical Joint Development .. ... et a e e e
Remote Manipulation Techniques ... ..o et s
Remote Maintenance Demonstration Facility ..o e
Heater-Insulation Unit Development .......ccoccoiiiiiii et st

Evaluation of Expansion Joints for Molten-Salt Reactor Systems ..o,
Heat Transfer Coefficient Measurement ...........ccoooiiiiiiiiis i et e

Design, Construction, and Operation of Materials Testing Loops .....cccooivieiiereiviiicecec
Forced-Circulation LLoops . .. oottt ra e et n s e eee b e en e beanae s
P il LL00PS oottt bbb b r e r e be bt e rean e eheebben st eeabeas

ENGINEERING RESEARCH .ottt s e e

Physical Property Measurements ...ttt e ettt s et st sr e e
VS OSIY o oieiieiei ettt ete e ket e h e e h oo ekt s e sr st abes bt s ehe 2R At e e et aen sme e ee et e reerae e e e rereere
Thermal Conductivity ..ottt ettt e e e e ere e se it ste et ebbr e eaes

—_— D0 O NNONON O W W
1.4.

1.5.

2.1.

2.2.

Enthalpy and Heat Capaeity .....ooeioieriieicecce e et eeeer ee s ev e e s e v s e e es s s e esanssans 38
SUPFACE TERSTON ittt ettt et ee s s e es e e eee s ee e aenens 39
TREPMA] EXPANSION .....ovoeceiii ettt et e e ee e e ee e ee e e e e e e e s s et e er et eres e e eesee e 39
Hydrodynamic Studies of MSR Core ... ..ot eee e e et eenesseeeeee e, 39
Molten-Salt Heat Transfer STUAIEs ..ot s e s et e aetesesene e s nesn s 43
INSTRUMENTATION AND CONTROLS ..ottt eeees e eea e evns e s eesena s eeeeeenes 48
Endurance and Stability Tests of Sheathed Thermocouples ..........cccouoeomieioreoeeeeeee oo, 48
Resistance-Type Fuel Level INdicator . ..ot eeeees e e s e eee s ee s 48
SCANMNING SWITERES (.ot sttt et s s s e ettt e e e s se et eeaeeraene 48
PressUre TraNSAUCErS ..ottt ettt et s st ee s ees s e et e se st et s e nraeseeeens 48
ADVANCED REACTOR DESIGN STUDIES ..ottt sttt sttt et e 49
An Experimental 5-Mw Thermal-Convection Reactor ......ccueeeeeieicieeeeieeeee e et e eaeeenee e 49
A 600-Mw Thermal-Convection ReaCtOr .........c.coiiieiuiiiiecteieeieeeeeee ettt e 51
Gas-Cooled Molten-Salt Heat EXChan@er ..............ocoooviiiiieioeee e e eer e ean e 52
PART 2. MATERIALS STUDIES
MET ALLURGY ottt et et ettt e e eea et e e s eeere et eeeeeereensenenas 57
Dynamic Corrosion StUdI@S ..ottt et st eeen et se et er e e ee s e et ereeenn e 57
Inconel Thermal-Convection Loop Tests ..ot er e 57
INOR-8 Thermal-Convection Loop TestS ..ooeoiiioieeoe oo eeeeeeeteees e e oot eees s e aees e 58
General Corrosion SHUGIES ...t et et esns e senenen et esaeannes 59
Carburization of Inconel and INOR-8 ...ttt e e n e 59
Brazing Alloys in Contact with Molten Salts ..o, 62
Mechanical Properties of INOR-8 ..ot e 64
B rICAHION SHUAIES (it et ettt et s et saet ettt st et e sn e rer et are e 66
Influence of Composition on Properties of INOR-8 ... . e 66
High-Temperature Stability of INOR-8 .......... ..ottt et 67
Status of Production of INOR-8 (Westinghouse Subcontract 1067) ...ooovviviiieiiivn e 68
Status of Production of Seamless Tubing (Superior Tube Company Subcontract 1112) ............ 68
Bearing Materials ...t e ettt b e eb et e e e 69
Welding and Brazing StUdies ...ttt v e st st e r et e 69
Weldability Evaluations .. ..o oiiuieiceee et ettt ee e e et en et et en et ese s sn et asbab e tnaneras 69
REMOte Walding ..o vttt ettt e e tbeeresre e e s sstsasssaasaseesseaasssaasatansserbeaenbesataeeaniras 70
Joint Development ... e et b et e e s s e eneee et 71
Component Fabrication ....o..iiccieiiiieieeeiis e eereees st e e et sv e st eeab e s srmeseebee s st e e seteensseaastn b baeesnnesess 72
Material and Component INSPECHION .. ..oiiiciiiis ettt et e bt eb st em et 76
RADIATION DAMAGE .ottt ettt et seesre sbe b e e saesiesen e e sas e saes 78
In-Pile Dynamic Corrosion Tests . e e s 78
INOR-8 Thermal-Convection Loop Assembly for Operation in the LITR ..o 78
LITR Forced-Circulation Loop Examination .......c.ccceiiiiiiiiiiini s e 79
ORR Forced-Circulation Loop Development ............coocviiiiiiiiiivniiiiie et 81
In-Pile Static Corrosion Tests ... serre s ete e ste e s ere e s r s see e ebes b bban o 81
2.3, CHEMISTRY Lottt bbbt hbe et eaeee e ee e raea et mteste et e s es s eesseeeeenene e

Phase Equilibrium STUies ...ttt ettt e a et eee e et v e e s et e e esesseaseeeeanes
Systems Containing UF ; and/or ThF , ..o
Solubility of PuF, in Alkali Fluoride~Beryllium Fluoride Mixtures ..........ccocconnucvuinarnnenes

Fission-Product Behavior ...t oo er e st en et
Solubility of Noble Gases in Molten Fluoride Mixtures ..........cccooooviiiimeveecieeeeeeeeeeeeeereeereen
Solubility of HF in LiF-BeF, Mixtures .....ccoccoiiiciiiciiniiec it necenenessecieseseceisessisecens
Solubilities of Fission-Product Fluorides in Molten Alkali Fluoride~Beryllium

Fluoride SOIVENTS ..o e ettt s sttt

Chemistry of the Corrosion Process ..ot et crt sttt e s e e ne
Activity Coefficients of NiF, in LiF-BeF , oo,
Solubility of NiF, in LiF-BeF, (61-39 Mole %) ......cccoovviii e
Experimental Determination of Chromium Diffusion Coefficients in

Molten Salt—Incone! Systems ...t bt ee e

Vapor Pressures for the CsF-BeF, System ..o
Fused Chlorides as Heat Transfer FIuids ..o eve s
Permeability of Graphite by Molten Fluoride Salts .........o.ococviiiiiiciceeeree e,
Preparation of Purified Materials ...ttt st e eerese e e
Preparation of CrF o oo e see e e et s e st e vt ns e ene e ans
Production-Scale Operations ...ttt eeae s resserete e et ere e e saan e seessreeneerenee

Experimental-Scale OPerations ... et ee e s et e et e e e e v e re s e eeeeeae e e e eseeas
Transfer and Service OPerations ...........iccoiiriieiiisice ettt st e s ess s e e eeseestsbeesteststea sue

xi
Part 1
REACTOR DESIGN STUDIES
1.1. DESIGN STUDIES

H. G. MacPherson

Reactor Projects Division

INTERIM DESIGN REACTOR

Conceptual design studies of a power reactor
have led to an ‘‘interim design reactor,’”’ which
is described in a report entitled Molten Salt Reactor
Program Status Report. Since this report will not
be issued for general circulation until September,
a brief description of the interim design reactor
will be given here, together with some design
information that is not included in the status
report.

The interim design reactor is a two-region,
homogeneous, molten-salt reactor with a single
fuel pump at the top of the reactor core, as shown
in Fig. 1.1.1.  The gross electrical output is
275 Mw, and, since 15 Mw is required in the plant,
the net power output is 260 Mw. The net over-all
plant efficiency is about 40%, and thus the reactor
has a thermal capacity of 640 Mw.

The molten salt fuel used in the reactor core is
initially a mixture with the following approximate
composition: 61.8 mole % Li’F, 36.9 mole % BeF,,
1.0 mole % ThF,, and 0.3 mole % UF,. The
volume of the system is such that the initial
critical inventory is about 600 kg of U235, Fuyel
reprocessing is initiated at the end of one year in
small batches equivalent to reprocessing the
entire fuel charge once per year. With this system
of reprocessing, the inventory of U235 builds up
to about 900 kg at the end of the first year and
remains approximately constant for the next
20 years. The inventory of U233 gradually builds
up to about 300 kg at the end of 20 years. The
blanket salt is a mixture of 71 mole % LiF,
16 mole % BeF,, and 13 mole % ThF,. The
average conversion ratio for this reactor is 0.53
over the 20-year period.

An off-gas system is provided in a side stream
of a circulating fuel circuit. The fuel is purged
of xenon and krypton by mixture with a helium
purge gas. The purge gas is circulated through
cooled charcoal beds of sufficient size to absorb
the xenon and krypton, and the helium is recircu-
lated to provide a continuous purge.

The circulating fuel, which is pumped from the
top of the reactor, is divided into four streams
that lead to four primary heat exchangers. The
four streams return to the reactor core and enter

tangentially at the bottom to provide a swirling
motion to the fuel as it rises in the reactor core.

Heat is interchanged between the fuel and the
steam in two sodium circuits in series, the first
being radioactive because of delayed neutrons
and the second being nonradioactive. For this
reactor system a fuel cost of 2.5 mills/kwh and
a total power cost of 8.9 mills/kwh have been
estimated.

The pertinent features of the interim design
reactor are given in Table 1.1.1.

REACTOR CORE CONFIGURATION
J. Y. Estabrook W. S. Harris

The shape of the core of the interim design
reactor is somewhat arbitrary; however, certain
criteria were adhered to in its definition. First,
the volume of fuel in the core was to at least
equal, but not greatly exceed, that of an 8-ft-dia
sphere, that is, 268 ft3. Second, the shapes of
sections taken through the walls of the core were
to be expressed as simple algebraic equations in
order to assure smooth shapes, easily calculated
volumes, and dimensional reproducibility. The
fuel inlet to the core was to consist of four 10-in.
sched-40 pipes bent slightly to impart a rotational
component to the fuel as it progressed through the
core. The four inlet pipes determined the minimum
diameter of the bottom of the reactor. The fuel
exit was fo consist of an 18-in.~dia pump inlet
sufficiently removed from the main body of the
core to make the fuel surrounding the pump
definitely separate from the critical mass of the
reactor.

Between the bottom and the large central
spherical portion of the core, a conical shape was
introduced, as shown in Fig. 1.1.1. The included
angle of the cone (60 deg) is probably the most
arbitrary part of the core shape and it will remain
so until hydraulic experiments are performed. A
large angle tended to minimize the fuel inventory,
but it may bring about intolerable flow separation.

The curve joining the spherical portion of the
core with the pump inlet was determined by
sketching in a curve, noting its point of tangency
(40 in. above the center of the sphere) and then
calculating the slope at that point (1.62088).
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 286364

 

 

 

 

 

 

 

\ !
‘ BLANKET

SECTION A-A \ PUMP
FUEL PUMP MOT‘OR
MOTOR ‘

 

 

 

 

   
 
    

 

BLANKET EXPANSION
TANK

 

 

 

 

SIPHON DRAIN

A=T | LN
FUEL LINE TO | 1~

HEAT EXCHANGER | IJI’ : |

 

FUEL EXPANSION
TANK

 

 

 

 

Fig. 1.1.1. Reactor Vessel and Pump Assembly.
Tuble ].]ll.

PERIOD ENDING JUNE 30, 1958

Reactor Plant Characteristics

 

Fuel

Fuel carrier
Neutron energy
Moderator
Primary coolant

Power
Electric (net)
Heat

Regeneration ratio
Clean (initial)
Average (20 years)

Blanket

Estimated costs
Total
Capital

Electric
Refueling cycle at full power
Shielding
Control
Plant efficiency
Exit fuel temperature

Steam
Temperature

Pressure
Second loop fluid
Third loop fluid

Structural materials
Fuel circuit
Secondary loop
Tertiary loop
Steam boiler

Steam superheater and reheater

Active=core dimensions
Fuel equivalent diameter
Blanket thickness

Temperature coefficient, (Ak/k)/°F
Specific power
Power density

Fuel inventory
Initial {(clean)

Average (20 years)
Critical mass (clean)

Burnup

>90% U235F4 (initially)

62 mole % LiF, 37 mole % Ber, 1 mole % ThF4
Intermediate

LiF-BeF2

Fuel solution circulating at 23,800 gpm

260 Mw
640 Mw

0.63
™~ 0.53

71 mole % LiF, 16 mole % Ber, 13 mole % ThF4

$69,800,000
$269/kw
8.88 mills/kwh

Semicontinuous

Concrete room walls, 9 ft thick
Temperature and fuel concentration
40:6%

1210°F at approximately 83 psia

1000°F with 1000°F reheat
1800 psia

Sodium

Sodium

INOR-8

Type 316 stainless steel
5% Cr, 1% Si steel

2.5% Cr, 1% Mo steel

5% Cr, 1% Si steel

8 ft
2 ft

—~(3.8 T 0.04) x 103
~ 1000 kw/kg
80 kw/liter

600 kg of U233
~ 900 kg of U23S

267 kg of U233

Unlimited

 
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

The curve had the appearance of part of an ellipse
with its center about 2 in. above the entrance to
the pump and perhaps 30 to 50 in. off to one side
of the core centerline. The general equations of
an ellipse having the ordinates shown in Fig. 1.1.2
were used to determine the distance () off the
core centerline:

b2x2? 4 aly? ~ 2ba2y +h2a? ~g2p2 -0

d b2
7% 1.62088 .
dx aZ(y___b)

By substituting the values of x and y shown in the
sketch into the general equations, it was possible
to solve for b without determining the coefficients
a and b; b was found to be 39.091. The determi-
nation of a and & was then made by simply
transferring the origin so that it was at the center
of the ellipse. The general equations were then

x2 y2
—_t =1,
a? b2

dy b2 x
a2y

The calculations gave a = 34.15 and & = 30.14.
When the result was checked for fit with the
spherical portion of the core, it was found that a
better match was obtained by slightly changing

UNCLASSIFIED
ORNL-LR—-DWG 31190

SLOPE AT THIS POINT =1.62088

¥ \
h 24.678 in.

»

 

 

 

— 2 in —-——————— 28 in.—————————»

 

Fig. 1.1.2,
with Pump Inlet.

Curve Joining Spherical Portion of Core

b to 39.078. The mismatch in slope at the point of
tangency with the sphere was only about 0° 17,
which is well within manufacturing tolerance for

shapes of this size.

WEIGHT ANALYSIS OF REACTOR SYSTEM
J. Y. Estabrook W. S. Harris

The interim design reactor was used as a basis
for preliminary rough estimates of the weights of
the various components of such a system.
mation on the weights is pertinent to remote
handling and maintenance problems, plant layout,
and construction planning and costs. For these
estimates it was assumed that sodium would be
used as an intermediate heat exchange medium
between the fuel and steam and that boilers of the
Lo&ffler type would be used. The data obtained
are presented in Table 1.1.2.

Infor=

EVALUATION OF FABRICABILITY
OF REACTOR VESSEL

E. J. Breeding

Fabricability was a major consideration in the
layout of the interim design reactor. Since the
first reactor does not need to be designed for mass
production, it was not considered to be essential
to maintain interchangeability in vessel parts.
it will be permissible for the fabricator to fit the
parts, within reasonable limitations, to the
tolerances he can expect with his equipment and
methods. This implies that the entire reactor
vessel should be fabricated in one shop adequately
equipped to accomplish the job. If parts were to be
fabricated by various suppliers, the dimensional
tolerances would have to be somewhat closer to
assure fitup upon assembly.

An arrangement of shapes that appear to be
fabricable according to conventional practice of
the pressure-vessel industry is shown in Fig. 1.1.3.
Comments on the numbered sections are presented
below.

Outer Blanket Shell

Section 1 is a conventional flanged section
designed as a support to which the fuel pump
assembly can be bolted.

Section 2 is a transition ring to provide a satis-
factory welding arrangement and stress distribution
in the neck region. This section can be machined

as part of section 1 or can be welded to section 1.
PERIOD ENDING JUNE 30, 1958

Table 1.1.2, Estimates of Weights of Components of a Molten-Salt Reactor System

 

 

Weight (Ib)
Empty equipment
Reactor vessel, including fuel and blanket salt pumps and motors 111,100
Fuel-toesodium heat exchangers (4) 62,300
Blanket salt-to=sodium heat exchanger 5,400
Fuel system superheaters (4) 279,000
Blanket system superheater 23,000
Fuel system reheaters (4) 50,000
Fuel piping 18,000
Blanket salt piping 4,000
Sodium piping 52,000
Sodium pumps and motors 80,000
Total 684,000
Fuel inventory
in reactor, including fuel in pump and expansion spaces (325 £+ at 122 !b/ffa) 39,650
In piping (290 #3) 35,380
In heat exchangers (224 ff3) 27,328
Total 102,358
Blanket salt inventory
In reactor, pump, and expansion space (920 13 at 201 1b/Ft3) 184,920
In heat exchanger (29 #3) 5,829
In piping (20 ft3) 4,020
Total 194,769
Sodium inventory
In fuel-to-sodium heat exchangers (160 #3 at 49.5 Ib/f‘ra) 7,920
In blanket salt-sodium heat exchanger (26 ff3) 1,287
In superheaters (768 ff3) 38,016
in reheaters (240 fts) 11,880
In piping (830 ff3) 41,085
Total 100,188

 

Section 3 is a 2:1 ASME elliptical shell head
with the center section removed. It is welded to
section 2.

Section 4 is a cylinder rolled from plate and
joined with one longitudinal weld.

Section 5 is a standard hemispherical shell head
with a flued opening.

Section 6 is a hemispherical shell head cut and
flared to mate with section 7.

Section 7 is a tapered conical section rolled from
plate and joined with one longitudinal weld.

Section 8 is a 2:1 elliptical shell with a flued
nozzle for joining with the blanket salt inlet pipe.

Inner Core Shell

Section 9 is a conical, dished section with a
flued opening.

Section 10, which is the pump barrel housing, can
be either a straight section of pipe or a cylinder
rolled from plate and longitudinally welded.

Section 11 is an elliptical shell head.
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL—LR—DWG 31191

BLANKET PUMP AREA

   

FUEL PUMP AREA

 

é secTion (1) g
L S LinmomT— P
e 2
///‘ 777”_" | @77 77 ‘ | ///// \:\‘_\----—::-".'.'.‘ L. Ttttz é

I

  
  

R

 

 

 

 

 

 

 

CORE SHELL

%,
/x ——BLANKET SHELL

 

 

';’//,/ e tT LoDl - — A
2 /
////////////////I//,/// ////
) Y
7 0

Fig. 1.1.3, Molten-Salt Reactor Yessel, See text for discussion of fabricability of numbered sections.
Section 12 is a special shape with reverse flues.

Section 13 is a cylindrical section rolled from
plate or made from straight pipe.

Section 14 is a plate with reverse flues.

Section 15 is a forged neck or a machined plate
weldment.

Section 16 is a hemispherical shell head with a
flared opening.

Section 17 is a hemispherical shell head cut and
flared to mate with section 18.

Section 18 is a tapered conical section rolled
from plate and longitudinally welded.

Section 19 is a tapered conical section rolled
from plate and longitudinally welded. The fuel
inlet pipe sections, the thermal sleeves, and
the outer shell section 7 will be a separate
subassembly.

Section 20 is an elliptical shell head.

Blanket System Pump Housing

Section 21 is a standard flanged section combined
with a nozzle section of straight pipe or a standard
flange machined on a cylindrical section.

Section 22 is a standard elliptical head with a
flued nozzle opening.

Section 23 is a cylindrical rolled-plate section
with the pump-discharge thermal sleeve welded in
as in section 19,

Section 24 is an elliptical shell head.

Section 25 is a special shape fabricated to form
a transition section between the vessels.

FUEL PUMP DESIGN

W. G. Cobb
The fuel pump proposed for the interim design
reactor is shown in Fig. 1.1.4. Fuel is fed

directly from the reactor core into the volute,
which is suspended from the roof of the expansion
tank and is submerged in the fluid. The drive
motor is completely canned and is located in @
clean portion of the purge gas system. The motor
and pump rotating assembly are completely
replaceable remotely as a cartridge unit. Fuel
salt and blanket salt flow through and around
structural parts for high-temperature cooling.
Radiation shielding of the motor is provided by
solid shield materials located below and around

PERIOD ENDING JUNE 30, 1958

the motor compartment. Temperature control of the
motor and shield materials is accomplished by
circulation of a low-temperature coolant through
passages lining the motor compartment. The pump
and motor rotating parts are mounted on a common
shaft that is supported by a salt-lubricated
orifice-compensated hydrostatic lower bearing,
which uses the impeller suction shroud as a
journal, and by a compound thrust and radial
bearing located immediately below the motor rotor
and supported on the cooled thermal barrier.

The upper bearing is a hemispherical orifice-
compensated hydrostatic type that uses the
pressurized helium purge supply for lubrication
and support. An approximate bearing configuration
is shown in Fig. 1.1.5. Thrust capacity is
provided by the lower continuous circumferential
pocket. Gas is admitted through several orificed
ports into the feed grooves at the upper and lower
pocket extremities to provide stability. Radial
load capacity is provided by the multiple pockets
surrounding the upper regions of the hemispherical
journal. Each pocket has an orifice-compensated
supply. The feed grooves in the thrust pocket and
the radial pockets have depths of approximately
3 mils. The thrust-pocket depth would probably
not exceed 1 mil. Instability interaction of the
thrust and radial forces is eliminated by using a
continuous circumferential bleed groove which
exhausts to the motor cavity. It is expected that
the high gas velocity through the inner portion of
the thrust bearing will prevent diffusion of the
primary fission products into the motor cavity.
The elimination of liquid lubricants avoids the
radiation damage and system contamination
problems encountered with their usage. Approxi-
mate gas conditions for bearing operation are
given below:

Gas supply pressure 50 psia
Motor cavity pressure 1642 psia
Shaft annulus pressure 19.2 psia
Thrust 1000 [b
Radial thrust 1000 Ib
Shaft ennulus flow 1495 scfm
Total thrust bearing flow 640 scfm
Totel radial bearing flow 5¢5 scfm
Total flow to motor cavity 9455 scfm
Total gas flow 11.5 scfm
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
F-2-02-054-7083A

FLOW RATE, 25,000 gpm

HEAD, 71 ft

SHIELD RETAINER 7 | SHAFT SPEED, 700 rpm

SEALING HEAD — / POWER REQUIRED, 1000 hp
/

HANDLING LUG

  
 
  
  
 
  
 
 
   
  
 
 
 
 
 
   
  

ELECTRICAL CONNECTION

|
LOW-TEMPERATURE MOTOR COILS

COOLANT CONNECTION
[l sHiELD Lock

HELICAL COOLANT PASSAGES—— ' ROTOR
GAS CIRCULATION ANNULUS LAMINATIONS |

e E
STATOR—- e

¥ 1 HYDROSTATIC '

GAS CONNECTION T O BEARING I

NUCLEAR RADIATION SHIELD ‘

GAMMA SHIELD

SUPPORT RING
NEUTRON SHIELD——

HIGH-TEMPERATURE
COOLANT JACKET —

PURGE GAS FLOW
MAXIMUM SALT LEVEL s

 

N
HIGH-TEMPERATURE ¥ 7
COOLANT PASSAGE -

BYPASS FLOW CHANNEL

SYSTEM EXPANSION VOLUME SALT

NORMAL OPERATING SALT LEVEL  FLOW
VOLUTE (DISCHARGE TO REAR) | \2

e

HYDROSTATIC SALT BEARING INLET
IMPELLER— SCALE Q 1 2 3 FEET

 

Fig. 1.1.4. Fuel Pump for Interim Design Reactor,

10
0.0015-in,

CLEARANCE 45.75-in. RADIUS

PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED
ORNL-LR-DWG 34192

 

 

   

 

 

 

GAS SUPPLY ||
PLENUM -—T
0.0017-in. CLEARANCE
SHAFT ANNULUS FLOW
Figc 1.]-50

Bearing of Pump Shown in Fig. 1.1.4.

NUCLEAR CALCULATIONS

Modifications of Oracle Program Sorghum for
Calculational Analyses of MoltensSalt Reactors

L. G. Alexander

The Oracle program Sorghum, a 31-group, zero-
dimensional, reactor ‘‘burn-out’* code, was
extensively modified as required for analyses of
molten-salt reactors.  Provision for automatic
insertion of the thorium cross sections corre-
sponding to the partial saturation of the resonances
was made, with the option available in the input
format. Similarly, a fuel option in the input was
added to provide a choice between U233 and U235
as the isotope whose concentration is adjusted to
bring the reactor to the critical condition. The
output edit was modified to give the following:
time in years; concentration of key fuel, in
atoms /cm?3; total inventory of fissionable material,
in kg (including U233 in blanket system); integrated
net burnup, in kg (initial inventory plus integral
of feed rate less total inventory at time T);

|
0" \
|

RADIAL BEARING
( POCKET / BLEED SLOT
. / 7

 

 

 

 

 

 

 

 

THRUST POCKET

Hemispherical, OrificesCompensated, Gas-Lubricated Bearing Designed To Be Used as Upper

inventories of all nuclear species in core and
blanket, in kg; fraction of neutrons causing fission
in each of the three fissionable isotopes (U233,
U233, Py?39); neutron absorption
fourteen species; average n; average v; regeneration
ratio; and neutron balance. These items are
punched at each edit. A subroutine was also
provided which edits simultaneously, via the
console typewriter, the time, critical concentration
of fuel, total inventory of fissionable material,
regeneration ratio, and balance. A routine that
provides for automatically beginning the core
processing cycle at the end of the initial period
(usually one year) was aiso added.

Several provisions for detecting defective
problems were made, including the console type-out
just described, a subroutine to dump the memory
on drive 1 periodically so that no more than 10 min
of computing time is lost in event of machine
error, and iteration overflow that stops the
computation in cases where the critical calculation
does not converge in 31 trials. The modified
program appears to be working satisfactorily.

ratios for
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Analyses of Reactors Fueled with U233
L. G. Alexander

The investigation of the initial-state nuclear
characteristics of two-region, homogeneous, molten-
fluoride-salt reactors fueled with U232 was con-
tinved. The data required for comparison of
reactors fueled with U233 or U235 qgnd having
thorium concentrations in the range from 0 to 1
mole % ThF, in the fuel salt were obtained, and
the results are plotted in Fig. 1.1.6. The
comparison is based on reactors fueled with U233
or U235 having a basic core salt containing
31 mole % BeF, and 69 mole % LiF and a blanket
salt containing 25 mole % ThF, and 75 mole % LiF.
The regeneration ratios ronge up to 0.95 at
inventories of less than 350 kg of U233 for g
600-Mw system having external fuel volyme of
339 ft3,

In the interim design reactor described above,
the fuel salt composition was 37 mole % BeF2—63
mole % LiF plus UF, and ThF,, and the blanket
salt had a composition of 13 mole % ThF ,-16
mole % BeF 5=71 mole % LiF. The performance of
reactors fueled with U233 and having these fuel
and blanket salts is shown in Fig. 1.1.7. The

UNCLASSIFIEDR
ORNL-LR-DWG 31193

 

  

 

 

 

 

 

 

 

 

 

 

 

 

1.2
TOTAL POWER. 600 Mw
EXTERNAL FUEL VOLUME: 339 ft°
FUEL SALT: 31mole % BeF, + 69mole % L|F+ UF, + ThF,
1.0 |- BLANKET SALT: 25mole % ThF, + 75mole % LiF
0.8 ’
O
5 35
@ u?> ENVELOPE N
g |
= 0.8
&
o
l
=
w
B
€ o4 ThF, IN FUEL SALT
{mole %)
o 0
A 0.25
o8 ; o 0.50
A 0.75
. 4
[ NOTE: CORE DIAMETERS INCREASE
TO THE LEFT |
0 | ]
0 100 200 300 400 500 600
CRITICAL INVENTORY (kg)
Fig. 1.1.6, Initial-State Nuclear Performance of

Two-Region, Molten<Fluoride«Salt Re-
actors Fueled with U233 Compared with Reactors
Fueled with U233,

Homogeneous,

12

UNCLASSIFIED
ORNL-LR-DWG 31184

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1.4
| L] ]
| f
- TOTAL POWER: 600 Mw (HEAT) i
EXTERNAL FUEL VOLUME: 339 3 ‘ '
1.2 |~ FUEL SALT: 37 mole % BeF, + 63 mole % ’——‘“
LiF + UF, + ThF4 :
BLANKET SALT: 13male % ThF, + 16 mole % *‘“; T
\ o | |
o ‘ | BeF, + 7t male % LIF-'I | ]
3 ‘ 1 87 3
2 | P
r 08 ‘ 7 ¥ %
=z | 6 ‘ \ | \ N |
2 | 8,
g ‘ > \ w \
o |
> L Bfi ‘ | L
Z 06 | |
Q ‘ l , | I
u! i ; 1 |
= — NUMBERS ON CURVES ARE Q%H*_J
CORE DIAMETER IN FEET ,
0.4 4‘ i { | ‘
| 4 ThF, IN FUEL SALT | —
i ‘ } {mole %) |
0.2 P A 0.25 J‘_ |
A 0.50 |
| L o 0.75 |L L
} . { 5 l \
0 [ 1 1 1 | ‘
0O 20 40 60 80 100 120 140 160 180 200
CRITICAL INVENTORY (kg OF U233
Fig. 1.1.7. Initial-State Nuclear Performance of

Two-Region, Homogeneous, Molten-Fluoride-Salt Re-

actors Fueled with U233,

regeneration ratios range up to 0.9 at critical
inventories of U233 that are about one-fifth the
critical inventories required for reactors fueled
with U235, It was clear from the trend of the
curves in Fig. 1.1.7 that the performance would be
improved by increasing the thorium concentration,
and therefore three reactors with 4 mole % ThF,
in the fuel salt were studied. The results,
presented in Table 1.1.3, give regeneration ratios
that slightly exceed 1.0. |t seems reasonable that
a further siight increase could be obtained by
adjusting the diameter and the thorium concen-
tration.

The long-term performance of the UZ233.fyeled
with an 8-ft-dia core was studied with
the use of the Sorghum code. An extract from
the results is shown in Fig. 1.1.8. As may be
seen, the regeneration ratio fell rapidly from the
initial value to about 0.96 during the first year
of operation, while the inventory rose about 200 kg
from the initial 560 kg. At this time processing
of the core fluid was started at a rate sufficient

reactor
Tflble 1.]-3-

PERIOD ENDING JUNE 30, 1958

Initial-State Nuclear Characteristics of Two-Region, Homogeneous, Molten-Fluoride-Salt

Reactors Fueled with U233 and Having 4 Mole % ThF4 in Fuel Salt

Fuel salt: 37 mole % BeF2 + 63 mole % LiF + UF4 + T!"lF4
Blanket salt: 13 mole % ThF4 + 16 mole % BeF2 + 71 mole % LiF

Total power: 600 Mw (heat)
External fuel volume: 339 f43

 

Core diameter, ft 6 8 12
U233 iy fuel salt, mole % 0.324 0.226 0.140
U233 gtom density* 12.14 8.48 5.27
Critical mass, kg of U233 149 246 479
Critical inventory, kg of y233 595 559 869
Neutron absorption ratios**

U233 (fissions) 0.8715 0:8660 0.8785

U233 (4,) 0.1285 0.1340 0.1215

Be-Li«F in fuel salt 0.0887 0.0999 0.1274

Core vessel

Li«F in blanket salt 0.0856 0.0616 0.0320

Outer vessel

Leakage

Th in fuel salt 0.6409 0.7734 0.9228

Th in blanket salt 03725 0.2517 0.1306
Neutron yield, i 2.19 2.21
Regeneration ratio 1.013 1.025 1.038

 

*Atoms (X 10! 9)/cma.

**Neutrons absorbed per neutron absorbed in u233,

to hold the concentration of fission fragments
constant. The regeneration ratio was approximately
stabilized, but the inventory continued to creep
up to about 950 kg in 15 years. Meanwhile, the
integrated net burnup, that is, total purchases less
inventory, had reached 250 kg. The net burnup
rate averaged about 16 kg per year.

Analyses of Reactors Fueled with Plutonium

D. Baxter

Two cases of two-region, homogeneous, molten-
fluoride-salt reactors fueled with plutonium were

studied. The cores contained no thorium and were
6 and 8 ft in diameter, respectively. The results
of the analyses are presented in Table 1.1.4.
In comparison with corresponding UZ2335-fueled
reactors, the critical inventories of these reactors
are about one-third as great, and the required
plutonium concentration appears to be well below
the solubility limit in the salt mixtures of interest.
The regeneration in the blanket was substantial;
it amounted to about 40% of the fuel burned. The
U233 formed in the blanket should be effective
in compensating for ingrowth of fission products.

13
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 31195

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

1.10 |
CORE DIAMETER: 8ft
FUEL SALT: 37 mole % BeF,+ 63mole %
1.05 FLiF+UF, +TnF,
o TOTAL POWER: 600 Mw {HEAT)
%r_: PLANT LOAD FACTOR: 0.8
= EXTERNAL FUEL VOLUME: 339 ft3
o
£ 1.00 '
@
w
2
w
o
w
@
\
\
0.95 —4 —
rCORE PROCESSING BEGINS
0.90 ’
1000 I
/——
WENT
\
800 ‘ “(Ofli'/
/r/
600 I
B N
g |+ CORE PROCESSING BEGINS
ND I
400 t
| o L
200 auRNY
[ <0 Net
A
0
0 2 4 6 8 10 12 14 16
TIME (years)
Fig. 1.1.8. Long-Term Nuclear Performance of Two-
Region, Homogeneous, Molten-Fluoride-Salt Reactors

Fueled with U233 and Having 4 Mole % ThF4 in Fuel
Salt,

14

ARGON AS A PROTECTIVE ATMOSPHERE
FOR MOLTEN SALTS

L. A. Mann

Helium has been used almost exclusively at
ORNL as the protective or ‘‘blanket’ gas for
molten salts in experiments in both radiocactive
and nonradioactive environments. Argon has been
used occasionally, and other gases have been
considered as possible alternates for use with

particular designs, materials, and operating
conditions (for example, krypton, xenon, neon,
hydrogen, nitrogen, carbon dioxide)., =~ Several

factors must be considered in selecting the cover
gas. The major considerations are, of course,
availability and cost of gas of the required purity.
In establishing the specifications for the gas, it is
necessary to study the effects of the gas on the
container and other contacted materials, as well as
the effect of the gas on the salt and the salt on
the gas. If the gas is to be used in a radiocactive
environment, determinations must be made of the
activity that will be induced in the gas and the
effect the activity will have on the system. For
molten-salt reactor use, the requirement of ability
to strip krypton and xenon from the fuel is an
added factor.

The scarcity and high cost factors immediately
eliminate krypton, and xenon from con-
sideration for use in large systems. The nonnoble
gases can be evaluated only in terms of specific
applications. The possible chemical activity of
the nonnoble gases requires that a determination
be made of the compatibility of the gas with all
materials contacted under the proposed operating
conditions.

neon,

A review of the various factors has indicated
that only helium and argon are suitable for molten-
salt reactor application. Cylinder argon costs,
at present, about 6.5 cents per standard cubic
foot, compared with about 4 cents for cylinder
helium. Pipe line helium costs about 2.5 cents,
but it is anticipated that the price will increase
in the near future. Helium has the disadvantage
that no considerable sources of helium have been
found to exist outside the United States. Calcu-
lations have indicated that the activity produced
in the A40 isotope of natural argon by neutron
PERIOD ENDING JUNE 30, 1958

Table 1.1.4. Initiol-State Nuclear Characteristics of Two-Region, Homogeneous,

Molten-Fluoride Reactors Fueled with Plutonium

Total power: 600 Mw (heat)
External fuel volume: 339 3

Fuel salt: 37 mole % BeF2 + 63 mole % LiF + PuF3
Blanket salt: 13 mole % ThF4 + 16 mole % BeF2 + 71 mole % LiF

 

 

Core diameter, ft 6 8

Thorium in fuel salt, mole % 0 0

Pu in fuel salt, mole % 0.045 0.013

Pu atomic density* 1.55 0.460

Critical mass, kg of Pu 19.7 13.7

Critical inventory, kg of Pu 78.1 31.1

Neutron absorption ratios**
Pu (fissions) 0.6043 0.6291
Pu {n,y) 0.3957 0.3709
Li-Be-F in fuel salt 0.1404 0.3093
Core vessel 0.1282 0.1459
Li-Be-F in blanket salt 0.0266 0.0233
Outer vessel 0.0047 0.0034
Leakage 0.0070 0.0033
Th in blanket 0.4625 0.3530

Neutron yield, 1 1.76 1.84

Regeneration ratio 0.463 0.35

*Atoms (X 10-]9)/cm3.
**Neutrons absorbed per neutron absorbed in y233,
capture is negligible compared with that of the Some of the materials being considered for

krypton and xenon that will escape from the fuel
in a molten-salt reactor.

BEARING TESTER DESIGNS
L. V. Wilson
Hydrodynamic Bearing Tester

A test rig was designed for experimental studies
of hydrodynamic bearings and bearing materials in
molten salts. The apparatus consists primarily
of a model PK centrifugal pump which was
modified so that a journal could be mounted on
the shaft at the normal impeller position, as shown
in Fig. 1.1.9. The bearing is flexibly supported,
and radial force may be applied to it through the
load column attached to the load beam. The force
for the bearing is applied to the load beam by an
air cylinder mounted on the spool piece. The
bearing torque is measured by changes in power
input to the motor that drives the pump shaft.

bearings have coefficients of thermal expansion
that are approximately one-third the coefficient
of INOR-8. Since a stress problem will exist
when the bearing assembly is raised to operating
temperatures, a compensating-column type of
journal mounting is being investigated. In the
design illustrated in Fig. 1.1.9, the journal is
mounted on eight radial load columns made of the
same material as the journal. These load columns
are located near the centerline of the shaft and
near the axial center of the bearing to keep the
differential expansion to a low enough value to be
absorbed by the elastic bending of the load
columns.  The bearing load on the journal is
transmitted to the shaft by the columnar reaction
of the load columns.

Results of operating of this test rig are described
in Chapter 1.2 of this report.
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

LOAD BEAM—— |

PIVOT

 

BELLOWS —

 

AN

 

LOAD COLUMN

Fig. ]-]ogu

16

 

® > \!_/ .

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
ORNL-LR-DWG 31186
B:T
If
!1
| !
| /
"LI‘
/
[
}
!
/
“.‘
! .*':
{ ( /
[ = /
_ i
! E 7
b L =
| .
| ! ‘
i i
B
=
=
> =

 

 

 

 

 

== SUPPORT PINS

 

 

L
7 v
N N
-

NI e
BEAR!NG/ L g JOURNAL

 

TO FILL AND DRAIN TANK

Apparatus for Experimental Studies of Hydrodynamic Bearings,
Hydrostatic RotatingsPocket Bearing Tester

A rig designed for water tests of a hydrostatic
orifice-compensated bearing is shown in Fig. 1.1.10.
The bearing pockets and orifices are located on
the rotating member and the pressure supply is
provided by the centrifugal pumping action of the
supply channels to the orifices. The rotating
element is relatively rigid, and the stationary
element, which is mounted on two ball bearings,

  

TACHOMETER

BEARING POCKET

  
  

PRESSURE TAF’/
AIR CYLINDER
IMPELLER

PERIOD ENDING JUNE 30, 1958

is free to move in the direction of the load, which
is applied by means of an air cylinder. Motion of
the stationary element relative to the rotary
element is measured by dial indicators. The
stationary element is a Plexiglas shell, reinforced
with cutout steel pipe, to permit observation of
beoring operation. Pressure taps to measure
pressure distribution are located in the Plexiglas
around the periphery of the bearing.

UNCLASSIFIED
ORNL-LR-DWG 31197

  
  
  

COOLING
WATER

i
J U

  

COOLING COIL

  
   
   
 
 
  
 

DIAL
INDICATCR

  
 

  

Fig. 1.1.10. Apparatus for Experimental Studies of Hydrostatic Rotating-Pocket Bearings,
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL—LR—DWG 31198

  
  
   
     
   
    
    
 

COOLANT QUTLET

COOLANT INLET

LEAKAGE CUTLET

GAS INLET

PRESSURE TAPS __—0ORIFICE

.____—JOURNAL

PUMP TANK

BEARING
POCKETS

BEARING
BYPASS
FLOW

VOLUTE

IMPELLER

Fig. 1.1.11. Apparatus for Experimental Studies of Hydrostatic Stationary-Pocket Beorings.

18

Bk by e s S P nma e Al S et
Hydrostatic StationarysPocket Bearing Tester

A rig designed for testing hydrostatic stationary-
pocket bearings in water and then in molten salts
is shown in Fig. 1.1.11. A model PK pump was
modified for this application by removing the lower
ball bearing and seal, replacing them with a
labyrinth seal, and mounting the journal of the
hydrostatic bearing on the impeller hub. The lower

bearing load is carried by the hydrostatic bearing,

PERIOD ENDING JUNE 30, 1958

which is supplied with high-pressure fluid from
around the outer periphery of the impeller. The
fluid flows from the impeller to the bearing
orifices, into the bearing pockets, and out to the
expansion tank. The load on the bearing is
produced by the radial hydraulic unbalance in the
volute acting on the impeller. The bearing load
and the direction of the load are measured by the
pressures in the bearing pockets and their
resultant vector.

19
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

1.2. COMPONENT DEVELOPMENT AND TESTING

H. W. Savage

W. B. McDonald

Reactor Projects Division

FUEL PUMP DEVELOPMENT
W. F. Boudreau A. G. Grindell

Development Tests of Salt-Lybricated Bearings

P. G. Smith W. E. Thomas
L. V. Wilson

Hydrodynamic Bearings. — Construction of equip-
ment for testing hydrodynamic bearings in salt
mixtures was completed, and cold and hot mechani-
cal shakedown tests were performed (see Chap. 1.1
for description of equipment design). The first
test was run to obtain a preliminary evaluation of
tolerable bearing loads, and the bearing consisted
of an INOR-8 journal and sleeve, since this material
was known to have excellent corrosion resistance
to salt mixtures. While dry, the bearing was
subjected to one thermal cycle from room tempera-
ture to 1200°F and return. Heating required 26 hr
and cooling 12 hr, and the journal was rotated in
the bearing one-quarter turn each hour by hand.

An attempt was then made to operate the bearing
in fuel salt 130 (LiF-BeF ,-UF,, 62-37-1 mole %)
at 1200°F with a shaft speed of 1200 rpm. The
bearing operated for about 20 min ot an applied
load of up to 100 Ib. When the load was increased
to 125 |b, the power consumption of the drive
motor increased rapidly, and within a few additional
minutes the load became sufficient to stall the
The test was interrupted at this point, and
the shaft and bearing assembly were removed for
examination. Results of the examination are not
yet available. A second similar bearing, modified
to assure better filling, is operating satisfactorily.

Tests are planned in which fuel 130 will be used
as the lubricant and some of the operations ex-
pected for the fuel pump in a molten-salt reactor
will be simulated in order to obtain bearing per-
formance data for comparison with the results of
calculations based on theoretical considerations.
The calculations related bearing clearance and
minimum film thickness to bearing load, speed,
and salt viscosity and were the basis for selecting
testing clearances of 0.003 and 0.005 in. between
radial journal and sleeve. Calculations of the

motor.

20

journal power required and of Sommerfeld’s number
were also made for bearings having either of the
two clearances specified. The INOR-8 bearing
tested had a radial clearance of 0.005 in.

Other design and experimental investigations are
under way to determine @ suitable means for
mounting refractory metal or cermet bearing journals
and sleeves (molybdenum and tungsten carbide
with 12% cobalt binder) to the basic structural
material. The design must take into account the
differences in thermal expansion between the
materials, and imposition of stresses sufficient
to distort the bearing material must be avoided.

Hydrostatic Bearings. — Calculations have been
made to determine load capacity, eccentricity,
bearing liquid flow, bearing pressure distribution,
journal speed, orifice size, bearing clearance, and
supply pressure for tests of hydrostatic bearings.
Fabrication of the test equipment (described in
Chap. 1.1) is nearly complete. Preliminary water
tests will be performed to obtain the bearing load
and pressure distribution with respect to test loop

flow, flow resistance, pump speed, and power.

Rotating-Pocket Hydrostatic Bearings. — The
test apparatus described in Chap. 1.1 for evaluating
the rotating pocket hydrostatic bearing,! with
water as a lubricant, was completed. The test
bearing consists of an aluminum journal 11.351 in.
in diameter and 2.5 in. in height. There are twelve
pockets, 0.090 in. deep, equally spaced around the
periphery, with all lands, both peripheral and
axial, 0.375 in. in length. The bearing is con-
structed from Plexiglas pipe, bored to 11.375 in. ID,
to give a radial clearance of 0.012 in. between
journal and bearing. Each pocket on the journdl
is supplied with pressurized lubricant by one vane
of the impeller. Each vane consists of a 3/4-in.-dic1
drilled hole and takes its inlet from the 4-in.-dia
impeller eye. The test-bearing impeller is designed
to deliver a 50-ft head (with water) at the orifice
face when rotating at 1300 rpm.

 

IB. W. Kinyon, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL.-2474, p 16.
Development Tests of Conventional Bearings
D. L. Gray W. E. Thomas

Organic-Liquid-Lubricated Bearings. — The high-
temperature sump pump being used to conduct an
experimental evaluation of the behavior of Dow-
therm A {eutectic mixture of diphenyl and diphenyl
oxide) as a lubricant has continued to operate
satisfactorily for a period of 1620 hr. The pump
is circulating fuel 30 (NaF-ZrF4-UF4, 50-46-4
mole %) at a temperature of 1200°F; the shaft
speed is about 2600 rpm; the temperature of the
Dowtherm A supplied to the bearings is maintained
at 180°F, with approximately a 5°F rise through
the test bearings.

Upon completion of 1000 hr of operation, the
pump was disassembled and the test bearings
were inspected. The double-row angular-contact
ball bearings that were submerged in the lubricant
during operation showed no detectable wear or
imperfections.  The diameters of the aluminum
bearing and the Inconel journal were measured and
the diametral clearance was found to have increased
by 0.001 to 0.0045 in., as measured at room
temperature. Since 0.0045 in. is the design
diametral clearance, this initial wear may be
attributed to ‘‘wear in"' rather than to poor lubri-
cation.

It was also noted that the O-rings (buna N
material) had increased in size upon exposure to
Dowtherm A, but the elasticity of the material had
not been affected greatly. No leakage was observed
from the O-rings during operation.

Bearing and Seal Gamma Irradiation. — By the
end of the quarter the pump rotary assembly being
operated in the canal at the MTR as a bearing test
under gamma irradiation had accumulated a gamma-
ray dose of 5.9 x 10° r. Since the bulk of the
Gulfcrest 34 bearing lubricant is external to the
radiation field, the oil has accumulated a gamma-
ray dose of about 10% r. Samples of oil totaling
931 ¢m?, collected at various time intervals from
the lower-seal-leakage catch basin and from the
bulk oil supply, as well as four different helium
samples, have been received from the MTR for
analyses. The rotary element is operating under
the following conditions: shaft speed, 4000 rpm;
total lubricating oil flow, 4 gpm; temperature of
oil inlet to bearing housing, 143°F; lower seal
helium purge flow, 60 liters/day; upper seal
helium purge flow, 500 liters/day.

PERIOD ENDING JUNE 30, 1958

The barrier heater used to simulate gamma
heating in the lower seal area completely shorted
out on March 3. The oil temperature level dropped
only 5 to 7°F, and the loss of the heater was not
considered sufficient cause for terminating the
fest.

For seven days during March, the test assembly
operated without gamma irradiation while necessary
maintenance was performed on other MTR canal
tests. A power failure in April stopped test
operation for approximately 20 min. During this
power failure, some 750 cm® of oil leaked from the
lower seal oil purge line before the helium purge
flow was regained.

Gas-Lubricated Bearing Studies
D. L. Gray

Gas-lubricated bearings are being considered
for use in molten-salt pumps. In one design,2 the
gas bearing is hemispherical. A study of the
literature has indicated that this type of bearing
is probably feasible, but it is known that gas-
lubricated bearings require very small running
clearances and are subject to self-induced vibra-
tions. Calculations have indicated that a thrust
bearing of 8.5-in. spherical radius, supported with
helium, as proposed for a 5000-gpm molten-salt
pump, will require a minimum supply pressure of
45 to 50 psia. A gas flow of 5 scfm will be re-
quired to maintain @ minimum clearance of 0.001 in.
in this bearing. For a 25,000-gpm pump, a thrust
bearing of 15.75 in. spherical radius that is de-
signed for a constant clearance between journal
and bearing at equilibrium conditions will require
a 5-scfm helium flow into the pump tank end a
4-scfm flow into the motor cavity, at a supply
pressure of 40 psia, in order to provide a minimum
clearance of 0.001 in. A small hemispherical
bearing approximately 8 in. in diameter will be
used to test the validity of these calculations.

Mechanical Seals
D. L. Gray
Labyrinth and Split-Purge Arrangement. — Oper-

ation was terminated of the NaK pump which was
modified, as previously described,? to include a

 

2W. G. Cobb and M. E. Lackey, MSR Quar. Prog. Rep.
Jan. 31, 1958, ORNL-2474, p 18.

3p. G. Smith and L. V. Wilson, MSR Quar. Prog. Rep.
Jan. 31, 1958, ORNL-2474, p 18.

2]
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

labyrinth in the pump shaft annulus between the
seal region and the pump bowl and a purge gas
inlet midway of its length so that a portion of the
helium purge gas moved upward in the shaft
annulus past the lower seal region and into the
catch basin and the remainder of the purge gas
moved down the shaft annulus and into the gas
volume of the pump tank. The pump had operated
for 3386 hr, but, during the last three weeks of
the test, it was observed that the pump tank
pressure responded slowly to changes in gas
supply pressure. [t was deduced that the small
holes drilled through the heat barrier shaft sleeve
to the labyrinth had become constricted. Efforts
to clear the passages were unsuccessful, and the
test was terminated when communication from gas
supply to the pump tank became impractical.

A post-test inspection of the pump rotary assembly
revealed that the lower surface of the heat barrier
was covered with NaK and oxides. Brown- and
black-colored material had practically filled the
gas supply holes which fed the labyrinth, circum-
ferential, V-groove plenum. This material was
later determined by chemical analysis to contain
approximately 7% carbon, and thus it was evident
that lubricating oil or graphite had played a part
in the formation of the constriction.

Approximately 0.064 in. of the lower seal graphite
nose piece had been worn away during the 3386 hr
of service. Both face surfaces of the lower seal
were free of NaK and, although worn, appeared to
be in excellent operating condition.

Bellows-Mounted Seal. — By the end of the
quarter, the bellows-mounted seal being subjected
to an endurance test in a NaK pump had accumu-
lated more than 5100 hr of operation. This modified
Fulton-Sylphon type of seal is being tested without
gas purge protection from the vapors of the NaK
being pumped. The main helical spring was removed
from the standard seal and replaced with 24 small-
diameter helical springs equispaced in a circle
around the bellows and concentric with it. There
are no elastomers in this seal, and therefore it
may prove to be suitable for a pump that is to
operate in a radiation field. As mentioned previ-
ously,? the seal leakage during the first 2000 hr
of operation averaged approximately 25 cm®/day,
but it then decreased to approximately 8 cm®/week

 

4s. M. DeCamp and W. E. Thomas, MSR Quar. Prog.
Rep. fan. 31, 1958, ORNL-2474, p 20.

22

and has remained at that rate despite two shut-
downs required to correct deficiencies in parts
unrelated to the test.

Radiation-Resistant Electric Motors for Use
at High Temperatures

S. M. DeCamp

The successful development of molten-salt-
lubricated bearings would permit consideration of
totally submerged canned-rotor pumps for molten-
salt service. Such a pump would require an electric
drive motor that would operate reliably at high
temperatures and would be resistant to damage by
A preliminary study has indicated the
need for high permeability magnet steels suitable
for 1250°F operation, current-carrying conductors
of acceptable strength and electrical conductivity
at 1250°F, and electrical insulation for use between
the conductors ond the magnet iron or steel that
would not be damaged by high temperatures.

An investigation of electrical insulating materials
was initiated. At present, ceramics and glasses
appear to be the most promising materials for this
application. Several wire manufacturers are de-
veloping insulated wire for missile electronic
components, and some of their products may be
suitable for the motor being considered.

A contract has been given to the Louis Allis
Company for the production of six test coils suit-
able for use at 1200°F. These coils will be insu-
lated with a coating of metallic oxides that appears
to have good radiation and thermal stability. The
first test coils, which are expected in September,
will be tested at the desired service temperature,
and if they have satisfactory thermal stability they
will be tested for nuclear radiation stability.

Work performed by others in the development of
motors for use at high temperatures has indicated
that loss of strength and high resistance of the
conductor will be problems at the contemplated
operating temperature of 1200°F. Nickel and copper
alloys are being considered for use as conductors.
At 1200°F, the Curie point of common electrical
steels is approached, and it may be necessary to
use cobalt steels.

radiation.

Piping Preheating Tests
D. L. Gray

Tests were made to determine the length of time
required to bring a typical fused-salt piping system
to a temperature (900°F) substantially above the
salt freezing temperature by preheating a portion
of a closed forced-circulation test loop with
electric-furnace elements and transmitting the heat
to the remainder of the loop by forced circulation
of helium within the system piping.

The preheating tests were made in a loop that
included a pump, piping, a venturi, and throttling
valves. About 1200 |b of Inconel was used in
constructing the loop. Forty-nine per cent of the
total loop surface was subjected to radiant heating
from the flat, ceramic-covered, radiant-wire heating
elements. The loop helium pressures and pump
speeds used in three tests are given below:

Loop Pressure Pump Speed
(psig) (rpm)
14 3475
98 3450
10 0

 

Fig. 1.2.1, Mechanical Joint Test Loop.

PERIOD ENDING JUNE 30, 1958

The required temperature of 900°F was attained
in a period of approximately 9 hr in the test for
which a system pressure of 98 psig was used. It
was not possible to reach the desired temperature
in either of the other tests.

DEVELOPMENT OF TECHNIQUES FOR REMOTE
MAINTENANCE OF THE REACTOR SYSTEM

E. Storto W. B. McDonald

Mechanical Joint Development
A. S. Olson

Further screening tests were conducted on the
three types of mechanical joints, described previ-
ously,® that are being developed to facilitate
remote removal and replacement of reactor com-
ponents. The loop shown in Fig. 1.2.1, which

 

5A. S. Olson, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 20.

M UNCLASSIFIED
PHOTO 31474

(1) Joint being tested, (2) Calrod heaters and shim stock binding,

(3) clamshell heaters, (4) loop frame, (5) molten-salt pump, (6) salt sump, (7) insulation support trough, (8) insulation.

23

 
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

was used for these tests, is capable of circulating

molten salts at temperatures up to 1500°F, and

controlled thermal cycling conditions can be

imposed. An assembled joint is shown in Fig. 1.2.2.

The joints are subjected to the following typical

test procedure:

1. joint assembled and leak checked while cold
(leakage specification, <1 x 10~7 c¢m® of helium
per second),

2. joint welded into loop and loop operated through
50 thermal cycles between 1100 and 1300°F,

3. loop drained and joint removed intact,

4. joint leak checked,

5. joint separated, remade, and leak checked.

None of the joints fested have given any indi-
cation of leakage during operation, and the leakage
rates found before and after the tests have been
satisfactory. Three joints that were photographed
upon separation after testing are shown in Figs.

1.2.3, 1.2.4, ond 1.2.5.

The freeze-flange joint shown in Fig. 1.2.3 was
separated without difficulty, the backup seal ring
was easily removed, and the frozen salt was
readily cleaned from the flange faces. The in-
dented-seal flange joint shown in Fig. 1.2.4 was
also separated without difficulty, but the gasket
stuck to one-half the joint and had to be machined

 

Fig. 1.2.30

sleeve, (3) groove for backup seal ring.

24

 

Freeze-Flange Joint Separated After Testing.

UNCLASSIFIED
PHOTO 31519

o
¥

 

Fig. 1.2.2.
After Removal from Test Loop.

Assembled Indented-Seal Flange Joint

UNCLASSIFIED |
PHOTO 31525 |

(1) Frozen salt between flange faces, (2) insert

 
o — - e — e e T e e 1 = =

PERIOD ENDING JUNE 30, 1958

 

UNCLASSIFIED
PHOTO 31132

 

—I T l ¥ l ¥ ' ¥ l I ' ¥ l T I v l_' l i l T | v ‘
O 1 2 3
INCHES

 

 

Fig. 1.2.4, Indented-Seal Flange Joint Separated After Testing.

UNCLASSIFIED
PHOTO 31423
Ew R

 

Fig. 1.2.5. Cast-Metal-Sealed Flange Joint Separated After Testing.

25
e e e, S e e ——

MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

from the flange face. The cast-metal-sealed flange
joint (Fig. 1.2.5) was heated to above the melting
point of the sealing metal in order to effect sepa-
ration, and heavy oxidation of the sealing metal
resulted.  Considerable force was required to
separate the flanges, and it was impossible to
remake a tight joint without reworking both sealing
faces.

The freeze-flange and indented-seal joints are
being reassembled for testing with sodium as the
circulated fluid, since sodium may be used as the
secondary coolant for a molten-salt reactor system.
All parts of the freeze-flange joint are being re-
used. A new gasket will be used in the reassembled
indented-seal flange joint.

A new freeze-flange joint for use in 4-in. pipe
has been fabricated and assembled, as shown in
Fig. 1.2.6. Details of the design of this joint may
be seen in Fig. 1.2.7. The 4-in. joint will be
installed in an existing salt-circulating loop for
testing. Provision has been made for continuous
leak monitoring of the joint during high-temperature
operation of the loop.

Remote Manipulation Techniques
C. K. McGlothlan

Experimental remote maintenance work on a NaK
pump which had been used in a high-temperature

UNCLASSIFIED
PHOTO 31405

 

Fig. 1.2.6. New Freeze-Flange Joint for 4-in. Pipe.

26

 
PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED
ORNL-LR-DWG 31199

GAP (~v g in.)

 
 
 
 

NICKEL-PLATED SOFT COPPER

CLAMP - BACKUP SEAL RING

    

   
   
   
 
   
    
  
  
    
   
    
    

 
 
  
  

AIR CHANNEL FOR COOLING

&
@@

o
“o

REINFORCING WEB

  

RING INSERT TO PROVIDE
LABYRINTH FOR SALT LEAKAGE

TN

 

/ |

e ———
T SALT STREAM -

 

 

Lh______/
WELD OF FLANGE
TO LCOP (4-in. PIPE)

 
 

THIN SECTION TO PROVIDE
TEMPERATURE GRADIENT FOR
FORMATION OF FROZEN SALT SEAL

&

SCREEN INSERT IN FREEZE ANNULUS TO
REDUCE SALT HOLDUP VOLUME AND
PROVIDE BASE FOR FROZEN SALT CAKE

PUMP OUT TUBE FOR
LOCATION QF GAS LEAK OR

TO BUFFER GAS LEAKAGE CLAMP LUGS

Fig. 1.2,7. Design Details of Joint Shown in Fig. 1.2,6,

27
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

system was completed as previously planned.‘s
The results of this work’ showed that remote
maintenance is definitely feasible for a pump or
any other readily accessible component in a fused-
salt system.

Good vision is essential to the performance of
maintenance operations by remote manipulation,
and direct viewing through cell windows, indirect
viewing with periscopes, and closed-circuit tele-
vision viewing have been considered for this
application. The use of television is contingent
on the availability of radiation-resistant cameras
and the development of stereo viewing. A review
of work in progress at ORNL and in industry has

given reasonable assurance that video cameras

 

éc. K. McGlothlan, MSR Quar. Prog. Rep. Jan. 31,
1958, ORNL-2474, p 24.

7w. B. McDonald, C. K. McGlothlan, and E. Storto,
Remote Maintenance Experimental Work on a Reactor
System Pump, ORNL CF-58-4-93 (April 23, 1958).

¢
%

=

can be constructed that will operate satisfactorily
for extended periods in gamma radiation of up to
1 x 10° r/hr. Since no good stereo television
system was found to be readily available, an
experimental optical stereo system was designed
and constructed. Stereo pair photographs were
used to simulate television monitor screens acti-
vated by a pair of television cameras remotely
located. The three-dimensional viewing achieved
with this system was very satisfactory, and there-
fore specifications were issued for the purchase
of a matched pair of closed-circuit television
systems, the monitors of which will be substituted
for the stereo pair photographs.

A study of the feasibility of remote welding by
mechanical manipulation was initiated. A pair of
Argonne model 8 manipulator arms was set up to
handle a standard Heliarc welding torch and filler
rod, as shown in Figs. 1.2.8 and 1.2.9. Two
welders succeeded in producing high quality welds

& UNCLASSIFIED
| PHOTO 31477
St e

 

 

Fig. 1.2.8. Remote Welding of Plate.

28
 

PERIOD ENDING JUNE 30, 1958

UNCL ASSIFIED
PHOTO 31479

 

Fig. 1.2.9. Remote Welding of Pipe.

with this apparatus on standard Inconel plate and
pipe test coupons. The welds passed x-ray and
dye-penetrant tests. A specimen of pipe opened
by a leak was also repaired. Both welders reported
that they had no difficulty in becoming accustomed
to the apparatus (Figs. 1.2.10 and 1.2.11), but they
both stated that inadequate viewing was a severe
handicap. The best results have been achieved by
viewing through binoculars, but this method results
in severe eyestrain. Experiments are now being
conducted on the development of camera lens
filtering and special illumination to permit remote
viewing of the welding process with television.

Remote Maintenance Demonstration Facility

An engineering layout has been approved for a
remote maintenance demonstration facility (Fig.
1.2.12), and detailing and fabrication of the com-
ponents are proceeding. An order has been placed
for a General Mills manipulator, and specifications

for a closed-circuit television viewing system were
issued, as stated above. An area has been cleared
for the facility in the experimental engineering
building (Building 9201-3), and design of the

manipulator runway is essentially complete.

Heater-Insulation Unit Development

Pilot model heater-insulation units® (Fig. 1.2.13)
were fabricated and are to be tested on a loop
circulating molten salts. The units are constructed
on the clamshell principle and are designed to be
applied to pipe sections and removed from them
by remote manipulation. Each unit will serve to
preheat its pipe section preparatory to filling of
the system and will insulate it during high-temper-
ature operation. Joints between units are covered
with clamshell inserts that contain only insulation.

 

8A. L. Southern, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 25.

29
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

| |

_ =

g

 

Fig. 1.2.11. Setup for Television Viewing of Remote Welding.

30

— e

UNCL ASSIFIED

| PHOTO 31475

HUNCL ASSIFIED
PHOTO 31474

 

 
) UNCLASSIFIED
< ORNL-LR-DWG. 31482

 

Fig. 1.2.12. Remote Maintenance Demonstration Fa-

cility.

EVALUATION OF EXPANSION JOINTS FOR
MOLTEN.SALT REACTOR SYSTEMS

R. L. Senn

A test program was initiated to evaluate expansion
joints that are commercially available, with respect
to their usefulness in molten-salt reactor systems.
If expansion jeints can be used in fuel and coolant
circuit piping, the extra space and fluid inventories
required for thermal expansion loops can be avoided.

Six expansion joints were ordered from three
different vendors. Each vendor will supply one
Inconel and one stainless steel unit. The Inconel
joints are to be tested with the molten salts of
current interest, and the stainless steel joints are
to be tested with sodium.

A test stand was designed and is presently
being built which will provide for simultaneous
testing of three expansion joints. Each joint will
by cycled through its maximum axial traverse once
every 2 hr for 1000 cycles. The cycle period is
sufficient to allow relaxation of stresses introduced
into the bellows by each half cycle. The joint will
be filled with the appropriate fluid and held at
1300°F and 75 psi during the test.

PERIOD ENDING JUNE 30, 1958

UNCL ASSIFIED
QRNL-LR-DWG, 31559

 

Fig. 1.2.13. Pilot<Model Heater-Insulation Unit.

HEAT TRANSFER COEFFICIENT

MEASUREMENT
J. C. Amos R. E. MacPherson
R. L. Senn

Processing of the data obtained from a heat
transfer coefficient test in which fuel 130 was
circulated in an available heat exchanger test
facility wos completed and a preliminary report
Descriptions of the test facility and
procedures were presented previ-
ously.'® Data obtained during an initial calibration
run for fuel 30 (NaF-ZrF4-UF4, 50-46-4 mole %),
as well as the data obtained with fuel 130 flowing
on the inside of the round heat exchanger tubes,
are compared in Fig. 1.2.14 with (1) the Dittus-
Boelter equation,

0.4 0.8
Nyo/Npt =0.023N 0

was issved.’
experimental

 

9

J.. C Amos, R. E. MacPherson, and R. L. Senn,
Preliminary Report of Fused Salt Mixture No. 130 Heat
]'.I;'ggjsfer Coefficient Test, ORNL CF-58-4-23 (April 2,

mJ. C. Amos, R. E. MacPherson, and R, L. Senn,
MSR Quar. Prog. Rep. Jan, 31, 1958, ORNL-2474, p 27.

31
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

32

0.4
Pr

NNU/N

UNCLASSIFIED

ORNL~LR-DWG 27859R
100

20
80
70

60

50

40

30 DITTUS—BOELTER EQ
N 0.4 _50p3 408
nu/Npy =0.023 Vg
FOR TURBULENT FLOW
OF NORMAL FLUIDS

20

MIXTURE 12 IN STAINLESS STEEL
FOR NICKEL  SOURCE: HOFFMAN

MIXTURE 30 IN HEAT EXCHANGER SHELL

O MIXTURE 30 INSIDE TUBES

 

A MIXTURE 430 INSIDE TUBES

400 600 800 1000 2000 4000 6000 8000 40,000
REYNOLDS NUMBER, g,

Fig. 1.2.14, Correlation of Results of Molten-Salt Heat Transfer Coefficient Measurements,
for turbulent flow of normal fluids in ducts; (2) an
experimental curve for molten salt mixture 12
(NaF-KF-LiF, 11.5-42-46.5 mole %) flowing inside
round nickel or stainless steel tubes; and (3) an
experimental curve for fuel 30 flowing on the shell
side of heat exchangers similar to the heat ex-
changer utilized in this test. As may be noted,
the data for flow inside the tubes are in general
agreement with the data for mixture 12 and for
fuel 30 flowing on the shell side above a Reynoids
number of 4000, but they are lower than the shell
side data at Reynolds numbers below 4000. This
difference at the lower Reynolds numbers has been
attributed to improved shell-side heat transfer at
Reynolds numbers normally associated with laminar
and early transitional flow as a result of the
turbulence promoted by the wire spacers ufilized
in the closely spaced heat exchanger tube bundle.

It was necessary during the course of this test
to formulate, on the basis of experimental data, an
equation for predicting the heat transfer per-
formance of the NaK flowing in the shell side of

the test heat exchanger. The resulting equation,

0.6
Ny, = 0.066(N, Ny )

!

is compared on the basis of the test heat exchanger
dimensions other empirical heat
transfer equations in Fig. 1.2.15. As may be seen,
there is good agreement with an equation developed
by Tidball'! for NaK flowing in baffled-shell heat
exchangers.

A molten-salt-to-air radiator has been designed
and is currently being fabricated locally to replace
the NaK-to-air radiator in the heat exchanger test
facility used to obtain the information described
above. Modifications have been started on the test
facility to accommodate this radiator and thus
provide for a molten-salt-to-molten-salt heat transfer
test.

with various

DESIGN, CONSTRUCTION, AND OPERATION OF
MATERIALS TESTING LOOPS

Forced-Circulation Loops
J. L. Crowley

Operation of forced-circulation corrosion testing
loops was continued. Repair work was completed

 

”C. B. Jackson (ed.), Liguid Metals Handbook.
Sodium-NaK Supplement, TID-5277, p 285 (July 1, 1955).

PERIOD ENDING JUNE 30, 1958

on three loops, and, by the end of March, nine
loops were operating ~ six that were fabricated of
Inconel and three of INOR-8,

On April 6, a circuit breaker failed, and the main
resistance-heating power supply was lost to six of
the nine operating loops. The salts in these loops
froze, and, when heat was reapplied, four of the
loops ruptured in the heated section because of the
excessive stress created by the expanding salt.
Two of the loops successfully resumed operation.

Two of the four ruptured loops have been repaired
and restarted, one was disassembled for exami-
nation, and one is yet to be repaired and restarted.
Samples of tubing were removed from the loops for
metallurgicol examination before repairs were made.
A summary of operating experience as of May 31,
1958, is presented in Table 1.2.1.

The bifluid loop, designated CPR, partially froze
just after it had passed a full year of accumulated
operating time because the salt-pump drive motor
failed as a result of an overload. The lack of
driving force on the centrifugal pump then caused
a seal failure. When the pump was removed to
replace the seal, a metallic coating was discovered
on the impeller and the associated liner assembly.
The coating is believed to have contributed to the
overload of the motor by causing interference. A
sample of the material was removed for analysis.

Two new loops were started. One is a standard
Inconel loop (designated 9377-3 in Table 1.2.1)
that is circulating fuel 131. The second is a
special-purpose INOR-8 [oop that contains graphite
specimens and is circulating fuel 130. This loop
(designated 9354-5 in Table 1.2.1) was described
in the previous report. 2

Another special-purpose loop (9354-4) for making
weight-loss studies of INOR-8 sample inserts was
completed and will be placed in operation when the
required fuel 130 charge is available. The samples
will be removed at staggered intervals, as described
previously, '2 in order to establish a corrosion rate
which can be extrapolated to several years of
operation.

As a result of the incident which caused the four
loop failures of April 6, improvement of test stands
to assure operational reliability has been accel-
erated. Adequate automatic protection of the salt
system against failures resulting from trouble with

 

12J. L. Crowley, MSR OQuar.

Prog. Rep. Jan. 31,
1958, OCRNL-2474, p 31.

33
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL—-LR—-DOWG 27866R

 

-

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0
e T
Neo ,
& ==, 0025 UVRIe P = T
- [ =
(1 N | |~
o- k
5 ¢ fi—f—
4 J ’
3
(1) FOR NaK INSIDE TUBES, MARTINELLI AND LYON
» REF. 11, p 73
(@ FOR LIQUID METAL INSIDE TUBES, KAUFMAN AND
3 LUBARSKY, NACA-TN-3336
= (3) FOR NORMAL FLUIDS IN BAFFLED SHELLS, DONOHUE,
. REF. 11, p 285
o (4) EXPERIMENTAL DATA FOR NaK IN SHELL
2 (5) FOR NaK IN BAFFLED SHELL, TIDBALL, REF. 11, p 285
2 (8) FOR NaK IN UNBAFFLED SHELL, MCGOFF, REF. 11, p 285
Z
= 0.9 ,_‘; VALUES USED IN EQUATIONS
T 0.8 0" 4 1 Np, = PRANDTL NUMBER = 0.006
o 0.7 ’4 Nwe { | D, = EQUIVALENT DIAMETER = 0.1284 in.
2 1® L J 4, = FLOW AREA = 0.00428 2
0.6 1 A, = HEAT TRANSFER AREA =~ 6.73 ##2
1.8
0.5 — 1 <fi) = 0.000001764
Ah
0.4 —I»l 1l I
03 o | |
8 |
N\N‘\ &
\
0.2 Net =
Nae
20 ~
z° A
@ N /
o4 L i
10,000 20,000 50,000 100,000 200,000 500,000 1,000,000
REYNOLDS NUMBER, Ag,
Fig. 1.2.15, Comparison of Various NaK Heat Transfer Equations.
conventional components has required mechanical In-Pile Loops
fznd elect'nccl redesign of the |oc'>ps'and changes D. B. Trauger
in operating procedures. The major improvements 3. A. Conlin P. A. Gnadt

to be incorporated in new loops and test focilities
are shown schematically in Fig. 1.2.16. A loop
that incorporates all the improved features has
been fabricated and is shown in Fig. 1.2.17 on its
portable dolly with one half of the cocler box
removed. Changes of operating loops are being
made as they are shut down for disassembly or
for repairs.

34

Out-of-pile tests were conducted which demon-
strated adequate heat-removal capacity of the salt-
to-air heat exchanger designed for use with the
in-pile loop that is to be operated in the MTR.'3
The temperature region near the salt freezing point

 

I3D. B. Trauger, J. A, Conlin, and P. A. Gnadt, MSR
Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 33.
GE

Table 1.2.1.

Forced-Circulation Loop Operations Summary as of May 31, 1958

 

 

Composition Approximate Maximum Minimum Maximum Hours of
Loop LOOP Number of Flow Reynolds Wall Fluid Fluid Operation at
Designation Moter!a| Circulated Rate Number Temperature Temperature Temperature Corditions Comments
and Size Floid*  ©Pm) a1 V100°F (°F) (°F) CF) Given
CPR Inconel, 3/8 in. sched 40, 122 1 5,000 1250 1095 1190 8801 Repairs being made on pump impeller
1in. OD x 0.035 in. wall, Sodium ~7 97,700 1085 1135
3, . .
7 in. 0D x 0.035 in. wall
B
9344.-1 Inconel, ]/2 in. CD, 123 2 3,250 1300 1100 1210 5125 Loop shut down for 90 hr to change motor
0.045 in. wall
9354-3 INOR-8 84 2.75 1200 1070 1150 3983 Loop froze once because of power loss;
Hot leg, % in. sched 40 4,500 thawed successtully
Cold leg, ]/2 in. OD, 5,400
0.045 in. wall
9344.2 Inconel, l/2 in, OD, 12 2.5 8,200 1200 1000 1100 2956 Loop froze because of power loss;
0.045 in. wall repaired and restarted
9377-1 Inconel, ]/2 in. OD, 126 2.0 1,600 1300 1100 1175 3413 Loop froze because of power loss; to be
0.045 in. wall repaired and restarted
9377-2 Inconel, I/2 in. OD, 130 2.0 3,000 1300 1100 1210 3055 Loop failed when frozen because of
0.045 in. wall power loss; terminated 4/6/58
9354-1 INOR-8, 1II/2 in. OD, 126 2.5 2,000 1300 1100 1210 2106 Hestelloy pump bowl replaced with
0.045 in. wall INOR-8 pump bowl; no incidents during
this period
9377-3 Inconel, ]/.z in. OD, 131 2.0 3,400 1300 1100 1215 1925 Loop froze once because of power loss;
0.045 in. wall thawed successfully
9354-2 INOR-8, 1/2 in. OD, 12 2.0 6,500 1200 1050 1140 1052 Hastelloy pump bowl replaced with
0.045 in. wall INOR-8 pump bowl and loop restarted
2/26/58; loop subsequently froze
because of power loss; failure occurred
during aftempt to restart
9354-5 INOR-8, 3/8 in. OD, 130 1.1 2,200 1300 1100 1205 501 Special loop containing graphite semples;
0.035 in, wall started 5/8/58
9354-4 INOR-8 130 Special loop containing INOR-8 inserts;
Hot leg, 3/8 in. sched 40 fuel salt not yet available
Cold leg, % in. OD,
0.045 in. wall

 

*Compesition

Composition

Composition 84:

Composition

122:
123:

Nal":-ZrFA-UF4 (57-42-1 mele %)
NaF-BeF2-UF4 {53-46-1 mele %)
Nc:F-LiF-BeF2 (27-35-38 mole %)
12:  NaF-LiF-KF {11.5-46.5-42 mole %)

Composition 126 LiF-BeF2-UF4 (53-46-1 mole %)
Composition 130: LiF-Ber-UF4 (62-37-1 mole %)
Composition 131: Li|:-BeF2-UF4 (60-36-4 mole %)

8561 ‘0€ INNI ONIANI goiy3d
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

EMERGENCY CONDITIONS WHICH WILL AUTOMATICALLY
PLACE LOOP ON ISOTHERMAL OR PRE HEAT CONDITIONS:

1., LOW TEMPERATURE.

. HIGH TEMPERATURE.

« LOSS OF TRANSFORMER POWER.

. LOSS OF BOTH CLUTCH SUPPLIES.

. LOSS OF PUM?P DRIVE MOTOR.

. COMPLETE POWER FAILURE (DIESEL GENERATOR STARTS).

. LOSS OF BUS DUCT SUPPLYING LOOP EQUIPMENT
(DIESEL GENERATOR STARTS).

~N O RN

 

MPROVEMENTS INCORPORATED [N REVISEL CORROSION LOOFP DESIGN:
DOUBLE SHEAVES ON PUMP DRIVE,
USE OF STEEL CORE ''v** BELTS.

MAINTAINED CONTACT ON PUMP MOTOR START BUTTON.
USE OF AUTOMATIC RESET ON TRANSFORMER POWER.

ELIMINATION OF SLOW RESPONSE HEATER ELEMENTS ON COOLER
COIL BY APPLYING DIRECT RESISTANCE HEAT AUTOMATICALLY 1IN
EVENT OF AN EMERGENCY CONDITION.

&, USE OF A LID ON CCOLER COIL DUCT WHICH DROPS AUTOMATIC-
ALLY IN THE EVENT OF AN EMERGENCY CONDITION,

@» «OO

UNCLASSIFIED
ORNL-LR-DWG 29715

LOOP CN ISOTHERMAL OR LOW HEAT CONDITIONS CONSISTS OF:

*A. HEATER LUGS "' AND "'D'* DISCONNECTED BY 1600 AMP
BREAKERTQ PROVIDE RESISTANCE HEAT THROUGH LUGS A"
AND "'C!' TO ENTIRE LOOP EXCEPT COCLER COIL.

B. RES|ISTANCE HEAT APPLIED DIRECTLY TO COOLER COIL BY
10 KVA POWER SUPPLY.

C. COOLING AIR OFF,
D. LID DROPPED ON COOLER DUCT TO ENCLOSE COOLER COIL.

*NOTE: ALTERNATE METHOD, IN THE EVENT CF FAILURE OF THE
110 KVA POWER SUPPLY, AUTOMATICALLY CONNECTS CLAM
SHELL HEATERS (ITEM 10) T PROVIDE SUFFICIENT HEAT TO PRE-
VENT FREEZE UP.

@ INSTALL A 1600 AMP CIRCUIT BREAKER TO AUTOMATICALLY SWITCH
TO LOW HEAT CONNECTIONS ON LOOP PIPING IN AN EMER-
GENCY.

8. IMPROVED OPERATION OF MAGNETIC CLUTCH SPEED ALARM BY USE
OF HIGHER QUALITY METER RELAY AND CIRCUIT CHANGES.

9. INCREASED FUNCTION OF LOW TEMPERATURE ALARM TO INCLUDE
AUTOMATIC ACTIONS FOR PREVENTION OF FREEZE UPS.

PROVISIONS MADE FORTHE USE OF CLAM SHELL HEAT AUTOMATIC~
ALLY IN THE EVENT OF A FAILURE OF MAIN 110 KVA TRANSFORMER
POWER.

(1) PROVISIONS MADE FOR USE OF EMERGENCY DIESEL GENERATOR, IN
THE EVENT OF A COMPLETE POWER FAILURE, FOR ISOTHERMAL OPER-
ATION OF CORROSION LOOPS,

Fig. 1.2.16. Improved Forced-Circulation Cortosion Testing Loop and Test Facilities.

was investigated extensively in these tests, since
low temperatures can be expected during a reactor
scram. |t was determined that a control system
can be designed which will permit the loop to
survive a reactor scram without danger of freezing
and without close operator supervision. The salt
in the heat exchanger was deliberately frozen and
remelted several times during the course of the
experiment.

36

Prototype pump tests revealed that gravity-filling
of the system will be difficult because of the
surface tension of fuel 130, and two other methods
of filling are being studied. One method is to fill
the pump sump with small increments of fuel. This
method is not compatible with filling the loop after
insertion in the reactor as was done previously
with other loops. Accordingly, tests are being run
with the prototype pump loop to determine the
 

PERIOD ENDING JUNE 30, 1958

URCU ASSIFIED .
PHOTO 31454

L
G

 

Fig. 1.2,17, New Forced-Circulation Loop.

feasibility of freezing and melting the loop. This
method would permit a complete loop checkout
prior to its shipment to NRTS for installation in
the MTR and would eliminate the uncertainties
associated with filling after complete assembly
when minor repairs could no longer be made. The
second method utilizes the gravity-filling system,
with the filling lines modified so that the pump
impeller and loop are filled before a pump sump

level is established. This method was demonstrated
successfully with water,

The in-pile loop assembly has been completed
to the point where the pump can be installed.
Further assembly work will be delayed, however,
until the new filling techniques are tested. All
parts have been fabricated and the shielding and
bulkhead sections are complete.

37
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

1.3. ENGINEERING RESEARCH
H. W. Hoffman

Reactor Projects Division

PHYSICAL PROPERTY MEASUREMENTS
W. D. Powers

Viscosity

The viscosities of three beryllium-containing
fluoride salt mixtures were experimentally de-
termined over the temperature range 500 to 900°C.
As previously described,’ the measurements were
made with the use of precalibrated efflux-cup
viscometers. The scatter of the data was above
average (110% in one case), and therefore the
data must be considered to be preliminary. Mean
viscosity values at three temperatures are given
in Table 1.3.1.

Some of the observed deviations of the data
might be explained in terms of the surface tension
of the beryllium salts. There is considerable
“humping’’ of the fluid surface, and thus the
determination of the cup efflux time is extremely
difficult. Further measurements of the three
mixtures listed in Table 1.3.1 and of other salts
are being made, and possible refinements of the
measuring technique are being studied.

Thermal Conductivity

A variable-gap apparatus is being fabricated
with which to measure the thermal conductivities
of beryllium-containing fluoride salt mixtures.
Contamination by the beryllium salt requires that
the apparatus be expendable, and, hence, it must
be relatively small and inexpensive. On the

 

]W. D. Powers, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 37,

other hand, a small system increases the un-
certainty in the measurement and control of the
heat flow through the test sample. It is believed
that the design chosen represents a satisfactory
compromise between the factors of cost, ease
of operation, and experimental accuracy. The
apparatus will consist of two solid stainless
steel cylinders, each 4 in. in diameter and 41/2 in.
long, arranged so that their vertical axes are
coincident. The lower cylinder will be an integral
part of the sample container, while the upper
cylinder will be supported and separated from
the lower cylinder by a set of support rings
attached to the sample container. These rings
will permit adjustment of the sample thickness
in 0.010-in. steps between 0 and 0.14 in. Heat
generated by cartridge heaters located within
the upper cylinder will flow downward through
the test sample to a heat sink located below
the lower cylinder. The magnitude of the heat
flow will be determined by calculation from the
measured temperature gradients in the two cyl-
inders.

Enthalpy and Heat Capacity

The measurement of the enthalpy and heat
capacity of the mixture LiF-BeF,-UF, (62-37-1
mole %) was completed. Values for these proper-
ties may be represented by the following equations:

Solid (100 to 300°C)
Hp ~ Hygog = ~9.6 + 0.293T + (2.18 x 10~7) T2

c, = 0.293 + (4.36 x 10~5) T

Table 1.3.1. Viscosities of Beryllium-Containing Fluoride Salt Mixtures

 

Salt (mole %)

 

Viscosity (cp)

 

 

LiF NaF BeF UF, At 600°C At 700°C At 800°C
53 46 ] 15.2 7.4 4,2

53 46 1 20.5 10.8 6.4

62 37 1 12.2 7.2 4.7

 

38
Liquid (471 to 790°C)
Hp ~ Hygoc = 24.2 + 0.4887 + (2.52 x 10~5) T2

c, = 0.488 + (5.04 x 10~ T

In the expressions above,
H = enthalpy in cal/g,

c, = heat capacity in cal/g-°C,
T = temperature in °C.

Surface Tension

Recent experiences in pumping a beryllium-con-
taining mixture through a small in-pile loop and
observafions made during viscosity measurements
generated interest in measurements of the surface
tensions of mixtures in this class of fluoride
salts. An apparatus is therefore being fabricated
with which to measure surface tensions by the
maximum-~bubble-pressure method. The maximum
pressure necessary to form o bubble at the end
of o small-diameter tube immersed in the salt
melt will be determined, and the surface tension
will then be calculated from the formula

5 r
y = p2 ’
where 8p is the pressure difference between the
inside and outside of the formed bubble and r is
the inside radius of the tube. The maximum-
bubble-pressure method was chosen for this in-
vestigation because (1) the experimental apparatus
is simple and presents no design and fabrication
difficulties, (2} instrumentation for measuring the
pressure is available and can be readily calibrated,
(3) the theory of the method is well developed
and on a sound basis, and (4) o new surface is
formed with each bubble so that, with a clean
tip on the bubbler, surface contamination should
not be a problem.
Preliminary studies made with water to provide
operafional experience gave the following results:

Diameter of Tube Surface Tension

(in.) (dynes/em)
0.026 68
0.032 70
0.046 72

The generally accepted value for the surface
tension of water is 72 dynes/cm at 24°C. The

PERIOD ENDING JUNE 30, 1958

difference between the accepted value and the
values is believed to be due to
in the measurement of the inside
An error of 0.001 in. would be
sufficient to account for the observed dis-
crepancies. Measurements with the salt mixture

NaNO,-NaNO,-KNO, (40-7-53 wt %) are currently

in progress.

experimental
inaccuracies
tube diameter.

Thermal Expansion

In evaluating the nuclear characteristics of a
reactor, it is necessary to know the quantity of
fuel present in the core at any time. For a homo-
geneous reactor, such as the MSR, the fuel
density, and hence the amount of fuel, is de-’
pendent on the core temperature. This relationship
between the core temperature and the fuel density
enters the nuclear calculations as the coefficient
of thermal expansion. Variations in the values
derived for this coefficient for a number of
beryllium-containing mixtures from density meas-
urements at ORNL and at the Mound Laboratory
have indicated the need for a direct measurement
of the thermal expansion of the salt mixture.

An apparatus has been designed and is being
fabricated with which to obtain the thermal
expansion by measuring the difference in height
of the fluid in the two legs of a U-tube when
the legs are maintained at different temperatures.
The liquid levels in the two legs will be de-
termined with adjustable metallic probes inserted
through the Lucite top of the inert-atmosphere
box containing the apparatus. The probes will
be hollow, with only the contact tip solid, since
the large thermal gradient which will exist along
the probe would make the length of a solid rod
both variable and indeterminate. Low-expansion
quartz rods will be inserted in the tubes; and
the movement of the upper ends of these rods,
as determined by fixed micrometers, will be used
to establish the liquid-level difference in the
U-tube. The initial ‘*zero’”” will be obtained by
maintaining the two legs at the same constant
temperature.

HYDRODYNAMIC STUDIES OF MSR CORE
F. E. Lynch

Hydrodynamic studies of three proposed molten-
salt reactor cores were continued with the use
of the phosphorescent-particle flow visualization
technique. The three cores being investigated

39
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

differ only in the associated entrance-exit systems
which provide (1) straight-through flow with
diametrically opposed entrance and exit (MSR-1),
(2) concentric entrance-exit flow with fluid entering
through the inner pipe and exiting through the
outer annulus (MSR-2), and (3) concentric entrance-
exit flow with fluid entering through the annulus
and exiting through the inner pipe (MSR-3). The
physical details of these models and the charac-
teristics of flow of the second type mentioned
above were described in the previous report.?

The results of full-field observations made with
a planar light source for the straight-through flow
model (MSR-1) are given in Fig. 1.3.1. The inlet
Reynolds modulus (based on the inside diameter
of the inlet pipe) was 74,000, which corresponds

to a flow rate of 1.38 ft/sec. It may be seen

 

2k, E. Lynch, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 35.

UNCL ASSIFIED
ORNL—LR--DWG 31200

T -1

/// | !
VELOCITY NEXT i
TO THE WALL 1S £
VERY LOW @

   

/PARTICLE MOTION
IN THIS REGION 1S
VERY UNSTABLE

Fig. 1.3.1, Flow Patterns Observed Within an lllumi-
nated Plane of the Straight-Through Core Flow Medel
{MSR-1).

40

that the main high-velocity flow short-circuited
the core and passed directly from the entrance
to the exit with essentially no spreading. The
remainder of the core was filled with slowly
rotating fluid that exhibited low-velocity flow
upward along the sphere surface and extremely
low-velocity flow near the core exit. This rotating
mass displayed some instability (perhaps as-
sociated with gross pressure fluctuations in the
system), with smaller eddies appearing as indi-
cated by the dashed lines of Fig. 1.3.1. In-
stantaneous velocity profiles for this same flow
are shown in Fig. 1.3.2 at the indicated axial
positions along the core. The flatness of the
velocity profile through the central jet is apparent
at all positions. The distortion of the exciting
beam and the profile, particularly noticeable in
(@), (b), and (f) of Fig. 1.3.2, resulted from the

curvature of the sphere.

Results were also obtained for flow in the third
type of model described above (MSR-3). A
photograph of the experimental system is shown
in Fig. 1.3.3. The central pipe was flared outward
at the top of the core so that the fluid entering
through the outer annulus was directed along
the sphere surface. For this system, the inlet
Reynolds modulus (based on the diameter of
the inlet side arm) was 70,000, which corresponds
to a flow rate of 1.3 ft/sec.

The gross velocity patterns obtained through
visual observation of particle motion within an
itluminated plane of MSR-3 are shown in Figs.
1.3.4 to 1.3.6. The complexity of the flow is
only partly indicated by these figures. The ob-
served patterns were generated, not by the annular
entrance per se, but rather by the manner in which
the fluid entered the annulus, as is evident in
Fig. 1.3.4, which shows that the fluid flowed
past the central pipe rather than parallel to it.
Lengthening the annular region or installing
turning vanes should correct this condition.
Such changes are to be studied in future tests.

The inset circle of Fig. 1.3.4 outlines the
nature of an off-center rotating eddy formed by
the asymmetrical inlet flow. When observed from
above (see Fig. 1.3.5), this eddy was found to
consist of two separate vortices, one located
120 deg from the inlet and the other slightly
less than 90 deg. Both tapered toward the center
and joined intermittently. The outer face of the
eddy did not contact the sphere surface. At
Ly

 

 

 

Fig. 1.3.2, Instantaneous-Velocity-Profile Photographs in Straight-Through Core Flow Model (MSR-1).

UNCLASSIFIED
ORNL-LR-DWG 31201

8561 '0f ANNI 9NIGNT @0i¥d3d
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Fig. 103130
Designed for Fluid Entrance Through Outer Annulus,

Concentric Entrance-Exit Model (MSR+3)

higher velocities, the vortices combined and the
eddy assumed a position perpendicular to the
inlet flow., Other aspects of the flow in this
geometry are shown in Fig. 1.3.6, in which the
plane of illumination includes the center of the
sphere. The inlet flow imparted a general rotation
to the entire fluid mass. Further, it was observed
that at the exit the flow split and some of the
fluid recirculated through the core. A number
of stagnant regions were noted. One was found
off the trailing edge of the inlet flare at a position
180 deg from the inlet where the incoming and
recirculating flows combined.  Another, found
within the inlet flare region, was due to bypassing
of the inlet flow. This stagnant region extended
around the annulus to 90 deg on each side of
the entrance. A third stagnant region was found
close to the sphere center in the region of the
eddy. As is also shown in Fig. 1.3.6, it was
found that the axis of the eddy rotated.

A series of instantaneous
photographs  obtained for this geometry are
presented in Figs. 1.3.7 and 1.3.8, which were

velocity profile

42

 

UNCLASSIFIED
ORNL-LR—DWG 3{202

 

STAGNANT
REGION — —t———
/ s — e
e o ——
/ -
- . FLOW N T
/ - — .
- -
-
g

 

 

 

   

      
  

\A—FLOW PATTERN
| WITHIN A ZONE
OF THE SPHERE

Fig. 1.3.4, Flow Patterns Within the Annular Entrance
and a Zone of MSR-3 As Observed in an llluminated

Plane.

UNCLASSIFIED
ORNL—LR—~DWG 34203

AXIS OF ROTATION COF

/ EQUATORIAL EDDY
!

CENTRAL
PIPE

 

OUTER LIP
OF FLARED
CENTRAL PIPE

 

INLET SIDE-ARM

AXIS OF ROTATION OF /
EQUATORIAL EDDY-

Fig. 1.3.5. Flow Patterns Within MSR-3 As Observed

from Above.
UNCLASSIFIED
ORNL-LR-DWG 31204

  
  
   
  
  

INTERMITTENT STAGNANT
REGION WHICH FLUCTUATED

AS INDICATED BY ARROWSj ~STAGNANT REGION

EXTENDED ARQUND
THE ANNULUS TO
30 deg ON EACH
SIDE OF INLET

-EDDY ROTATED ABOUT
AX[S SHOWN BY DASHED
LINES. THE AXIS CF
ROTATION OF THE INTER-
MITTENT STAGNANT
REGION WAS THE QUTER
EDGE OF THE EDDY.

Fig. 1.3.6. Flow Patterns Within MSR=3 As Visualized

in an llluminated Plane.

taken at positions 90 deg apart. The downward
flow along the sphere wall in the region opposite
the inlet is apparent in Fig. 1.3.8, while Fig.
1.3.7 gives evidence of the upward surface flow
in the region closer to the inlet. Further, Fig.
1.3.8, (&) through (d), shows clearly the eddy
described in Fig. 1.3.5.

Recent studies by Taylor? of flow in a sphere
with a central pipe entrance and an annular exit
have shown that the depth of penetration of
the inlet stream is a function of the extent of
projection of the central pipe into the sphere.
A bubble-photograph technique was used in these
studies., Taylor also investigated the effect
of flaring the inlet pipe. With no projection into
the sphere, it was found that the inlet stream
did not reach the bottom of the sphere. This
condition persisted until the projection extended
to one-fourth of an inlet-pipe diameter. Con-
siderable vibration was observed when flared
tubes were used. The amplitude of vibration
increased for tube projections of one-fourth
diameter or less and also varied with the ex-
pansion being greater with a 40-deg

3A. F. Taylor, Preliminary Studies of the Flow
Patterns in Two Designs of Homogeneous Test Reactor
Core. Part I. Re-entrant Core, IGR-TN/CA-324 (sub-
ref HARD(A)/P-12) (March 23, 1956).

angle,

 

PERIOD ENDING JUNE 30, 1958

included-angle flare than with a 20-deg flare.
Vibrations with the straight tube occurred only
for projections of less than one-fourth diameter.

The model shown in Fig. 1.3.9 (MSR-2A) was
fabricated in order to study the effect of extending
the central pipe in the MSR-2 model. The central
pipe was extended three-fourths of a pipe diameter
into the sphere (compared with one-fourth diameter
for MSR-2), and the flare at the top of the sphere
in the region of the join with the exit pipe was
increased.  Bubble photographs of the over-all
flow are shown in Fig. 1.3.10. The inlet Reynolds
modulus was 118,000, which corresponds to a
flow rate of 3.25 ft/sec in the inlet pipe. An
unsteadiness in the flow pattern may be seen
by comparing the two photographs of Fig. 1.3.10.
In Fig. 1.3.104, a well-defined eddy may be seen
in the right-hand region of the sphere, while the
left side shows a number of smaller eddies and
more random patterns. In Fig. 1.3.105, taken a
short time later, the patterns are reversed. |t
was also noted that the three-fourths-diameter
projection gave a higher velocity at the bottom
of the sphere than did the original one-fourth-
diameter projection, A series of instantaneous
velocity-profile photographs obtained with this
model at a Reynolds modulus of 108,000 (flow
rate of 3 ft/sec) are shown in the consecutive
exposure reproduced in Fig. 1.3.11. The points
of excitation are indicated by arrows,

MOLTEN-SALT HEAT TRANSFER STUDIES
H. W. Hoffman

A preliminary series of tests was completed
with the salt mixture LiF-Ber-UF4 (53-46-1
mole %) flowing through an electrical-resistance-
heated small-diameter Inconel tube.  Typical
results are shown in Fig. 1.3.12, where the data
are presented in terms of the Colburn j-factor.
It may be seen that the limited results closely
agree with the correlation, j = 0.023NES'2, for
ordinary fluids. Some difficulty was experienced
in establishing a satisfactory heat balance for the
system, and experiments are under way to diagnose
and correct this difficulty. An existing system
that contains a pump is to be modified so that
long-time exposure data can be obtained with this
salt mixture. Long-time exposure will be required
in order to determine what adverse conditions,
such as those which would arise as a result of
film formation, will exist in this system.

43
vy

(c)

  
 

Fig. 1.3.7.

 

 

 

&
CAMERA
POSITION
~
+ _—_

ANGLE OF EXCITATION

Instantaneous«Velocity-Profile Photographs in MSR-3.

(d)

UNCLASSIFIED
ORNL—LR—DWG 31205

 

1Y0dIY SSIYO0dd WYHO0dd ¥yOLIvIY LAVS-NILTIOW
)4

 

 

ANGLE OF
EXCITATION

(¢)

4

CAMERA
POSITION

Fig. 1.3.8, Instantaneous«Velocity-Profile Photographs

       

UNCLASSIFIED
ORNL—LR—DWG 31206

 

{6}

 

 

(d)

in MSR-3. Photographs taken 90 deg from those of Fig. 1.3.7.

 

2561 "0 INNT 9NIANT @ol¥3d
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
PHOTO 34310

 

Fig. 1.3.9. Modified MSR-2 Central-Pipe-Inlet and Annular-Exit Flow Model (MSR-2A).

46
Fig, 1.3.11, Consecutive
Profile Photographs of Flow in MSR-2A.

 

 

UNCLASSIFIED
PHOTC 44497

Instantaneous-Yelocity-

PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED
PHOTO 44496

'Bi

UNCLASSIFIED
ORNL—LR-—-DWG 31207

2 2t

Nep Mp,

r

©
o
O
o

  
  

T

0.002

—
|

COLBURN FACTOR, j

 

S ‘
Lo ‘
ooot b o v o 1.
2 x 103 5 10% ? 3x10%
REYNOLDS MODULUS, A

Fig. 1.3.12. Heat Transfer with LiF-Ber-UF4
(53-46-1 Meole %).

47
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

1.4. INSTRUMENTATION AND CONTROLS
H. J. Metz

Instrumentation and Controls Division

ENDURANCE AND STABILITY TESTS
OF SHEATHED THERMOCOUPLES

C. M, Burton
fnconel-sheathed, 0,250-in,-OD, Chromel-Alumel

thermocouples with magnesium oxide insulation
and hot-junction closure welds made by the
Heliarc welding process are being tested in order
to determine the expected accurate life of a
thermocouple in service and to determine the effect
of a liquid metal or a molten-salt environment on
the integrity of the closure weld. More than
10,000 hr have been accumulated in a test in which
sheathed thermocouples are immersed in sodium
held at 1500°F, The temperature of the sodium
bath is reduced once each week to 1300°F and to
1100°F for readings at these temperatures. Of
the 38 thermocouples included in this test, only
2 have failed because of weld closure deficiencies.
Although the temperature readings from the various
thermocouples being tested do not correspond
exactly to the reading from the calibrated standard
thermocouple as a result of individual variations
and of different positions in the pot of heated
sodium, drifts from the initial temperature readings

are within +0.75%.
RESISTANCE-TYPE FUEL LEVEL INDICATOR
R. F. Hyland

Test facilities were prepared for an investigation
of the suitability for molten-salt reactor service
of the resistance-type fuel level indicator. Two
Inconel elements of the type shown in Fig. 1.4.1
were fabricated for these tests without temperature
compensation. Tests of these elements in fuel 130

(LiF-Ber-UF4, 62-37-1 mole %) will provide the

 

INTERNAL STRUCTURE

data needed for design of temperature-compensated
INOR-8 elements and indications of the long-term
stability of the elements.,

SCANNING SWITCHES
A, M, Leppert

Design work is under way on modifications
required to improve the Delta switch manufactured
by the Detroit Controls Corp. for use in switching
low-level transducer signals. Switches of this
type are useful for monitoring transducer signals
for alarm purposes and for commutating, transmitting
over a pair of wires, and decommutating as many
as 80 signals. The modified units are to have
stainless steel housings, and the individual con-
tacts will be insulated with Ceramicite.,! The
present units have Bakelite bodies which hoid and
insulate the individual contacts. The modifications
of this mercury-jet type of switch are being made
to increase bearing life, reduce the mercury con-
tamination, and eliminate electrostatic charges
during operation,

PRESSURE TRANSDUCERS
J. W. Krewson

A test program was initiated for determining the
suitability of various types of pressure trans-
ducers for use in a reactor containment cell where
freedom from drift, errors, and failure would be
essential, The available instruments are being
tested under a single pressure condition in order
to determine their static drift behavior, and other
instruments are to be obtained. The relationship

between temperature and driftisalso being studied.

 

1 Ceramicite is manufactured by Consolidated Electro-
dynamics Corp.

UNCLASSIFIED
PHOTO 29725

 

Fig. 1.4.1, Resistance-Type Fuel Level Probe,

48
PERIOD ENDING JUNE 30, 1958

1.5. ADVANCED REACTOR DESIGN STUDIES
H. G. MacPherson

Reactor Projects Division

AN EXPERIMENTAL 5-Mw
THERMAL-CONVECTION REACTOR

J. Zasler

The history of reactor technology has indicated
that the development and demonstration of a
reactor concept requires the operation of o small
experimental reactor and a medium-sized pilot
plant, In the case of molten-salt power reactor
development, it appears that the simplest and most
reliable experimental reactor would be based on
thermal convection of the fuel. A design study
has indicated' that a 5-Mw thermal-convection
reactor would be large enough to provide experi-
mental data and yet be small enough to keep the

chief disadvantages of the thermal-convection
system - excessive fue! volume and large heat
exchangers — from being major factors. Further,
the 5-Mw thermal-convection reactor could be
converted to a 50-Mw pilot plant by adding a fuel
pump and increasing the capacity of the heat dump.
Proposed layouts for the 5-Mw reactor system
are shown in Figs. 1.5.1 and 1.5.2, and the
dimensions and operating conditions for 5- and
50-Mw service are given in Table 1,5.1. Provisions
would be included for connecting the blanket and
fuel regions so that the reactor could be operated

 

1], Zasler, Experimental 5 Mw Thermal Convection

Molten Salt Reactor, ORNL CF<58+6-66 (June 13,1958).

UNCLASSIFIED
ORNL-LR-DWG 31208

 

 

 

 

 

 

 

 

 

 

 

 

a i 2 ..
3 AIR REACTOR Lo
- LOCK CELL L

| MAINTENANCE -
AREA DO
MANIPULATOR -

 

 

 

 

 

SODIUM PUMP [

TURBINE-GENERATOR ROOM
{FOR FUTURE EXPANSION)

 

5-Mw RADIATOR

 

 

FUEL-TO-SODIUM &
HEAT EXCHANGER ]

 

 

 

 

 

 

F @ o o» i
FUEL EXPANSION TANK

 

 

 

   

 

 

—

 

 

 

~CUTTING PLANES FOR INSTALLING

e

 

 
 

 

 

 

   

 

 

et : e 50-Mw SODIUM SYSTEM
1 O a . a ’ TR T ™ 1. """
BLANKET-TO-SCDIUM L e S b
K EXCHAN’GER ‘- :1.:‘ e @l \ /é\\‘z‘\\\ i W 7
S S ";_somum DRAIN TANK

 

 

 

 

 

 

 

   

 

 

 

 

 

 

 

 

5 0 5 10 15 20
e — e n—]
FEET

Fig. 1.5.1. Elevation Drawing of 5«Mw Experimental Molten-Salt Reactor Plant,

49
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 31209

 

 

 
 
 
 
 
 
 

 

 

CONTROL
ROOM

 

- ~x

  

TURBINE GENERATOR ROOM
(FOR FUTURE EXPANSION]

 

 

FUEL-TO-SODIUM & -

 

 

HOT HEAT EXCHANGERE: -
.1 MAINTENANCE R
AREA

 

 

 

 

 

 

 

 

REACTOR

 

 

 

¥

 

 

 

! MAINTENANCE

    

BLANKET-TO-

 

 
  
     
  
 
 
   

SODIUM PUMP

 

 

   

 

5-Mw

o 84 f1,
"IRADIATOR ‘

 

 

AREA SODIUM HEAT
CH

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

T

  
 
  

BLOWER

 

 

 

5 0 5 10 15 20
e e

 

Fig. 1.5.2, Plan View of 5-Mw Experimental Moliten-Salt Reactor Plant,

as a one-region system, The 5Mw reactor en-
visioned could be constructed of components
already developed. [n order to provide for 50-Mw
operation, the fuel expansion tank was designed
so that a sump-type fuel pump could be installed
in it, and the sodium lines leading to the heat
exchanger were sized to handle the flow required
in the 50-Mw system,

The 5-Mw reactor would serve to demonstrate
the feasibility of continuous operation of a molten-
salt reactor, to provide in-pile corrosion data on
removable samples in hot and cold legs, to develop
and demonstrate remote maintenance procedures,
and, by replacing the air heat dump with a steam
heat dump, to demonstrate sodium-to-steam heat
transfer. When the experiment had been completed,
the system could be converted for 50-Mw operation
by installing a sump-type fuel pump in the fuel
expansion tank, a new sodium system consisting
of a 10,000-gpm pump and a sodium-to-steam heat
exchanger, and a turbine-generator system. The
fuel-to-sodium heat exchanger designed for 5-Mw

50

operation would be satisfactory for 50-Mw operation
because of the reduction in the fuel film resistance
that would occur as a result of going from laminar
flow at the lower power to turbulent flow at the
higher power.  Calculations indicate that the
bianket circuit could be designed so that thermal
convection would provide adequate circulation of
the salt at both power levels, but if necessary a
pump could be installed, Although such a plant
would not be identical with a large-scale power-
producing plant, there would be sufficient similarity,
especially with respect to control, corrosion, and
maintenance problems, so that successful operation
would lead directly to the design and construction
of a large power plant.

A rough estimate has indicated that the 5-Mw
plant would cost about $10,000,000, and that for
an additional $10,000,000 it could be converted to
the 50-Mw plant, The arbitrary power levels and
estimated costs would, of course, be subject to
change if optimization studies were carried out for
these systems.
PERIOD ENDING JUNE 30, 1958

Toble 1,5.1, Dimensions and Operating Conditions for Experimental
5~ and 50-Mw Molten<Salt Reactor Systems

 

 

Power output, Mw (thermal) 5.19 50

Reactor
Core diameter, ft 5 5
Blanket thickness, ft 0.5 0.5
Power density, w/em® 1.5 15
Diameter of riser and downcomer, in. 10 10
Height of fuel heat exchanger above 20 20

reactor centerline, ft

Fuel velocity in riser, ft/sec 0.64 6.17
Fuel head, ft 0.39 12.66
Fuel volume, #3 120 120
Fuel flow, gpm 158 1515
Sodium flow, gpm 578 9250

Heat exchanger
Tube inside digmeter, in. 0.6 0.6
Tube wall thickness, in. 0.050 0.050
Tube length, ft 20 20
Number of tubes 250 250
Shell outside diameter, in. 18 18
Shell wall thickness, in. 0.375 0.375
Fuel inlet temperature, °F 1210 1210
Fuel outlet temperature, °F 1010 i010
Sodium inlet temperature, °F 850 850
Sodium outlet temperature, °F 1100 1000

A 600-Mw THERMAL-CONVECTION REACTOR Heat exchanger pressure drop, Ib/§t2 27

J. Zasler

An analysis similar to that of Romie and Kinyon?
was used as a basis for a study of the possibility
of a thermal-convection reactor of approximately
600-Mw thermal output. A system with the follow-
ing dimensions would be required if a fuel-to-
sodium heat exchanger, with sodium in the tubes,
were used:

Core diameter, ft 8
Fuel temperature drop, °F 200
Height of heot exchanger above 25

reactor centerline, ft

Diameter of riser, ft 4

Heat exchanger (based on use of 4)

Number of tubes (in each exchanger) 6125

Outside diameter of tubes, in. 0.70

Tube length, ft 11.4

Fuel volumes, ft

In reactor and expansion tank 300

In riser and downcomer 800

In heat exchangers 675
Total 1775

 

2F, E. Romie and B. W. Kinyon, A Molten Salt

Natural Convection Reactor System, ORNL CF-58-2-44
(Feb. 5, 1958),

51
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

The 1775 ft3 value is to be compared with the
530 ft3 estimated for a reactor system in which the
fuel is circulated by a pump. The greater fuel
volume would be reflected in a higher cost for the
larger inventory of U233, the lower conversion
ratio, and the increased size of equipment, The
cost of power from such a system would probably
be approximately Y mill/kwhr higher than for the
system in which the fuel is circulated by a pump,

GAS-COOLED MOL TEN-SAL T HEAT EXCHANGER
R. E. MacPherson

A design study was completed of the heat ex-
change equipment that would be required for gas
cooling of a molten-salt reactor. For the study,
it was assumed that the reactor had an output of
574 Mw (thermal) that was divided among four
primary salt-to-gas heat exchangers. Helium,
steam, and hydrogen were considered as coolants,
and the effects of varying tube sizes, coolant
pressure level, and coolant inlet temperature (at a
fixed outlet temperature) were investigated,

The following dimensions and operating con-
ditions were assumed for a heat exchanger of the
type illustrated in Fig. 1,5.3 in the preparation of
Table 1.5.2, which gives operating cost data as o
function of the type of coolant:

Fins per inch 7.32

Salt pressure drop 40 psig
Enriched salt cost $1335/41°
Blower power cost 9 mills/kwhr

It may be seen that, for a given set of conditions,
hydrogen is the most effective coolant,
and steam are reasonably comparable, since the
larger-diameter heat exchanger required with steam
offsets the advantage of shorter length, There
incentives for choosing
steam over helium, since components for high-
temperature steam systems are well developed.
In addition, leakage of steam to the atmosphere
would be less of a problem inasmuch as the re-
placement cost would be negligible. However, the
of a steam-to-salt leak would be

Helium

are certain economic

consequences
quite serious.,

By lowering the coolant inlet temperature to
700°F, a heat exchanger with helium as coolant
appears to be more attractive dimensionally than
the heat exchanger with hydrogen at an inlet
temperature of 850°F. The hazards of hydrogen
usage would be avoided, but other problems would
be introduced by the gas being at a temperature
lower than the salt freezing temperature,

The results presented in Table 1.5.2 were based

 

 

 

Salt LiF-BeF ,-UF , (62-37-1 on a heat exchanger with circumferentially finned
mole %, fuel 130) tubing with the salt making four serpentine passes
Salt inlet temperature 1210°F across the helium stream. Inconel was used ex-
Salt outlet temperature 1075°F clusively as the material of fabrication for both
Coolant outlet temperature 1025°F tubes and fins to avoid any question of material
Coolant operating pressure 300 psig compatibility. Mechanical bondingof the fins to the
Heat load 143.5 Mw tubes was assumed., An optimization of the tubing
Tubing size 0.5 in. size on the basis of the assumed operating con-
Wall thickness 0.050 in. ditions indicated a tube diameter of 0.5 in. Larger
Fin dimensions tube diameters led to excessive tube lengths, and
Outside diameter 1.024 in. smaller diameters led to large numbers of tubes
Thickness 0.023 in: in the matrix.
Table 1.5.2. Gas+Cooled Molten-Salt Heat Exchanger Operating Costs
Based on Enriched Salt Cost plus Blower Power Cost
Coolant inlet Heat Exchanger Size Yearly Operating
Coolant
G Temperature Length Diameter Cost for One
° CF) (ft) (f+) Heat Exchanger

He lium 850 32 9.8 $216,000

Hydrogen 850 18 8.9 $198,000

Steam 850 26 11.2 $211,000

Helium 700 17 8.2 $191,000

 

52
PERIOD ENDING JUNE 30, 1958

           
           

 
 

  

= = =

by -, e

_||h \\\ |N|I1.|\lllll,./flfu1

=3 TR

3 = = =

= T W
- (e Rt
5 z 7 =\ J
= 3 N\\\ e
| - = = —
o Q n.\\w\\\ e H
© 8 e 3
= S e 3
1
<L
wn

=

 

              

) g AN = QIR 7 X,

\.,%
W \
i \
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s
w
7 7

  

%
7
\ Ittt

   

A
7,
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\.:\.\ \\ 4
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L [ 7
ST

   
    

   

          

o - SN\,
T - . o ST - - < =
s \\V\\\\\\\\\\ - = A N
L = - -
, =

   

FINNED TUBE MATRIX
COOLANT INLET

Fig. 1.5.3. Gas-Cooled Heat Exchanger.

33
Port 2
MATERIALS STUDIES
2.1. METALLURGY

W. D. Manly

D. A. Douglas

A. Taboada

Metallurgy Division

DYNAMIC CORROSION STUDIES
J. H. DeVan J. R. DiStefano

Corrosion experiments in which Inconel and
INOR-8 thermal-convection and forced-circulation
loops are being utilized were continued as part
of a three-phase corrosion program, previously
described.! No forced-circulation loop tests were
completed during the quarter, and therefore no
additional results of examinations are available.
The status of all forced-circulation loops now in
operation is presented in Chap. 1.2 of this report.
Tests of three Inconel and eight INOR-8 thermal-
convection loops were completed during the quarter,
however, and results of examinations of these
loops are presented here.

Inconel Thermal-Convection Loop Tests

The three Inconel thermal-convection loop tests
which were completed conclude the first phase of

UNCLASSIFIED
To14431 |

 

 

Fig. 2.1.1. HoteLeg Section of Inconel Thermal-
Convection Loop 1178 Which Circulated LiF-Ber-UF4
(62-37-1 Mole %) for 1000 hr at 1250° F. 250X.

 

corrosion testing of this alloy in contact with
fused fluoride mixtures. The purpose of this
initial testing phase was to obtain information
with which to evaluate the relative corrosive
characteristics of 12 fluoride mixtures in Inconel
thermal-convection loops operated for 1000 hr at a
maximum hot-leg temperature of 1250°F with a
temperature difference across the loop of 170°F.
The hot leg of Inconel loop 1178, which circulated
fuel mixture 130 (62 mole % LiF-37 mole %
BeF2-] mole % UF,), was found to have voids
to a depth of | mil, as shown in Fig. 2.1.1. The
second loop (1191), which circulated fuel mixture
131 (60 mole % LiF-36 mole % BeF,—4 mole %
UF,), was found to have hot-leg attack in the
form of intergranular voids to a depth of less than
1 mil, as shown in Fig. 2.1.2, The third loop

 

]J. H, DeVan, J. R, DiStefano, and R. S, Crouse,
MSR Quar. Prog. Rep. Oct. 31, 1957, ORNL.2431, p 23,

- UNCLASSIFIED
$ T.14698

 

 

 

Fig. 2.1.2. Hot-Leg Section of Inconel Thermal-
Convection Loop 1191 Which Circulated LiF-BeF,-UF,
(60-36-4 Mole %) for 1000 hr ot 1250°F, 250X.

57
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

{1192) was operated to check previous test results
obtained with fuel mixture 125 (53 mole % NaF-46
mole % BeF2—0.5 mole % UF4—0.5 mole % ThF4).
The previous loop (1174) that circulated this fuel
mixture at 1250°F was attacked to a depth of
2 mils. The check of the results was considered
to be necessary (1) because corrosion theory
indicates that UF , should effect a higher corrosion
rate than comparable additions of ThF, and
(2) because a loop (1163) which had circulated a
mixture which was the same, except that the
ThF, was replaced with UF ;, showed quite limited
corrosion to a depth of less than 1 mil. The

examination of the check loop (1192), however,
showed attack to a depth of 4 mils in the hot
section. No explanation for the increased attack is
readily apparent at this time. Chemical analyses
of the salts circulated in these loops are presented

in Table 2.1.1.
INOR-8 Thermal-Convection Loop Tests

The operating conditions and results of eight
INOR-8 thermal-convection loop tests are presented
in Table 2.1.2. These are the first results reported
for INOR-8 loops operated under MSR conditions.
As may be observed, the results are favorable;

Table 2.1.1. Chemical Analyses of Salts Circulated in Inconel Thermal«Convection Loops for 1000 hr
ot a Hot-Leg Temperature of 1250°F

 

Major Constituents Minor Constituents

 

 

Loop Salt Salt Composition (wt %) (
o ppm)
No, No. (mole %) Sample Taken
U Be Th Ni Cr Fe
1178 130 62 LiF =37 BeF2—1 UF, Before test 6.81 8.88 100 65 230
After test 6.76 8.50 80 160 240
1191 131 60 LiF-36 BeF2-4 UF, Before test 21.8 7.00 320 40 210
After test 19.9 6.77 145 75 175
1192 125 53 NaF —46 BeF2 -0.5 Before test 2.26 8.39 2.66 175 60 210
UF4 -0.5 ThF,
After test 1.60 8.00 2.56 45 160 180

 

Table 2,1,2, Operating Conditions and Results of INOR-8 Thermal-Convection Loop Tests

 

 

Loop Salt Salt Composition Temperature of Hot Section Operating Time Depth of Attack
No. No. {(mole %) (°F) {hr) (mils)

1162 123 53 NoF 46 BeF, 1 UF, 1250 6360 0.75

1164 124 58 NaF =35 BeF2—7 ThF4 1250 1000 0

1165 12 11.5 NaF 46,5 LiF -42 KF 1250 1340 0

1179 130 62 LiF-37 BeF, -1 UF, 1250 1000 0

1194 12 11.5 NaF -46.5 LiF-42 KF 1125 1000 0

1195 84 27 NaF =35 LiF-28 BeF, 1125 1000 0

1197 126 53 LiF—46 BeF,~1 UF, 1250 1000 0

1205 129 55.3 NaF -40.7 ZrF, -4 UF, 1250 1000 0

 

58
only loop 1162, which operated for 6300 hr, showed
measurable attack. The attack in loop 1162, as
shown in Fig. 2.1.3, was in the form of widely
scattered subsurface voids. The attack ranged
to a depth of 0.75 mils at the hottest point of the
loop. Chemical analyses of the salts circulated in
these INOR-8 loops are presented in Table 2.1.3.

GENERAL CORROSION STUDIES

E. E. Hoffman
W. H. Cook D. H. Jansen

Carburization of Inconel and INOR.8

The study, previously described,? of the problems
associated with the use of graphite in direct
contact with molten-salt fuel in o reactor was
continued. Since carburization occurs slowly, if
at all, in a molten-salt system, specimens for
studies of the effect of carburization on the
mechanical properties of Inconel and INOR-8 are
obtained by exposure of the specimens in a sodium-
graphite system. The apparatus used for preparing
the specimens is shown in Fig. 2.1.4. Tensile
specimens 0.040 in. thick and specimens for
metallographic and chemical onalyses that consist

of 0.040-in.-thick plates and 0.25-in.-thick bars

 

2E. E. Hoffman, W. H. Cook, and D. H. Jansen, MSR

PERIOD ENDING JUNE 30, 1958

are held at 1600°F for 40 hr both in the sodium-
graphite environment and in an argon environment.
The specimens heated in an argon environment
serve as control specimens for determining the
effect of the heat treatment.

 

 

UNCLASSIFIED | [ »
T.14646 ¥

o

z

 

 

 

010

 

 

 

# . - #
#2 T F e FS L R I
P = -t & Sy 011

 

 

 

 

 

 

 

 

-
a1 5 .
SR gl e TR TN
Ay v\ oo ee : o2
A e kS g P
oA Yt
* A * 3 .013
2 AP 2
Wooatt 20 T N e, oo '
Wew o h e Lo Lol e s
* - £ .- .‘?V ’ -"_ ¥ o E ‘ . ‘ .*
Ty \M T > e Ll 0I5
. . v A ’ & - fi L M ‘I,
S %A . & ok cedvd | x
T - x k. = » < e Fre %g “@.' y
v o - L t : " Ty, e O
= " éi ¢ e-_ o e g0 P ? {’” -y i W0
; + .?( . F‘w - v..-. h“ : " .% o
« . S - S o — oy . ‘-
:“: P A w R -‘. " :;. a“?'...’ < s c*a
Fig. 2.1.3. Hot<Leg Section of INOR-8 Thermal-

Convection Loop 1162 Which Circulated NaF-BeF,-UF,

 

 

 

 

Quar. Prog. Rep., Jan. 31, 1958, ORNL-2474, p 54. (53-46-1 Mole %) for 6360 hr at 1250°F. 250X.
Table 2,1.3. Chemical Analyses of the Salts Circulated in INOR-8 Thermal-Convection Loops
Major Constituents Minor Constituents
Loop Neo. Sample Taken (wt %) (ppm)

Be U Th Cr Ni Fe
1162 Before test 8.87 3.25 445 75 110
After test 8.50 2.95 250 72 580
1164 Before test 4.46 25.5 200 85 420
After test 5.15 20 300 20 170
1179 Before test 7.03 8.97 120 175 200
After test 650 8.64 10 215 70
1195 Before test 9.33 30 1730 85
After test 8.72 30 30 250
1197 Before test 10.2 6:62 15 65 30
After test 9.75 6.30 150 55 265

 

59
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 27988R

—-—— HASTELLOY B
CONTAINER

 

 

 

 

 

 

 

 

 

 

SODIUM BATH e

 

 

INCONEL CONTROL SPECIMENS
(IN ARGON ATMOSPHERE)

INOR-8 SPECIMENS ——

 

 

 

 

INCONEL SPECIMENS

 

 

T

TR

 

 

INOR-8 CONTROL SPECIMENS
(IN ARGON ATMOSPHERE)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

AT

 

GRAPHITE BLOCK

Fig. 2.1.4. Apparatus for Carburization of Inconel and INOR-8 Specimens.

60
Metallographic exeminations of the 0.040-in.-thick
plates exposed to the sodium-graphite system
showed Inconel and INOR-8 to be carburized to
approximately the same depth. Heavy carburization
was found to a depth of about 0.003 in. and some
evidence of carburization could be detected to a
depth of 0.012 to 0.014 in., as indicated in
Fig. 2.1.5. Chemical analyses, presented in

PERIOD ENDING JUNE 30, 1958

Table 2.1.4, indicated that the INOR-8 was more
heavily carburized than the Inconel; that is, a
greater carbon concentration was found in millings
taken from the cross section and surface of the
INOR-8 alloy than in those from the Inconel.
Tensile tests were run in room-temperature air
on three specimens from each of the control and
carburized units. Since the results were in good

UNCLASSIFIED |

 

| UNcLASsIFIED |

 
  

 
 

 

 

 

 

 

 

 

 

 

 

 

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Fign 2.].5-

Specimens 0,040 in, Thick of (a) Inconel and (b) INOR-8 (Heat SP-16) After Exposure in a Sodiums-

Graphite System for 40 hr at 1600°F, Etchant: copper regia, 500X.

Table 2.1.4. Carbon Analyses of Millings Taken from 0.25-in.=Thick Bars of Inconel and INOR-8
Carburized in a Sodium-Graphite System for 40 hr at 1600° F

 

Sample

Carbon Found (wt %)

 

INOR-8 Specimen

Inconel Specimen

 

Millings from cross section
First 0,010 in. from surface {(0-0.010 in.)
Second 0.010 in. from surface (0.010-0.02Q in.)

0.04 0.06
0.25 0.33
0.09 0.08

 

61
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

agreement, values are presented in

Table 2.1.5.

average

As may be seen in Table 2.1.5, carburization of
the Inconel increased its tensile strength and
yield strength and reduced its ductility. The
same carburizing treatment lowered the tensile
strength, slightly increased the yield strength,
and greatly reduced the ductility of INOR-8 (heat
SP-16).

Apparatys is being prepared for studies of
INOR-8 and Inconel in contact with a graphite
and molten salt system.

Brazing Alloys in Contact with Molten Salts
E. E. Hoffman D. H. Jansen

Two precious-metal-base brazing alloys showed
promising corrosion resistance to LiF-Ber-UF4
(62-37-1 mole %, fuel 130) in preliminary tests,3
and therefore other commercially available brazing
alloys having high gold or silver contents are
currently being tested at 1300°F. The compositions
of the alloys being tested and their approximate
brazing temperatures are listed in Table 2.1.6.
The solidus points of all these alloys are well
above 1250°F, the highest temperature expected in
a molten~salt power reactor.

In order to obtain long-term dynamic corrosion
data on brazing alloys in fuel 130, a series of
tfive alloys will be tested in duplicate by inserting
brazed lap in the hot legs of thermal-

The designations of the alloys

joints

convection loops.

 

3D. H. Jansen, MSR Quar. Prog. Rep. Jan, 31, 1958,
ORNL-2474, p 59.

that will be tested and their compositions are
given below:

Composition

Coast Metals alloy Nos» 52  89% Ni—5% Si—-4% B-2% Fe

Coast Metals alloy Nos 53  81% Ni—=8% Cr—4% Si—-4%

B-3% Fe

General Electric alloy 70% Ni=20% Cr-10% Si

NO. 8]

Au=-Ni 82% Au—18% Ni

Copper 100% Cu

The configuration that will be used for their

insertion in the hot leg of a thermal-convection
loop is shown in Fig. 2.1.6. The loops are ready
for filling and should be in operation in the near
future. A series of three identical loops is to be

Table 2,1.6. Precious-Metal-Base Alloys
Being Tested in Static Fuel 130

Test temperature: 1300°F
Test duration: 500 hr

 

Brazing Temperature

 

Material Range

(°F)
Pure silver 1760-1900
90% Ag—10% Cu (coin silver) 16001850
72% Ag-28% Cu 1435-1650
50% Ag—33.3% Au-16.7% Cu 15251600
75% Au~20% Cu-5% Ag 1650~1850
42% Au—40% Ag-18% Cu=0.6% Zr 15081600

 

Table 2,1.5. Results of Tensile Tests of Carburized and Control Specimens of Inconel and INOR-8

 

 

Tensile Strength Yield Strength at 0,2% Offset Elongation

Material (psi) {psi) {% in 2-in, gage)
Inconel

Control 75,878 20,906 47.25

Carburized 91,532 28,577 25,67
INOR-8

Control 116,338 50,542 49,50

Carburized 98,535 52,828 7.75

 

62
PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED
ORNL-LR-DWG 29965

 

 

 

  
  
 

   

SPECIMEN HANGER

COLD LEG—mm

 

 

SPECIMEN
(TEN)

 

 

 

 

 

Fig. 2.1.6. Configuration for Tests of Brazing Alloy Specimens in the Hot Leg of a Thermal-Convection Loop,

63
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

operated for 1000-hr, 5000-hr, and 10,000-hr periods

before the specimens are removed for examination.

MECHANICAL PROPERTIES OF
D. A. Douglas

The tensile properties of INOR-8 at various
temperatures were measured at this laboratory and
at Haynes Stellite Company on material from
similar heats that contained 0.02% carbon and at
Battelle Memorial Institute (BMI) on material from
a heat that contained 0.06% carbon. The ORNL
data are presented in Fig. 2.1.7 and are compared
in Figs. 2.1.8 and 2.1.9 with the Haynes Stellite
and BMI data. As may be seen from Fig. 2.1.8,
there is good correlation of the data. The BMI
results were expected to be slightly higher than
the others as a result of the higher carbon content.
The yield strength shows the expected trend, but
the tensile strength at 1250°F appears to be low.
The elongations measured at rupture in the same
experiments are shown in Fig. 2.1.9. The
relatively brittle failure which occurred in the

INOR-8

UNCLASSIFIED

BMI test at 1250°F explains the low tensile
strength reported in Fig. 2.1.8. The temperature
dependence of Young's modulus for INOR-8 is
shown in Fig. 2.1.10,

Relaxation curves obtained from data taken at
1200, 1300, 1400, 1500, and 1600°F are presented

in Fig. 2.1.11. Considerable relaxation occurs at

1400°F and above within the first 6 min of the

 

 

 

 

 

 

 

 

 

 

 

 

 

 

   
 

 

 

 

  
  

 

 

 

 

 

 

 

 

 

 

 

  
   
   

 

 

 

 

 

 

 

 

 

 

(x 10%) ORNL—LR-DWG 28709R
110 10
l 17 [ 1
TENSILE STRENGTH | E } ‘
100 - | | L 100
| i ‘ ‘ ‘ :
| \ | . .
90 b—— + : } 90
; ; |
80 —————— - ,,J_‘ - N —1 80
‘ i
- 70 : Lo -+ | - 70
a2 . ELONGATION (%) ! .
0
- T ’ 32
2 N\ =
—- i -— 60 2
w 60 f ‘ I
- ) o
” %
50
40
30
{20
| i i
S | | )
0 400

800 1200
TEMPERATURE (°F)

1600

Fig. 2.1.7. Tensile Properties of INOR-8 as a Function

of Temperature.

64

test., At 1300°F and below, incubation periods of
(107} UNCLASSIFIED
130 ORNL-LR-DWG 31214
! T T ] T ] T T T
120 [-B~k—-+ ULTIMATE TENSILE STRENGTH—————r—F rfi
. . ; ;
\"“‘*\! 1 ! | ‘ | : |
"o -"'"-a__‘ : ‘[ | i |
—_— ~ ST
S |
100 [— ORNL ; =] . -
50 ] ! >~ g HAYNES STELLIT
= o~ ' .
2 : L~
5 80 - ; f f
w ,
€ 70 : Lo
L | ol
? 60 ‘ S
- SN | '
50 +——+——0.2% OFFSET YIELD STRENGTH ——
| SSSSes=mmgo e[ B
e —— e
30 | HAYNES .~ A RS e
20 i L | 4
o 200 400 600 800 1000 1200 1400
TEMPERATURE (°F)
Fig. 2.1.8, Comparison of Tensile Properties of

INOR-8 Obtained at ORNL, Haynes Stellite Company,
and Battelle Memorial Institute (BMI).

UNCLASSIFIED
ORNL-LR-DWG 31212

68

'*jfi BMI
W e o o o e -

HAYNES STELLIT

ELONGATION AT RUPTURE (%}

 

Q 200 400 600 800 1000 {200 1400
TEMPERATURE (°F)
Fig. 2.1,9. Comparisons of Tensile Elongations of

INOR-8 Specimens Tested at ORNL, Haynes Stellite

Company, and Battelle Memorial Institute,
1 hr and longer occur before creep commences. It
appears that, even at 1200°F, INOR-8 will
eventually relax to a stress of 2000 to 3000 psi.

Creep tests of INOR-8 (heat SP-16) are under way
at temperatures ranging from 1100 to 1500°F with
fuel 107 (NoF-KF-LiF-UF,, 11.2-41-45.3-2.5
mole %). The data obtained thus far are plotted in
Fig. 2.1.12 os creep rate vs stress for temperatures

UNCLASSIFIED
ORNL-LR-DWG 29238

— 1 1T 17 1T 7T 7T 1

 

26

YOUNG'S MODULUS
n
5

 

na
N

 

 

L1 | 1
0 200 400 €600 800 1000 1200 1400
TEMPERATURE (°F)

A 1L

 

™
o

PERIOD ENDING JUNE 30, 1958

of 1100, 1200, and 1300°F. Stress-rupture and
creep results obtained in tests in air at Haynes
Stellite Company are presented in Fig. 2.1.13, and
data obtained at ORNL in molten salts under
similar test conditions are presented in Fig.2.1.14.
The correlation of the data is excellent.

UNCLASSIFIED
ORNL-LR-DWG 31213

(X 103)
26 —
o4 1200°F
25 [ 4300°F!
L = ‘ B

20 1£}OO°F. f [T T :

18 i ‘ el
=16 ‘ , ‘ ey
@ 1500°F | T~ : : |
;;‘4 ! NG J”—rl -
o1 j B : ‘ —
T40
o

8

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6
4
2
0

o1 1o 20 50 iC0

 

 

 

 

 

 

 

 

 

 

 

 

 
   

 

 

 

 

 

 

 

 

 

   

 

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig, 2.1,10. Young’'s Modulus for INOR-8 as a Fig. 2.1.11. Relaxation Curves for INOR-8 Stressed
Function of Temperature, to 0,1% Strain at Temperatures from 1200 to 1600°F.
UNCLASSIFIED
50,000 OTNL—LRI—DWG 312‘;4
' T |
Loy | | L
T o T
40,000 —[ l ’ ll ( ’ } J— { I ‘ |
| l /
35,000’-——fl {H / ’ { | | J
30,000 %—— | 1 i ( | { {
W 1100°F |
{ 1200°F ‘
= 25,000 b _J’_
2
. /
& 20,000 /
| |
/|
v
V] A
/|
15,000 [T ’7 o
L L

 

 

108 2 5 1074 2 5

 

 

 

 

 

 

1073 2 5 1078 2 5 10~

MINIMUM CREEP RATE (% /hr}

Fig. 2.1.12, Creep Rates vs Stress for INOR-8 Tested at Various Temperatures in a Molten-Salt Environment,

65
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

(X 10%) UNCLASSIFIED
100 ORNL—LR—DWG 34215

50

20

STRESS(psi)

 

1 2 5 10 20 50 100 200 500 1000
TIME (hr)

Fig. 2.1.13. INOR-8 Creep and Stress-Rupture Data
Obtained in Air by Haynes Stellite Company.

UNCLASSIFIED

 

   

STRESS (psi)

10 yr |

o2 L LT VL L

|
1yr§
|

 

 

 

 

1 10 100 1000 10,000
TIME (hr) TO 1% STRAIN

100,000

Fig. 2.1.14,
Tested at

Time to 1% Strain vs Stress for INOR-8
Yatious Temperatures in a Molten-Salt

Environment,

FABRICATION STUDIES

J. H. Coobs
H. Inouye T. K. Roche
Influence of Composition on Properties of INOR-8
T. K. Roche

An investigation was completed of the influence
of composition variation on the creep-rupture
strength at 1500°F and the microstructure of alloys
containing 10-20% Mo, 5~10% Cr, 4-10% Fe,
0.5% Al, 0.5% Mn, 0.06% C, and the balance nickel.
A report has been prepared that presents the
details of this investigation,4 and the results are
summarized below.

The compositions of the alloys studied were
varied systematically in an effort to determine the
effect of each element on the alloy strength.
Creep-rupture tests at a stress of 10,000 psi were
run on all the alloys in the annealed condition.
The criteria used to evaluate the strengths of the

66

alloys were the times required to produce strains
between 1 and 10%.

The results could not be explained simply in
terms of composition variation, since the principal
factors affecting the strength of the alloys were
solid-solution elements, carbide and noncarbide
aging reactions, the presence of M6C-fype carbides
in the microstructures, and grain size. From the
standpoint of creep-rupture strength, it was possible
to conveniently group the alloys according to the
three concentrations of molybdenum studied: 10,
15, and 20%. It could be concluded from the
chemical and microstructural analyses that the
relative contribution of each of the factors to the
strength of the alloy varied among the groups.

The combined effects of solid-solution strength-
ening by molybdenum and the increase in quantity
of dispersed M C-type carbides which this element
promoted in the annealed materials were the
predominant factors which increased the strength
of the alloys grouped by molybdenum content.
The only exception noted was the 20% Mo-7%
Cr-10% Fe alloy, which precipitated a noncarbide
phase as a consequence of crossing a new phase

boundary. This phase contributed noticeably to
creep-rupture strength in the later stages of
testing.

The contributions of chromium and iron to the
strengths of the alloys within the individual groups
could not be established with certainty because of
simultaneous variations in other factors affecting
creep-rupture behavior. In order to obtain a better
indication of the strengthening influence of
chromium and iron, creep-rupture studies were
conducted on low-carbon ‘‘high-purity’’ alloys.
Although the analysis of the data was complicated
by the presence of a limited amount of carbide
precipitation and by grain-size variations, the
influence of chromium was found to be significant
when 5 to 10% was added to a 15% Mo-bal Ni
base, with the presence of 10% chromium in the
base composition resulting in the most pronounced
strengthening influence. The strengthening effect
of iron was interpreted as being insignificant when
amounts up to 10% were added to a 15% Mo-7%
Cr—bal Ni base. A general consideration of all
the data obtained from this investigation favorably

 

41, K. Roche, The Influence of Composition Upon
the 1500°F Creep-Rupture Strength and Microstructure
of Molybdenum-Chromium-Iron-Nickel Base Alloys,
ORNL-2524 (June 24, 1958).
supported the composition specification selected

for the alloy INOR-8.

High-Temperature Stability of INOR-8
H. lnouye

Embrittiement studies of INOR-8 in the temper-
ature range of 1000-1400°F were completed for
aging times up to 2000 hr. Data from tensile
tests of such specimens indicate that aging
produces no significant changes in the properties
of the alloy. Two heats containing different carbon

PERIOD ENDING JUNE 30, 1958

contents were found, however, to have significantly
different properties. The lower carbon content
alloy (heat SP-19, 0.06% C) behaved in a more
ductile manner than the higher carbon content
alloy (heat 8M-1, 0.14% C).

The data obtained in the tests are compared in
Figs. 2.1.15 and 2.1.16 with data for Hastelloy B,
whose tendency to embrittle was previously
discussed.’

 

SR. E. Clausing, P. Patriarca, and W. D. Manly, Aging
Characteristics of Hastelloy B, ORNL-2314 (July 30,
1957).

UNCLASSIFIED
ORNL—-LR—DWG 29617

 

160

 

 

 

140

 

 

 

 

120

 

 

 

 

 

 

 

100

o AGED 500 hr
m AGED 2000 hr

 

TENSILE STRENGTH (1000 psi)

 

 

   

A AGED 1000 hr
0o AGED 2000 hr

———HAST B —

v AGED 500 hr
A AGED 1000 hr

 

 

 

 

 

 

 

 

 

 

 

ELONGATION (%)

 

 

 

 

 

 

 

 

 

1000

1100

 

1200

1300 1500

AGING TEMPERATURE (°F)

Fig. 2.1.15. Room-Temperature Tensile Properties of INOR-8 and Hastelloy B After Aging in the Temperature

Range 1000-1400°F for Times Up to 2000 hr.

67
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Status of Production of INOR-8
(Westinghouse Subcontract 1067)
H. Inouye
Five air-melted heats of INOR-8 consisting of
about 5000 |b each were prepared at the Blairsville
Metal Plant of the Westinghouse Electric Corp.
About 20,000 |b of finished products are to be

furnished from these heats,

Status of Production of Seamless Tubing
(Superior Tube Company Subcontract 1112)

H. [nouye

The first shipment of seamless INOR-8 tubing
(0.500 in. OD, 0.045 in. wall) was received during
the quarter. It consisted of approximately 150 f1
of tubing which was processed by the Superior
Tube Company's production department without

UNCLASSIFIED
ORNL—LR—DWG 29616

 

 

 

 

 

 

 

 

 

 

 
 

 

  

 

   

  

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

  

 

 

 

 

 

140 | | | |
S HEAT 8M4// ———HAST B |
. v ANNEALED x ANNEALED 0 ANNEALED
'3420 -0 AGED 500 hr ® AGED 500 hr vV AGED 500 hr
o ¢ AGED 1000 hr A AGED 1000 hr A AGED 1000 hr
S m AGED 2000 hr 0 AGED 2000 hr ——
£ 100 ’i%%?‘:‘« ‘
\ v mme o GED 5 sua
: b e i~
=z SISO RS - N SO N
lfig [ \\ -
— 80
w)
wl
-
2 N\
= 60 =~
40
3
= SRR /57
=z K
S L~ 7
o R e
S R R AT S 7
5 v vy o
-
. QTR =
—— S —— o ’-—"T
———— ——::: .—--—-—‘
1000 1100 1200 1300 1400 1500

TEST TEMPERATURE (°F)

Fig. 2.1.16. Elevated-Temperature Tensile Properties
ature Range 1000~1400°F for Times Up to 2000 hr,

68

of INOR-8 and Hastelloy B After Aging in the Temper-
difficulty. This was the first production run of
seamless tubing by a commercial vendor. The
production involved an extrusion from a 9-in.-dia
drilled billet to a 4-in.-OD, ]/z-in.-wall tube shell
by the International Nickel Company (on a purchase
order), reduction by Babcock & Wilcox to a 2-in.-0OD,
0.187-in.-wall tube, and tube reduction and re-
drawing by the Superior Tube Company to the final
size. Despite the circuitous route involving four
manufacturers, the tubing appears to be of excellent
quality. (The inspection of this material is
described in a subsequent section of this chapter.)

Bearing Materials

J. H. Coobs

INOR«8 Journals, — Two 3-in.-dia journals of
INOR-8 were flame-sprayed with INOR-8 for use
in bearing tests. One journal was built up with
5 mils of sprayed coating and the second with
20 mils. The journals were first sprayed with a
light layer (~2 mils) of molybdenum to provide a
rough bonding surface, and after being sprayed
with INOR-8 were heat-treated at 2250°F to bond
the coatings.

During the heat treatment, the 5-mil coating on
the first journal separated at one end. This journal
was recoated after being rough-machined to provide
a better bonding surface. The final coating was
more than 25 mils thick and was successfully
applied and finished. These results indicate that
the thickness of the sprayed coating for a 3-in.-dia
journal should be a minimum of 20 mils.

An experiment was also run in which an INOR-8
journal was flame-sprayed with molybdenum. A
sample of ]'/2-in.-dia INOR-8 pipe was threaded to
provide a good bonding surface, spray-coated with
about 20 mils of molybdenum, and subjected to
thermal cycling. The great difference between the
thermal expansion of INOR-8 and that of molybdenum
resulted in severe cracking of the coating during
the first thermal cycle. After 23 additional thermal
cycles from room temperature to 1350°F, the
molybdenum layer was partially separated from the
INOR-8 base. The large cracks that developed in
the coating during the first thermal cycle and the
area where the coating broke away from the

INOR-8 pipe may be seen in Fig. 2.1.17.

Molybdenum Journals. — Several l-in.-dia speci-
mens of molybdenum rod were sprayed with
molybdenum in an effort to develop a suitable
procedure for spray-coating molybdenum journals.

PERIOD ENDING JUNE 30, 1958

Preliminary tests indicated that a surface which
had been grit-blasted and etched was most suitable
for bonding, and that a high-temperature heat
treatment was needed for producing reliable
coatings. Evaluation tests are now under way on
two specimens of 21/4-in.-dia molybdenum rod.
These rods were rough-threaded to provide a good
bonding surface. One sample was then cleaned by
bright-annealing at 1800°F, and the other was
etched with chromic-sulfuric pickling solution.
Both specimens were sprayed with about 20 mils of
molybdenum and then heat-treated at 2250°F for
2 hr in order to bond the coatings.

WELDING AND BRAZING STUDIES

R. L. Heestand
G. M. Slaughter

Weldability Evaluations

P. Patriarca
E. A. Franco-Ferreira

Results of tests of specimens of Haynes heats
SP-16 and -19 that were machined from V-in. plate
and were sent to Rensselaer Polytechnic Institute
for hot ductility testing are presented in Table 2.1.7.
Since weldability tests of these two alloys had
indicated severe weld-metal and base-metal
cracking of the SP-16 material and good weldability
of the SP-19 material, the results of these tests
were expected to be of value in determining
whether welding techniques and materials or the
base material composition was the cause of the
difference in weldability. = As may be seen, the
ductility of the SP-19 alloy upon heating to 2200°F
was significantly higher than that of the SP-16
alloy. Further, the ductility of the SP-19 alloy
was not appreciably impaired by prior heating to
2300°F, but it was impaired somewhat by prior
heating to 2350°F and above.

Additional weld test plates were made for
mechanical property studies of welded joints and
for radiographic, metallographic, and hardness
examinations, as well as to obtain general
information pertaining to the welding characteristics
of the materials under conditions of high restraint.
In general, the weldability of the INOR-8 alloys
appears to be good. No weld cracking difficulties
have been experienced with any of various heats,
except Haynes heat SP-16. The welds have been
found to be sound, in general, and the welding
deposition characteristics of the material are
comparable with those of Inconel or the stainless

69
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
Y-26035

 

Fig. 2.1.17.
Temperature to 1350°F.

steels. A new International Nickel Company filler
metal, designated INCO Weld A, which was designed
for the joining of dissimilar materials, is being
tested for applicability to the joining of Inconel to
INOR-8. This new filler metal is reported to be
subject to aging at high temperatures {contains
3% titanium), and therefore its mechanical proper-
ties both in the as-welded and aged conditions

should be studied.

All-weld-metal reduced-section tensile bars
(0.252-in. dia) were prepared from the test plates
of Westinghouse heat M-5, Haynes heat SP-19, and
Inconel, and tensile tests were conducted at room
temperature, 1200, 1300, and 1500°F. The results
are presented in Table 2.1.8. The Inconel data
were obtained in order to provide a comparison
with the data for INOR-8. Comparisons of wrought
sheet and weld metal of both Inconel and material
from heat SP-19 are shown in Fig. 2.1.18.

Although moderate ductilities were obtained at
the lower temperatures for both heats of INOR-.8

70

INOR-8 Pipe Spray-Coated with Molybdenum and Heated Through 24 Thermal Cycles from Room

weld metal, the ductilities of both heats were
10% or lower at 1500°F. Although the ultimate
tensile strengths are approximately the same, the
yield strengths of the two heats of weld metal
were much greater than those of the wrought sheet.
For example, the yield strength of heat SP-19
wrought sheet at room temperature was approxi-
mately 44,000 psi, while it was approximately
30,000 psi at 1200°F. The ductilities of INOR-8
in both forms are comparable. An examination of
the data also reveals that, in general, INOR-8
possesses a slightly higher ultimate tensile
strength than that of Inconel, a significantly higher
yield strength, and a markedly lower high-temper-
ature ductility.

Remote Welding

E. A. Franco-Ferreira

A test specimen was supplied to Welding
Processes, Inc., Wilmington, Delaware, for semi-
remote welding by a machine being developed for
Tflble 2.].7-

PERIOD ENDING JUNE 30, 1958

Results of High-Temperature Ductility Tests on Hoynes SP-16 and SP-19 Materials

 

Reduction in Area

Ultimate Tensile Strength

 

 

 

Thermal Cycle (%) (psi)

SP-16 SP-19 SP-16 SP19
1800°F (on heating) 40.8 63.4 53,600 56,400
2000°F (on heating) 30.8 73.8 41,300 42,800
2200°F (on heating) 1.8 72.8 3,400 30,200
2300°F (on heating) 1.1 6.0 20,000
2400°F (on heating) 0.5 0.0 1,300 1,800
2400°F cooled to 2300°F 046 0.0 600 5,600
2400°F cooled to 2200°F 0.1 1,700
2400°F cooled to 2000°F 1.8 14,000
2400°F cooled to 1800°F 4,5 17,300
2350°F cooled to 1800°F 33.2 55,200
2350°F cooled to 2000°F 30.8 44,400
2350°F cooled to 2200°F 0.3 24,900
2300°F cooled to 1800°F 5%.1 59,000
2300°F cooled to 2000°F 68.4 45,900
2300°F cooled to 2200°F 63.8 31,800

 

the Westinghouse PAR Project. The specimen was
a 10-in.-dia Inconel pipe with a 5/B-in. wall. The
weld was made with the pipe axis vertical; nitrogen
was used as the backup gas; ond the arc was
shielded with helium. A continuous arc time of
5 hr was required for the 59 passes of the weld.

Yiews of the face and root of the sample weld
are shown in Figs. 2.1.19 and 2.1.20. The large
amount of root push-through was apparently the
result of internal mismatch in the pipe samples
rather than to an inherent defect in the welding
procedure, Radiographic inspection at ORNL
showed the weld to be completely sound, with no
evidence of porosity. A cross section of the weld
may be seen in Fig. 2.1.21. The structure is that
of a normal Inconel weldment.

Joint Development
G. M. Slaughter

property tests were made on cast
silver in order to provide information pertinent to
the design of a flanged joint with a cast metal seal

Mechanical

(see Chap. 1.2, this report). The 0.252-in.-dia
tensile bars prepared from cast silver were tested
at room temperature, 1200°F, and 1400°F. The
test results indicated that the room-temperature
mechanical properties were about the same as
those given in published data, that is, 20,000-psi
tensile strength, 8,000-psi yield strength, and 50%
elongation in 1 in. At 1200°F, the tensile strength
dropped to 2400 psi and the yield strength to
1900 psi, with a corresponding elongation of 6 to
10%. At 1400°F, the tensile strength was 1500 psi,
the yield strength was 1000 psi, and the elongation
was 10%.

Half-sections of two flanged joints with cast-
metal seals that were tested under simulated
operating conditions, as described in Chap. 1.2,
are shown in Figs. 2.1.22 and 2.1.23. The sedl
material used for the joint shown in Fig. 2.1.22
was silver, whereas the seal material used for the
joint shown in Fig. 2.1.23 was a silver-copper
alloy. The examination indicated that moderate
oxidation of the components had occurred during

71
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

opening and closing of the flange and that the
oxidation had impeded wetting of the base metal.
The joint with the silver-copper alloy seal appeared
to be less subject to nonwetting than the joint
containing the pure silver seal, probably because
of the lower temperatures required to remelt the
alloy during opening. A typical interface between
the silver-copper alloy and the Inconel base metal

is shown in Fig. 2.1.24.

Table 2.1.8, Comparative Tensile Properties
of All-Weld-Meta! Specimens of INOR-8 and Inconel

 

 

 

(As Welded)
INOR-8
Temperature Westinghouse Haynes Heat Inconel
Heat M-5 SP-19
Ultimate Tensile Strength (psi)
Room 116,615 116,725 95,295
121,260 115,095 93,975
1200° F 72,350 74,965 67,570
75,575 65,930 68,945
1300°F 62,855 59,075 58,740
67,500 64,080 59,015
1500°F 54,595 53,780 37,480
57,040 52,175 36,505
Yield Strength at 0,2% Offset {psi)
Room 75,385 78,425 56,790
80,840 76,390 55,980
1200°F 55,580 54,595 37,790
57,240 50,525 44,100
1300°F 52,900 38,610
50,300 37,390
1500°F 50,520 44,815 34,630
51,130 44,930 34,285
Elongation (% in l-in. gage)

Room 41 37 411

52 38 45

1200°F 18 19 49

18 17 41

1300°F 12 16 38

12 40

1500°F 4.5 9.5 77

5.5 10 65

 

72

Component Fabrication
G. M. Slaughter

The Griscom-Russell Company, Massillon, Ohio,
is developing an internal tube welding procedure
that would be applicable to the attachment of tubes
to tube sheets in heat exchangers if back-brazing
were impractical, as it would be for thick tube
sheets or exceptionally large heat exchangers.
In order to make tube-to-tube-sheet joints with the
procedure being developed, the tube sheet is
drilled with holes of the same dimension as the
inside diameter of the tubing, and bosses are
machined or otherwise formed on the underside of
the tube sheet to permit a butt welding operation.
Inert-gas-shielded tungsten-arc welding is then
performed internally with a specially constructed
rotating-electrode mechanism designed to provide a
full penetration weld. A test joint is shown before
and after welding in Fig. 2.1.25.

The joints made by this GriscomRussell
procedure have included 1]/2-in.-OD, 0.072-in.-wall
to ]/2~in.-0D, 0.070-in.-wall tubing, but most of the
experience has been with the larger size tubes.
The primary production item has been a heat
exchanger made for the Knolls Atomic Power
lLaboratory of 2.25% Cr—1% Mo steel, which con-
tained 76 welds. The tubes were 11/2-in. OD and
15/] -in. OD, and both sizes of tubing had 0.072-in.
wafis. A power supply which consists of a
200-amp Vickers Controlarc rectifier with a program
timer was used for this welding. The weld program
high current, rapid current
decay to the normal welding current, and slow
decay at the weld termination. The tungsten
electrode was seated in a modified copper collet
assembly, and a stainless steel gas-focusing
assembly was used to direct the shielding gas onto
the weld zone.

Test welds made with 1-in.-0D, 0.070-in.-wall,
type 316 stainless steel tubes were obtained for
examination at ORNL. Incomplete fusion of the
locating lips of the headers was found in some
areas of some of the tubes, but this could probably
be remedied by careful determination of the optimum
welding variables.  Moderate oxidation of the
outer surfaces of the welds was found, but this
condition was probably the result of inadequate
purging by the makeshift welding fixture being
used. Nondestructive and metallographic ex-
aminations of these welds have not been completed,
and it is not yet known whether they contain

included an initial

e R B e
PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED

ORNL—LR—-DWG 31246

 

 

 

—B— [NOR-8 WELD METAL (ULTIMATE TENSILE STRENGTH)
~=B== [NOR-8 WELD METAL (YIELD STRENGTH)

—(— [NCONE{. WELD METAL {ULTIMATE TENSILE STRENGTH)
=== [NCONEL WELD METAL (YIELD STRENGTH)

e @e== |[NOR-8 SHEET (ULTIMATE TENSILE STRENGTH)
~=@== INOR-8 SHEET (YIELD STRENGTH)

—0— INCONEL SHEET (ULTIMATE TENSILE STRENGTH)} FROM
=-0-=- INCONEL SHEET (YiELD STRENGTH) INT. DUCKEL

 

(X 103)
160
120
R -
.—‘—
.’% -
e
C 80
i
e
L
o
}._
o)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

40 |
O
80 I
—ge— NOCR-8 WELD METAL )
—pe  (NCONEL WELD METAL
—@= [NOR—8 SHEET
———O—— INCONEL SHEET
;6 o L
= ———
g 40 | \ . /
1
<J o - Y;’
O
=
O
J \
L .\
| »
5._;
B
—
0
ROOM 1400 1200 1300 1400 1500
TEMPERATURE TEMPERATURE (°F)

Fig. 2.1.18, Comparative Average Mechanical Properties of INOR-8 Heat 5P-19 and Inconel Weld Metal and

Wrought Sheet.

73
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED © UNCLASSIFIED

Y-26262 Y-26261

  

Fig. 2.1.19. Face of an Experimental Weld Made by a

Fig. 2.1.20, Root of the Experimental Weld Shown in
Semiremote Welding Process Being Developed by

Fig. 2.1.19,

Welding Processes, Inc.

UNCLASSIFIED
Y~25972

 

Fig. 2.1.21. Microstructure of the Experimental Weld Shown in Figs. 2,1.19 and 2,1,20, Etchant: electrolytic
oxalic acid, 10%. 2X.

74

   
PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED
Y-26103

  

UNCLASSIFIED
Y-26029

   

0

Fig., 2.1.22. Half-Section of a Flanged Joint with a

Cast Silver Seal. Figs 2.1.23. Half-Section of a Flanged Joint with a

Silver-Copper Alloy Seal,

-~ - UNCLASSIFIED
‘ | " Y-26241

INCHES

0,02

 

 

 

 

 

Fig. 2.1.24. Interface Between Silver«Copper Alloy and Inconel Base Metal. Unetched. 100X.

75
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
Y-26219

   

 

Fig. 2.1.25. Joint Before and After Welding by
Internal Welding Procedure.

undesirable pores; however, Griscom-Russell
engineers have indicated that repair of o weld to
eliminate porosity has been accomplished by
repeating the fusing of the complete joint.

An important factor to consider with this method
of heat exchanger fabrication is that the tubes
would have to be cut quite accurately to the proper
length. This would be a definite requirement if
the tubes were straight, since any slight mismatch
would inhibit assembly. The use of a bayonet or
U-tube heat exchanger would probably be necessary.
Inspection of the KAPL heat exchanger welds was
accomplished with a 1 x 1.5-mm spheroidal thulium
source (3/4 curie). The source was placed inside
the tube, and film was wrapped around the outside
of the weld. A small quantity of Y%-in.-OD, 0.050-
in.-wall INOR-8 tubing and 34‘-in.-di<:| and 3-in.-dia
INOR-8 bar is being prepared for test welding at
Griscom-Russell.

MATERIAL AND COMPONENT INSPECTION

J. W, Allen R. W. McClung
J. H. DeVan R. A. Nance
G. M. Tolson

Penetrant and radiographic methods were used
for the inspection of 675 ft 5 in. of INOR-8 tubing.
Defects greater than 5% of the wall thickness
were cause for rejection, and on this basis only
13 ft 5 in. of the tubing was rejected. One lot of
approximately 150 ft of seamless Weldrawn INOR-8
tubing that appeared to be of particularly high
quality was inspected by radiographic, dye
penetrant, ultrasonic, and eddy current methods

76

and only 5 ft 8 in. was found to be rejectable.
About five discontinuities were detected by
radiography, but none were found by use of the
encircling-coil eddy-current technique. The ultra-
sonic inspection revealed the presence of several
cracks which had not been detected by any of the
previous inspections, and approximately 3 ft of
tubing was rejected because of these cracks.
Some of these rejected areas were split, and dye
penetrant inspections of the inner surfaces were
made. In each case, longitudinal discontinuities
were found on the inner surface, and therefore
these sections were studied metallographically.
The discontinuities proved to be cracks that
started at the inner surface and proceeded to
depths of up to 0.005 in., as shown in Figs. 2.1.26
and 2.1.27; the cracks were approximately ]/2 to
1 in. in length. No discontinuities were found
on the outer surfaces. In all previous inspections
of small diameter tubing, most of the cracks were
found on the outer surface.

In addition to the INOR-8 tubing, 273 ft of INOR-1,
-2, «3, -5, and -6 experimental alloy tubing was
also inspected. Although it was checked only for
gross defects, it was necessary to reject 95 ft of

this tubing. Of the 75 ft of INOR-8 pipe inspected,

  
    

UNCLASSIFIED
Y.26239

 

 

EEE— -

Fig. 2.1.26. A Typical Large Crack Found by
Ultrasonic Inspection of I/z-in.-ID, 0.045-in,~Wall INOR-8
Tubing., 100X. Reduced 32%.
none was rejected; and of the 37 ft of {NOR-8
bar inspected, 6 in. was rejected.

Two expansion bellows were inspected for high-
temperature sodium service by dye penetrant and
vacuum techniques and were found to be acceptable.

The pump bulkhead for an in-pile test loop was
inspected after it had been machined. Dye
penetrations were found, but, since none of the
penetrotions were in areas where leak tightness
was essential, the part was accepted.

JUNCLASSIFIED
| ¥.26233

 

 

Fign 2.]-27.
Inspection of %+in,-OD, 0.045-in.-Wall INOR-8 Tubing.
100X. Reduced 32%.

Multiple Cracks Found by Ultrasonic

PERIOD ENDING JUNE 30, 1958

A duplex-tube heat exchanger, Inconel over
stainless steel, was inspected in finished form and
found to be acceptable. Fifty-six Hoke valves
were inspected with intended use as the criterion
for the evaluation. Two valves that were rejected
were subsequently repaired, inspected, and found
to be acceptable.

A pump impeller casting of INOR-8 (Fig. 2.1.28),
which was prepared by the Haynes Stellite
Company, was inspected by radiography. Although
the general appearance of the casting was good,
it was rejected because of large pores. Radio-
graphic inspection was also used to check the
spacing of the wires in 636 ft of Calrods, and no
deviations from the predetermined standards were

found.

UNCLASSIFIED
T-15132

   

Fig. 2.1.28, INOR-8 Pump Impeller Casting Made by
Haynes Stellite Co.

77
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

2.2, RADIATION DAMAGE

G. W, Keilhoitz
Solid State Division

IN-PILE DYNAMIC CORROSION TESTS
W, E. Browning

J. E. Lee, Jr. H. E. Robertson
W. H. Montgomery R. P. Shields

INOR-8 Thermal-Convection Loop Assembly
for Operation in the LITR

Assembly work was continued on the electrically
heated INOR-8 mockup of the INOR-8 thermal-
convection loop being constructed for circulating
molten salt fuel in an irradiation facility in the
LITR. Installation of thermocouples and heaters
was completed (Fig. 2.2.1). High-current power
connections for the heaters were made to the fuel

     
  
 

tube for generation of power to simulate fission
power. An automatic control system for the cooling
air was designed ond fabricated that includes
provisions for metering the distribution of air
flow between the two coolant channels on the loop,
A panel was prepared for the individual controls
of the various platinum clamshell heaters that are
installed on the loop. The various components
of the loop are now being joined and thermal
insulation is being installed. When the mockup
has been completely assembled, it will be operated
under conditions simulating those expected in the
in~pile loop in order to demonstrate the workability
of the design and to provide operating information
for use during the inepile experiment,

UNCLASSIFIED
PHOTO 43945

(HHITTHI

 

Fig. 2.2.1. Electrically Heated Mockup of Thermal-Convection Loop for Operation in the LITR.

78
The parts for the fuel system of the in-pile
model have been fabricated, Assembly of these
parts awaifs operation of the electrically heated
mockup. The charcoal to be used in the xenon
adsorber has been prepared for the in-pile experi-
ment by heat treatment in dry helium. Design
details of the outer can for the in.pile loop are
being worked out,

LITR Forced-Circulation Loop Examination

The Inconel forced-circulation loop which circu-
lated a molten salt in a vertical hole in the LITR
for 235 hr at a maximum temperature of 1600°F
and a temperature differential of 230 to 250°F
was examined for corrosion, and the fuel mixture
was analyzed for fission products., The corrosion
was the same as that which would have been
expected under the same conditions in the absence
of radiation,

A  section of Inconel tubing including fuel
(NcF-ZrF4-UF4, 60.8-27.4-11.8 mole %) was cut
from each region that contained a thermocouple,
Several sections were examined metallographically,
and a maximum depth of penetration of 4 mils
was found. The 4 mils of penetration was in a
section of the hot leg of the loop approximately
12 in. from the hairpin nose portion of the loop.
Etched ard unetched pieces of metal from the

UNCLASSIFIED

 

RMG 2143 7
s ,=."E i . ; -~.:: <]
“r
\ * '
L S
(a) . il w- s

PERIOD ENDING JUNE 30, 1958

hot leg of the loop are shown in Fig. 2,2.2, and a
section of tubing from the cold leg of the loop
is shown in Fig. 2,23, The penetration found
in a section of the hairpin of the loop is shown

in Fig, 2.2,4.

UNCLASSIFIED
RMG 2140

 

Fig. 2.2.3. Section of Inconel Tubing from Cold Leg

(1290°F During Operation) of LITR Vertical In-Pile
l.oop. Etched, 250X.

UNCLASSIFIED

   

 

_ K

Fig. 2.2,2. Sections of Inconel Tubing from Hot Leg (1490°F During Operation) of LITR Vertical In-Pile Loop.

(a) Unetched, (b) Etched. 250X,

79
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
RMG 2151

 

Fig. 2.2.4. Section of Inconel Tubing from Hairpin Nose Loop (1510°F During Operation) of LITR Vertical

InePile Loop. Unetched. 250X.

A new fuel salt sampling method, in which the
salt is melted out of the tubing, was tested to
determine whether the salt and the metallographic
specimens were affected. In order to compare the
old and the new techniques of sampling, one sample
of salt was drilled from the tube and a second
sample was obtained by suspending the tube in an
atmosphere of helium and heating it by induction
until the salt dropped out as a pellet.

The results of radiochemical analysis of the
salts obtained by the two methods are given in
Table 2,2,1. The data show that the two sampling
techniques are equivalent as far as radiochemical
results are concerned. Analyses of the fuel
samples for Fe, Cr, and Ni are under way,

In order to test the effect of the melting-out
technique on metallographic samples, specimens
from which the fuel was drilled and from which
the salt was melted were submitted for examination,
A  metallographic sample cut from a piece of
tubing from which the fuel was melted is shown
in Fig, 2.2.5. The temperature was deliberately
allowed to go 200°F higher than necessary during
melting of the fuel, and the specimen was held
at this temperature for 1 min, Normally the salt

80

drops out within 15 sec after the induction heater
is turned on. The longer heating period was used
to exaggerate any effect of the melting-out pro-
cedure. It is apparent in Fig. 2,2.5 that no
significant amount of corrosion occurred during
the melting-out operation.

Table 2,2,1. Results of Rodiochemical Analyses of
Fuel Salt Samples Qbtained by Two Methods
from Inconel Forced-Circulation Loop

Cperated in the LITR

 

Amount Found (me/mg)

 

Constituents

 

Analyzed For In Drilled In Melted
Sample Sample
Z¢73 6.7 x 102 6.7 x 102
cs!37 1.40x 104 1.41x 10~4
Ce'44 6.85 x 10~3 7.65 x 1073
5r89 1.84 x 10™2 1.92 x 1072
5¢70 5.45x 10™¢ 5.68% 10~4

 
 

PERIOD ENDING JUNE 30, 1958

" UNCLASSIFIED
RMG 2136

Fig. 2.2.5. Section of {nconel Tubing Held at 1390°F for 1 min During Melting-Out of Fuel Sample, 250X.

ORR Forced-Circulation Loop Development

Preparations are being made for the operation
of a forced-circulation loop in an ORR facility.
The loop assembly housing and lead tube were
tested in the reactor and were found to be capable
of withstanding the force of the flowing reactor
cooling water. Flux profile measurements are
under way at reduced reactor power levels to
determine the effect of a poison load simulating
that of the in-pile experiments.

Studies directed toward the development of
bearings which will withstand the radiation con-
ditions surrounding the pump of the in-pile loop
are under way. An ordinary carbon-steel ball-
bearing assembly modified to use a silver separator
and a ball-bearing assembiy that has special
titanium carbide balls are to be tested out-of-pile
without lubrication,

Removal of xenon from the atmosphere above
the fuel by circulating helium sweep gas through
the pump tank, through a charcoal holdup trap,
through the pump motor can, and down the shaft
annulus to the pump tank to complete the cycle is

being considered. Removal of the xenon and
tission-product gases and recirculation of the
purified sweep gas to prevent upward diffusion
of the fission gases may prevent damage of the
lubricant in the bearings, Experimental tests and
calculations are planned for evaluating the proba-
bility of the success of this arrangement.

The motor of the pump used in in-pile loops is
being redesigned to increase the available torque
and to eliminate a universal joint which is judged
to be a possible source of trouble.

IN-PILE STATIC CORROSION TESTS
H. L. Hemphill

Inconel capsules were assembled for testing
the stability of graphite in contact with molten-
salt fuel (Fig. 2,2.6) and were shipped to the
MTR for irradiation at 1250°F. INOR-8 capsules
were fabricated for similar irradiation and are
being assembled. The metal stock from which
these capsules were made was selected from
tubing that contained flaws, since flawless tubing
of the proper dimensions was unavailable,

W. E. Browning

8]
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED
ORNL-LR-DWG 30083

 
     
   
   
 

ALL DIMENSIONS ARE IN INCHES

INNER SHELL

-in. 0D x 0.150-in.1D BARREL
(INCONEL)

0.375-in.0D x 0.350-in.ID SHELL
{347 STAINLESS STEEL)

PLUG{HI-D GRAPHITE)

3 HOLES 120° APART

0.150-in. 0D x 0.100-in. 1D INNER SHELL
{HI-D GRAPHITE)

 

 

 

DRILL THRU, 2 HOLES PLUG  PLUG, ,SHELL ,BARREL _-INNERLINER
/90°APART FOR PINS / -J( / /
T L L 4_iu_1 z >y &
k A— -
— T e =T

 

 

 

 

SR

 

 

\oven. -

WELD ALL AROUND  SILVER SOLDER/\‘/.s—in.DFnLL, 3 HOLES
120 ° APART FOR PINS

SILVER SOLDER

ASSEMBLY

Fig. 2.2,6. Capsule for Testing the Stability of Graphite in Contact with MoltensSalt Fuel Under lrradiation in
the MTR.

82
PERIOD ENDING JUNE 30, 1958

2.3. CHEMISTRY
W. R. Grimes

Chemistry Division

PHASE EQUILIBRIUM STUDIES
Systems Containing UF, and/or ThF,

R. E. Thoma H. A. Friedman
H. Insley C. F. Weaver

The System LiF-Ber-ThF4-UF4. — Detailed
studies are being made of the phase equilibria
characteristics of the quaternary system LiF-
BeF ,-ThF ,-UF, in the composition region 26
to 40 mole % BeFZ, 1 to 3 mole % UF,, and
1 mole % ThF ,. A fused-salt breeder-reactor fuel
having a liquidus temperature of 440 + 5°C and
containing no more than 36 mole % BeF, can
be chosen from this composition region. Liquidus
curves for some LiF-BeF,-ThF ,-UF , compositions
in this region are shown in Fig. 2.3.1.

The System LiF-BeF,-ThF,. — A partial phase
diagram of the system LiF-BeF,-ThF, was pre-
senfed previously,! and a more complete phase
diagram based on inferences from the best thermal
analysis and quenching data presently available
is shown in Fig. 2.3.2. Recent results of optical
and x-ray diffraction examinations of quenched

 

R, E. Thema et al., MSR Quar. Prog. Rep. Jan. 31,
1958, ORNL-2474, p 81,

UNCLASSIFIED
ORNL-LR -DWG 29668R

 

 

 

  
 
 
 
  
   

x 1 mole % UF, +1mole % ThFy

  
  
 
  
 
   
 

g0 } N\ L ® 2mole % UF, + 1mole % ThF, #
. o> 3mole % UF, +1mole % Thi,
| |
455 ‘o — e - —_—
S | : 2LiF-BeF LIQUID | “
2. + LiQUID . J
L b o o o e
x 450 | iU id i =
S o] e —
[ | |
w445 re— T TTLIF B, TR T T T
5 | +L1Quip \
[ % !
- o LiQuio. b
440 - 2LiF BeF ‘ |
+ LIQUID . | ‘
4 3 5 },7 — — 77 ——— e—— — _ I - _— ,,7{ P _I‘_ __[
ol 1]
30 3 32 33 24 35 36 37 38 39
BeF, (mole 7o)
Fig., 2.3.1. Liquidus Temperatures in the System

LiF'BEFz'UFA“Th F4.

samples indicate that the following three eutectics
occur within this fuel solvent system:

Composition (Mole %)

 

Melting Point (°C)

LiF BeF2 ThF4

66 29 5 427
65 30 5 429
47 51.5 1.5 ~360

Two experimental factors preclude the immediate
construction of an unequivocal polythermal phase
diagram.  First, LiF-ThF, solid phases do not
appear in the ternary system LiF-BeF ,-ThF, as
pure phases; that is, they contain BeF,. The
type of solution which occurs is not, as yet,
known. Since no LiF-ThF, compounds occur as
pure phases in the ternary system, all Alkemade
lines become areas in the polythermal diagram
rather than lines and thus lose some significance
relative to defining compatibility triangles.

Second, no hydrolysis products have been ob-
served in any LiF-BeF.,-ThF, samples, although
these materials convert UF , into UD, so readily
in LiF-BeF ,-ThF ,-UF, mixtures that stringent
measures are required to ensure their absence.

UNCLASSIFIED
CORNL-LR-DWG 29666

      

7 LiF-SThF‘r\ L-LiF-2Thf,
\. ‘

  
 

3 LiF-ThFy,-

s

2LiF-BeF,” LiF-Befp~

Fig. 2.3.2. The System LiF-BeF,-ThF,.

83
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Previous reports '+ of results of LiF- -BeF ,-ThF
quenches showed consistent liquidus and secondary
phase temperatures and rarely reproducible solidus
temperatures. |t can be inferred that such vari-
ations in solidus values are indicative of the
presence of a quaternary liquid containing hy-
drolysis products in solution. The most recent
quench data for mixtures with compositions close
to that of the 2LiF.BeF,-7LiF.6ThF 4-LiF-2ThF
eutectic and to that of 'rhe 2LiF.BeF -7L|F 6ThF ,-
3LiF-ThF, eutectic were obtcnned with somples
in whlch only prehydrofluorinated BeF, was
used. The results from these data are much more
self-consistent than those obtained with as-
received BeF .

The System LiF- ThF,~UF,. — In quaternary
mixtures of LiF-BeF -ThF UF4, solid solutions
of LiF-ThF, and L|F UF
cipitated from molten
secondary phases.

compounds are pre-
|iquids as primary or
An understanding of the
UF“-ThF4 phase behavior in molten-salt reactor
fuels whose compositions are chosen from the
quaternary system LiF-BeF ,-ThF ,-UF, is there-
fore dependent on a rellobiy accurofe L|F ThF ,-
UF , phase diagram,

A comprehensive study of the phase equilibria
in the system LiF- ThF ,-UF, has been made,
and the phase diagram for 'rhe system is pre-
sented in Fig. 2.3.3. A report giving complete
results of the thermal analysis and of optical
and x-ray diffraction examinations of quenched
samples is being prepared.

The salient characteristic of the system LiF-
ThF ,-UF,, as would be expected, is the extensive
formation of solid solutions. In the course of
the investigation detailed studies were made
of methods for the precise determination of UF,
and ThF, concentrations in primary phases. |t
was found that concentrations precise to within
+0.5 mole % could be determined with the petro-
graphic microscope for UF , or ThF ,.

The supplemental results obfcmed from recently
completed thermal-gradient quenching experiments
were consistent with the results presented pre-
viously,! and liquidus isotherms could be drawn
for the entire system. Also, data from the recent
experiments defined the phase boundaries of the

 

2R, E. Thoma, Resulis of X-Ray Diffraction Phase
Analyses of Fused Salt Mixtures, ORNL CF-58-2.59
(Feb. 18, 1958). .

84

3LiF-ThF, compound and established the solu-
bility limit of UF, in this phase at the solidus.
The compound 3LiF-UF , has no region of stability
in either the binary system LiF-UF4 or the
ternary system LiF-ThF -UF,. It is probably
isostructural with 3LiF.ThF ,. Solubility studies
of mixtures with the composition 3LiF.UF,-
3LiF-ThF , (75mole % LiF) show that 3LiF . Th(U)F,
solid solutions containing as much as 15.5 mole %
UF, may be formed. This solid solution is green,
is biaxial negative, and has a birefringence of
approximately 0,008, with an optic angle of
approximately 30 deg. The indices of refraction
of 3LiF-Th(U)F4 as a function of UF,
are shown in Fig. 2.3.4.

content

Solubility of PuF, in Alkali Fluoride~Beryllium
Fluoride Mixtures

C. J. Barton R. A. Strehlow

Further information was obtained on the solu-
bility of PuF in alkali fluoride—beryllium fluoride
mixtures, During the quarter, additional data
were obtained for several of the solvent com-
positions previously tested and for one new
solvent mixture. Some of the previously reported
solubility values were changed slightly as the
result of repeated analyses, and, in addition,
the solvent compositions were more accurately
defined. All the data obtained to date are given
in Table 2.3.1 and are shown graphically in
Figs. 2.3.5 and 2.3.6. The values marked with

asterisks in Table 2.3.1 are believed to be in-

correct, and therefore they were omitted from
the graphs. The data® indicate that the solu-
bility of PuF, in LiF-BeF, mixtures is at a

minimum for mixtures containing about 63 mole %
LiF and that it is at @ minimum in the NaF-BeF,
system for mixtures containing about 57 mole %
NaF. Comparison of PuF, solubility data with
the results of CeF, solubility determinations?
showed that CeF, was slightly more soluble
than PuF, for the same solvent composition and
temperature. A study of the effect of CeF, on
the solubility of PUF3 in LiF-BeI:2 (63-37 mole %)

is under way.

 

3c. 4. Barten, W. R. Grimes, and R. A, Strehlow,
Solubility and Stability of PuF3 in Fused Alkali Flu-
oride—Bery!lium Fluoride Mixtures, ORNL<2530 (June 11,
1958).

“W. T. Ward, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 94, and subsequent section of this
chapter.
PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED
ORNL~-LR-DWG 28215R

 

ThF,
114
/
/
LiF-4ThF TEMPERATURES IN °C
4 7
0
%
77
0
%
LiF-2ThF, A
9
& ke
P890 g
/N s,
o, o
/N ‘
N/
7LiF-BThF, , ROS
\\
L ? R\
vo & D
P595{ 65, % %
E560 Ad‘
O
3LIF-ThF, B
9,
605 % /
E5704 o ~
‘ 6
C 9
5 & S 3 %
/
/N D 8D e A %
@ % N 0 0
3% R A\E \uos 6. &
% O Q \ %
% /N
7 | ,
LiF 500~ “490 610 770 LiF- 4 UF UF,
845 4LiF-UR, 7LiF-6UF, 4 1035

Fig. 2.3.3. The System LiF-ThF,-UF,.

Because there is interest in the possibility BeF, (65.4-34.6 mole %), obtained from Table
of converting Th232 to fissionable U2?3® in a 2.3.1, are compared with data for PuF, solubility
plutonium-fueled fused-salt reactor, a brief study in LiF-BeF-ThF, (62.4-36.6-1.0 mole %, cal-
was made of the effect of 1 mole % ThF, on the culated composition). A plot of solvent com-
solubility of PuF, in LiF-BeF, (63-37 mole %). position vs PuF, solubility3 indicated that the
The results of the study are presented in Fig. difference in the LiF-to-BeF, ratios of the two
2.3.7, in which data for PuF, solubility in LiF- solvents would have very little effect on the

85
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

INDEX OF REFRACTION

1.540

1.530

1.520

1.510

1.500

1.480

1.480

Fig.

UNCLASSIFIED
ORNL-LR-DWG 31247

—

 

 

 

 

 

 

 

MAXIMUM EXTENT OF —
3LiF - ThF,
SOLID SOLUTION

 

 

 

 

¢

 

 

 

 

 

O o0—

 

UF, CONTENT (mole %!}

2,3.4,

Solid Solution.,

PuFy (mole % )

o
o

0.2

oA
10

Fig.

 

UNCLASSIFIED
ORNL—LR—DWG 29605
TEMPERATURE (°C)
600 550

650 500

74.3 LiF-28.78eF,

|
51,7 LiF ~ 48.3 BeF,

65.4 LiF~34.6 BeF,

56.3LiF-43.7Bef,

5 11.0 1.5 12.0 12.5 13.0 13.5 14.0
10,000 /7 (°K)
2.3.5. Solubility of PuF; as a Function of

Temperature for LiF-BeF2 Solvents,

86

Indices of Refroction for 3LiF-ThF4

PuF, (mole %}

UNCLASSIFIED
ORNL—LR—DWG 29606

650 550 500

 

TEMPERATURE (°C)

 

600
_l_ T ] i{
565NGF 1?5LnF 26BefF, |
1

|
|
|

  

 

|

 

 

 

 

 

 

o
Gn

 

L
57.0 NaF-43.0Bef;~

5

 

 
 
 

 

   

 

 

 

 

 

 

 

 

0.2 L ~
~~\4
| \ 49?NGF 50.3BeF;,
\
o4 [
10,0 10.5 11.0 11.5 12.0 12.5 13.0 13.5
10%/°K
Fig. 2.3.6. Solubility of PuF; as a Function of

Temperature for NoF-BeF, and NoF-LiF-BeF, Solvents.

UNCLASSIFIED
ORNL-LR-DWG 31220

 

TEMPERATURE(°C)
650 600 550 500
1.0
0.8
06
3
@
g 04
>
=
—
@
o
|
® o2
u”? o SOLUBILITY IN LiF-BeF,
a (65.4 - 366m0|e%)
® SOLUBILITYIN LiF- BeF UF
(62.4-36.6-1mole%)
04
0.08
10.0 10.5 1.0 1o 2.0 12.5 130 135
10%/°K
Fig. 2.3.7. Effect of Addition of ThF, on Solubility

of PuF3 in an LiF-BeF2 Mixture,
PERIOD ENDING JUNE 30, 1958

Table 2.3.1. Solubility of PuF, in Alkali Fluoride~Beryllium Fluoride Mixtures

 

Solvent Composition

 

 

 

(mole %) Fi|trcfionoTemperdfure Filtrate Analysis
NoF LiF BeF, v Pu lwt %) PuF3 (mole %)
49.5 50.5 552 1.18 0.22
600 1.73 0.33
600 1.79 0.34
651 2,72 0.52
57.0 43.0 538 1.17* 0.22*
552 1.63* 0.31*
600 1.35 0.26
600 1.26 0.24
609 1.26 0.24
650 1.48* 0.28*
652 2.21 0.42
706 3.40 0.66
63.4 36.6 550 1.54 0.29
598 2.43 0.46
600 2,00* 0.38*
650 4.40 0.85
51.7 48.3 463 1.02 0.16
549 2,44 0.38
599 2.96* 0.47
654 5.76 0.93
56.3 43.7 494 1.04 0.15
560 1.89 0.28
602 3.15 0.48
653 5.47 0.86
550 1.98 0.30
599 2.04* 0.31*
649 6.24* 0.95*
65.4 34.6 532 1.15 0.16
600 1.78 0.27
643 4.30 0.63
713 28.7 546 4,00 0.56
597 5.90 0.85
650 8.48 1.26
56.5 17.5 26.0 500 2.92 0.51
554 7.68 1.43
565 2.61* 0.46*
600 7.59 1.41
634 13.0 2.58
655 6.45* 1.18*

 

*Daubtful results excluded from Figs. 2.3.5 and 2.3.6.

87
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

solubility of PuF,, but it is expected that the
solubility of PuF, in pure LiF-BeF, (63-37
mole %) will be determined in the near future in
order to obtain @ more direct comparison. Al-
though the scatter of the data shown in Fig.
2.3.7 precludes firm conclusions, it seems likely
that the addition of 1 mole % ThF, to an LiF-
BeF, (63-37 mole %) mixture will not appreciably
affect the solubility of PuF, in the solvent.

FISSION-PRODUCT BEHAVIOR
G. M. Watson F. F. Blankenship

Solubility of Noble Gases in Molten
Fluoride Mixtures

N. V. Smith

Studies were made of the solubilities of argon
in NaF-LiF-KF (11.5-42-46.5 mole %) and of
helium in LiF-BeF, (64-36 mole %), and numerical
values expressed as Henry's law constants are
presented in Tables 2.3.2 and 2.3.3 and Fig.
2.3.8. For comparison, the previously reported®
solubility constants of helium and neon are
also shown in Fig. 2.3.8.

 

SN. V. Smith, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL.-2474, p 91.

The trends of the data are the same as those
previously observed for the solubilities of the
noble gases in mixtures containing ZrF,. The
solubilities follow Henry's law, increase with
increasing temperature, and decrease with in-
creasing molecular weight of the gas. However,
an interesting difference in solubility is ex-
hibited by helium in the LiF-BeF, mixture. The
heat of solution of helium is considerably lower
in LiF-BeF, than:in NaF-KF-LiF or in mixtures
containing ZrF, ond is not approximated by the
product TAS calculated from the gas law for
isothermal expansion of an ideal gas from an
initial concentration, C , at the temperature and
pressure of saturation, to a final concentration,
C o that is numerically equal to the measured
solubility.  The entropy of solution diminished
by the entropy of expansion of an ideal gas
over the same concentration range, Cg to C
is considerably larger in the mixture containing
BeF, than in any of the molten fluoride mixtures
previously studied. The heats and entropies of
solution for helium, neon, and argon in NaF-

KF-LiF and for helium in LiF-BeF,, as well

 

®). H. Shaffer et al., MSR Quar. Prog. Rep. Oct. 31,
1957, ORNL-2431, p 41.

Table 2.3.2. Solubility of Argon in NaF-KF-LiF (11.5-42-46.5 Mole %)

 

 

Temperature Saturating Pressure Solubility .
(°C) {atm) (moles of urgon/cm3 of melt) K
x 108 x 10~8
600 1.01 0.834 0.825
1.50 1.378 0.917
2,02 1,929 0.957
Av  0.90 0,05
700 1.00 1.744 1.744
1.50 2,705 1.799
2,04 3.811 71.87]
Av 1.80 £0.04
800 1.004 3.4 3.397
1.51 5.161 3.422
2.033 6.842 3.365

Av 3.40 £0.02

 

*K = ¢/p in moles of gas per cubic centimeter of melt per atmosphere.

88
PERIOD ENDING JUNE 30, 1958

Table 2,3.3. Solubility of Helium in LiF-BeF2 (64-36 Mole %)

 

 

Temperature Saturating Pressure Solubility .
°C) (atm) (moles of he lium/em? of melt) K

x 10~8 x 108

500 1.009 7.46 7.39
1.521 11.46 7.53

1.996 15.08 7.56

Av 7.49

600 1.103 11.74 11.59
1.570 18.76 11.95

2,107 22,40 11.1

Av 11,55

700 1.004 14,81 14,75
1.526 23,75 15.56

1.963 28,41 14,47

Av 14.93

800 0.996 19.36 19.47
1.52 29,63 19.49

1.978 38.55 19.49

AV 19149

 

*K = c/p in moles of gas per cubic centimeter of melt per atmosphere.

as the corresponding TAS products, are given
in Table 2.3.4. As may be observed, the heats
of solution in NaF-KF-LiF can be estimated to
within 10% by considering the gas to be ideal
and calculating the entropy of isothermal ex-
pansion over the concentration interval defined
by a single experiment. This is not the case,
however, for helium in LiF-Ber. The fundamental
significance of these observations is not imme-
diately apparent, and efforts to interpret the data
will be deferred until additional measurements
have been made.

Solubility of HF in LiF-BeF, Mixtures
J. H. Shaffer

The investigation of the solubility of HF in
LiF-BeF,

composition

mixtures as a function of solvent
in the range 0-55 mole % BeF,

was almost completed, and, in order to avoid
repetition, it is planned to withhold presentation
of the results until the measurements are com-
pleted. It may be mentioned that, in mixtures
with low alkali fluoride content, the solubility
of HF in LiF-BeF, mixtures is approximately the
same as in NaF-ZrF, mixtures. As the alkali
fluoride content is increased, however, the solu-
bilities of HF in the two mixtures differ markedly.
For example, in mixtures containing about 90
mole % alkali fluoride, the HF solubilities are
about tenfold lower in the LiF-BeF, system
than in the NaF-ZrF, system. This comparison
is based on the inferpolation described previously”
of available data for the NaF-ZrF, system.

 

7). H. Shaffer, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 93.

89
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

UNCLASSIFIED

ORNL-LR-DWG 31221
TEMPERATURE (°C)

(x1079) 800 700 600 500
1000

HELIUM IN NaF-KF-LiF

100 HELIUM IN
iF-BeF,

NEON IN
aF-KF-LiF

K [moles OF GAS/(cm® OF melt) (atm )]

IN
NaF-KF-LiF

 

104/ °K

Fig. 2.3.8. Solubilities of Noble Gases in NaF-KF-
LiF (11,5-42-46,5 Mole %) and in LiF-BeF2 (64-36
Mole %).

Solubilities of Fission-Product Fluorides in
Molten Alkali Fluoride—~Beryllium
Fluoride Solvents

W. T. Ward

Determinations of the solubility of CeF, over
the temperature range of 450 to 700°C were com-
pleted for both LiF-BeF2 and NaF-BeF, solvents
ranging in composition between 50 and 70 mole %
alkali fluoride. It was found that the solubility
of CeF, passed through a minimum at approxi-
mately 62 to 63 mole % alkali fluoride in both
solvents. The solubility values were somewhat
in NaF-BeF, than in LiF-BeF, of cor-
responding composition, as may be seen in Fig.
2.3.9, in which the solubilities in terms of weight
per cent cerium are plotted as functions of
solvent composition for three different
peratures. The same solubility values in terms
of mole per cent CeF, in the filtrates are shown
in Fig. 2.3.10.  Solubility values taken from
these curves are listed in Table 2.3.5.

less

tem=-

Chemical Reactions of Oxides with Fluorides
in LiF-KF

J. H. Shaffer

As shown previously,® it is possible to effect
some chemical separations of solutes in molten

 

83. H. Shaffer, MSR Quar. Prog. Rep. Janm 31, 1958,
ORNL-2474, p 99.

Table 2.3.4. Enthelpy and Entropy Changes Occurring upon Solution of Some Noble Gases
in Molten Fluorides at 1000°K

 

Solution

 

 

ldeal Gas
Sol vent Gas
AH (cal/mole) AS (ev) Expansion TAS
NaF-KF-LiF Helium 8,000 -0.25 8,250
(11.5-42-46.5 mole %) Neon 8,900 - 0.97 9,870
Argon 12,400 -0.10 12,500
LiF-BeF2 Helium 4,850 ~3.73 8,580

(64-36 mole %)

 

90
PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED

o UNCLASSIFIED
ORNL—LR—DWG 3t222 ORNL~LR—DWG 24223

 

] [ { l ]

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

|
glf T
2.2
8 ! 2.0
7 / 1.8
C
/ -~ 1.6
_6 o
S € 14
E 700°C Z';
5 ° W
E \\/ R
5 / | E
Da \ = 10
= z
- / 700°C o
o / | 8 os
o3 }r
%Oooc \J/ J 0.6
> . A 600°C _|
N
0.4
4 \'X 'X\ Ly 0.2 o
‘x‘*L"""‘*soo"c \&"w.. 500°C 500°C
‘ *Tex=x= 500°C o
0 ’ 50 60 70 80 50 60 70 80
50 60 70 80 50 & 70 80 LiF IN LiF-BeF, SOLVENT  NoF IN NaF-BeF, SOLVENT
LiF IN LiF-BeF, SOLVENT  NaF IN NaF-BeF, SOLVENT (mole %) (mole %]
(mole %) A
Fig. 2.3.9. Solubility of Ce‘sF3 as Ce (wt %) in Alkali Fig. 2.3.10. Solubility of CeF, as CeF, (Mole %) in
Fluoride—Beryllium Fluoride Solvents, Alkali Fluoride—Beryllium Fluoride Solvents.
Table 2.3.5. Solubility of Cer* at 500, 600, and 700°C in Nc:F-BeF2 and LiF-BeF, Solvents
as a Function of Solvent Composition
Alkali At 700°C At 600°C At 500°C
ali
Fluoride In NaF-BeF2 [n LiF-Bo;-.F2 In NaF-BeF, In LiF-BeF, In NaF-BeF, In LiF-BeF2
in Solvent CeF, Ce CeF, Ce CeF, Ce CeF, Ce CeF, Ce CeF,
(mole %) (wt %) (mole %) (wt%) (mole%) (wt%) (mole%) (wt%) (mole®) (wt%) (mole®) (wt%) (mole%)
50 3.6 1.25 5.9 1.65 1.84 0.60 2.9 0.79 0.87 0.29 1.3 0.31
55 2.8 0.91 5.2 1.40 1.33 0.41 2.3 0.61 0.58 0.19 0.86 0.22
60 2.15 0.72 4.7 1.24 0.95 0.30 2.00 0.51 0.38 0.12 0.68 0.168
63 2.13 0.74 4,66 1.18 0.82 0.26 1.97 0.48 0.32 0.093 0.66 0.162
65 2.55 0.85 5.1 1.28 0.88 0.28 2.06 0.50 0.30 0.076 0.70 0.166
70 6.2 1.95 7.7 1.9 2.4 0.77 3.2 0.79
75 1.07

 

*The experimental error of the measurements is estimated to be +5%.

91
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

LiF-KF by selective precipitation as oxides.
Relatively sharp separations of uranium from
cerium and of zirconium from cerium were demon-
strated by the stepwise addition of solid CaO
to solutions containing UF ,-CeF, and ZrF,-
CeF, in molten LiF-KF. It was also found,
however, that CeF , and BeF , coprecipitated from
the LiF-KF solvent containing BeF,-CeF, when
Ca0 was added. It appeared to be pertinent,
therefore, before attempting investigations of
LiF-BeF, solvent to determine the
characteristics of BeO as a precipitating agent
for uranium, zirconium, and cerium from solutions
made up with the simple solute LiF-KF. [t would
be desirable, of course, to use BeO as a pre-
cipitating agent in mixtures containing BeF,,
since it would not introduce any foreign con-
stituents to the solution.

The results of an experiment in which BeO
was used as the precipitating agent for zirconium
are presented in Fig. 2.3.11. The concentration
of zirconium remaining in the LiF-KF solvent
is shown as a function of time. A filtrate was
obtained before addition of a 20-fold excess of
BeO pellets, and filtrates were obtained at
intervals after the addition. As may be seen,
the precipitation of the zirconium took place in
]/2 hr or less after the addition of the BeO, but
several hours were required to reach the stoichio-
metric concentration of beryllium in the solvent.
On the basis of this experiment alone, it is not
possible to state why the concentration of beryl-
lium takes some time to build up to the stoichio-
metric value, but it may be surmised that zirconium
oxide forms as a layer over the pellets of BeO
and that it therefore takes time for the BeF,
to diffuse out. It is interesting to note that the
the free energies of formation of the pure solids®
show a positive free-energy change of +16.6 kcal
for the reaction

mixtures,

2BeQ + ZrF4 \-_—" ZrO2 + QBeF2 .

The partial precipitation of cerium that occurred
upon addition of 20-fold excess of BeO to an

 

A. Glassner, A Survey of the Free Energies of
Formation of the Fluyorides, Chlorides, and Oxides of
the Elements to 2500°F, p 6, 20, ANL-5107 (Aug. 1953).

92

UNCLASSIFIED
ORNL~LR-DWG 34224

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

80
e
70 ¢ —T
L~ ;
.
&0
© |
2 3
5
5 50 :
» !
5 ;
o |
S |
‘% 40 H i |
o ] i f
@ |
£ |
o !
% 0 ZIRCONIUM REMAINING IN SOLUTION
E 30 | {DETERMINED RADIOCHEMICALLY)
N @® BERYLLIUM REMAINING IN SCLUTION
,<_[ {DETERMINED CHEMICALLY )
- | |
20 ! [
! i
10 ‘
|
\
o SO~ 0 ‘ L o
0 5 10 15 20 25

TIME (hr)

Fig. 23,11, Results of Reaction of ZrF, (Hf'8'F )
with BeO in LiF-KF (50-50 Mole %).

LiF-KF mixture containing CeF, is indicated in
Fig. 2.3.12. This experiment confirmed the
previous observation® that the oxides of cerium
and beryllium coprecipitate. Since it appeared
in this experiment that the reaction was surface
controlled, an additional 30-fold excess of BeO
was added after heating the solution to 900°C.
{t was surprising to find after this drastic treat-
ment that the cerium had not completely pre-
cipitated and that its concentration had only
decreased from 57 to 36 meq/100 g of solution.
No satisfactory explanation of this behavior is
immediately apparent.
UNCLASSIFIED
ORNL-LR-DWG 31225

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

140
O CERIUM REMAINING IN SOLUTION
¢ (DETERMINED RADIOCHEMICALLY)
120 | ® BERYLLIUM REMAINING IN SOLUTION ___|
{DETERMINED CHEMICALLY)

= 100
2
5
©
o 80
o
Q ®
S
g
[}
E
a 60
Z O—
O
[T
i
<1
',_-
£ 40

20

o
0 5 0 15 20 25

TIVE (hr)

Fig. 2.3.12, Results of Reaction of CeF3 with BeQO
in LiF-KF (50-50 Mole %) at 600°C.

CHEMISTRY OF THE CORROSION PROCESS
G. M, Watson G. J. Nessle
Activity Coefficients of NiF, in LiF-BeF,
C. M. Blood

The activity coefficients of NiF, dissolved in
a molten mixture of LiF-BeF, (62-38 mole %) are
being determined by using techniques described
previously, 10.11 The results of experimental
measurements at 600°C are summarized in Table
2.3,6, which gives the experimentally determined
partial pressures of HF and H,, the mole fractions
of NiF, in the melt, and the equilibrium quotients.

 

1OC. M. Blood, W. R. Grimes, and G. M. Watson,
Activity Coefficients of Ferrous Fluoride and of Nickel
Fluoride in Molten Sodium Fluoride—Zirconium Flu-
oride Solutions, paper 75, Division of Physical and
Inorganic Chemistry, 132nd Meeting of the American
Chemical Society, New York, Sept. 8~13, 1957.

Ne, M Blood, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 105,

PERIOD ENDING JUNE 30, 1958

The equilibrium quotients are also shown graphi-
cally in Fig. 2.3.13 as functions of the mole
fraction of NiF, in the melt.

An examination of the results indicates that,
within the experimental precision achieved, the
equilibrium quotients are independent of the mole
fraction of NiF, over a range from approximately
0.9 x 10=% to 21 x 1074, The activity coeffi-
cients of NiF, in this solvent, calculated in the
manner previously described, '%1! have values
of 2347 and of 515 with respect to the solid and
liquid standard states, respectively. Brewer's'?
tabulation of thermodynamic properties was used
in the calculations, and it is now known that the
assumed melting point for NiF, is seriously in
error. However, additional efforts to refine the
values of the calculated activity coefficients will
not be made until it is possible to obtain some
experimental values for the free energies of
formation of pure crystalline NiF, at high temper-
atures (500-600°C) by measuring the equilibrium
quotients by the present method at concentrations
approaching saturation. An attempt to obtain such
values at 500°C is under way.

Solubility of Nin in LiF-—-BeF2 (61-39 Mole %)
C. M. Blood

Measurements were made in order to establish
a concentration limit below which NiF, does not
precipitate from LiF-BeF, os a complex compound
or in the pure state. Experimental procedures
were used in this investigation that were similar
to those used previously'® to determine the
solubility of FeF, in LiF-BeF,. The results are
summarized in Table 2.3.7 and are shown graphi-

cally in Fig. 2.3.14,

The solubility of NiF, was found to be inde-
pendent of the amount of NiF, added. Further-
more, petrographic examination'? of powdered
samples of the filtrates revealed that the satu-
rating phase was pure NiF,. Accordingly, the
activity coefficients of solid NiF, at saturation
in this solvent are given directly by the reciprocal
of the solubility expressed as mole fraction (see

 

12| | Brewer et al., Natl. Nuclear Energy Ser. Div.
IV, 19B, 65, 110, 202 (1950).

13R. J. Sheil, MSR Quar. Prog. Rep. Jan. 31, 1958,
ORNL-2474, p 107.

4R, A, Strehlow, personal communication.

93
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Table 2,3.6. Equilibrium Quotients at 600°C of the Reaction Nin(d) + H,(g) -‘—_A Ni(s} + 2HF(g)
in LiF-BeF2 (62-38 Mole %)

 

 

Pressure of HF Pressure of H2 NiF2 in Melt .
(atm) (atm) (mole fraction) K
x 10~2 x 10~4 x 10™4
0.283 3.89 1.418 1.45
0.292 3.84 1.302 1,71
0.288 3.86 1.273 1.69
0.286 3.87 1,273 1.66
0.244 4.10 0.897 1.62
0.233 4.16 0.897 1.46
0.242 4,11 0.897 1.59
0.265 3.98 0,955 1.85
0.250 4.06 0.926 1.66
0.259 4.02 0.949 1.76
0.319 3,69 1.504 1.83
0.322 3.67 1.504 1.88
0.307 3.76 1.620 1.55
0.323 3.67 1.620 1.76
0.454 2.96 3.963 1.76
0.456 2.95 4,050 1.74
0.448 2.99 4,165 1.61
0.457 2.94 4,107 1.73
0.737 1.43 19,15 1.98
0.734 1.44 19.75 1.89
0.749 1.36 20.74 1.99
0.744 1.39 20.54 1.94

AV ].73 io.]z

 

*Kx = PaF/(XNiF2PH2)‘ where X is mole fraction and P is pressure in atmospheres.,

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

UNCLASSIFIED
(x 10%) ORNL—LR—DWG 31226
4.0 l ‘ l ‘
35 1 K, =(1.73£0.42) x 10% . - : !
Kols)= T7.37 - ( i‘
x 30 Kyl1)=336 . o ?
7 ¥ (s) = 2347
E y (1} = 545 |
5 | | i
]
e ~ e
:/ ./ ‘/_/ 2 e T 7 7 7 7 7 . s
= .-.‘{ e T | | — I T —
o 1.5 | —« o [ [ [ |[ o ‘ .
e ! ‘ ‘ ‘
2 ‘ | !
C 10 i " ‘ i 8 -k R
| | : ;
| | |
o0 | | | r | |
0 ’ | ! . 4
] 2 4 6 8 10 12 14 16 18 20 22 (x10 )

Nifp IN MELT {mcle fraction}
Fig. 2.3,13. Equilibrium Quotients for the Reduction of NiF, by H, in LiF-BeF, (62-38 Mole %).
94
PERIOD ENDING JUNE 30, 1958

Table 2.3.7. Solubility of NiF, in LiF:-BeF'2 (61-39 Mole %)

 

 

 

Temperature Ni Added Solubility of NiF, Activity Coefficient of NiF ,(s)
(° Q) (wt %) (mole fraction) at Saturation
694 2.85 0.0171 58.5
649 2.14 0.0128 78.1
597 1.33 0.00783 127.7
551 0.90 0.00529 189.0
499 0.58 0.00337 296.7
omNeLAsSIFIED Experimental Determination of Chromium Diffusion
TEMPERATURE (°C) Coefficients in Molten Salt—Inconel Systems
5 152 20 600 500 R. B. Evans R. J. Sheil

 

2
&
:g AH=12,000 cal /mole
E 102
>
=
-
m
>
-]
c 5
0
o
L
=z

2

1072

10.0 10.5 14.0 11.5 +12.0 12.5 £3.0 13.5

10%/°K

Fig. 2.3.14.
Mole %).

Solubility of NiF, in LiF-BeF, (61-39

Table 2.3.7). These activity coefficients com-
bined with the equilibrium quotients at saturation
obtained from reduction experiments will be useful
in the experimental determination of free energies
of formation of pure crystalline NiF, at high
temperatures.

The available data on diffusion coefficients for
chromium in nickel-base alloys may be divided
into two distinct groups. The first group consists
of high-temperature data (T > 900°C) which were
obtained through classical self-diffusion measure-
ments based on labeled-chromium (C¢3') distri-
butions within alloy specimens. > No molten
salts were involved. The second group consists

of data based on the exchange

(1) Cro+ Cr*F2 ;B CrF2 + Cro

AF°=0and K =1,

which tokes place when labeled and unlabeled
chromous fluorides are dissolved in a molten salt
contained by Inconel. Rates of the depletion of
Cr*F, with time and the corresponding Cr®
distribution in the Inconel were measured at
relatively low temperatures (800 to 600°C),'¢
Diffusion data at these — and even lower — temper-
atures are important from the standpoint of esti-
mating the to be
expected in molten-salt power reactor components.

A series of experiments for rechecking the data
of the second type, which consist of values

long-term corrosion rates

obtained at four temperatures, and to extend the

 

15p, L. Gruzin and G. B. Fedorov, Doklady .Akad.
Nauk. S.S.5.R. 105, 264 (1955).

V6R. B. Price et al., A Tracer Study of the Transport
of Chromium in Fluoride Fuel Systems, BMI-1194
(June 18, 1957).

95
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

data to temperatures below 600°C would be highly
desirable. Since solid-state diffusion mechanisms
may vary over wide temperature ranges, extrapo-
lation of high-temperature data to lower temper-
atures could introduce large errors in calculated
diffusion rates, The experimental procedures
which are to be used in future investigations are
described below.

Depletion Method. — If consideration is given to
an Inconel-molten salt system in which the molten
salt initially contains dissolved CrF, and Cr*F,
and the Inconel contains no Cr°*, a random
exchange will take place, as shown by Egq. i,
although the net change of total chromium is zero.
The combined action of the exchange reaction
and the diffusional forces within the Inconel will
result in @ gain of Cr*® in the Inconel and a loss
of Cr*F, from the salt, If the depletion of Cr*F,
activity in the salt (corrected for time decay) is
measured as a function of time, a diffusion
coefficient for chromium in the metal may be
calculated by means of the following relationship:

by=o ~ bt 22
(2) — - l—e erfc(a\/t) ,
t=0
where
t = time, sec,
bt=0=counfs/g-min {at time count is
made} of a filtered salt sample
taken at ¢ = 0,

b, = counts/gemin (at time count is
made) of a filtered salt sample
taken at ¢,

A [CF] Py —

D, sec"],

 

aq =— —_
v [CrF2] Psalt

A/V = ratio of the salt-exposed area of

Inconel to the salt volume, em™7,

[Cro]/[Cer] = weight fraction ratio of chromium
in Inconel to chromous fluoride
(as Ce™™) in the salt,
pw/psah = density ratic of metal to salt,
D = diffusion coefficient, cm?/sec.

Equation 2 is based on a simultaneous solution
of the linear diffusion equation

2
aCC rO* J CCrO*

(3) =D ,
ot ax2

 

96

and the equation resulting from a balance of the
instantaneous transfer rates of labeled chromium
from the salt to the metal, or

(4) 0 M ) % DA a[C (0,1)]

The variable x is distance within the lnconel

measured in the direction of diffusion, in cm;
Cc,o. is concentration of Cr°, in g/cm®; and
Mc g is the amount of Cr*F, in the melt, in g.

r F2 2

The boundary conditions applied to obtain this
solution are: (1} the Inconel is infinitely thick
in the x direction, (2) the initial concentration of
Cr®* in the lnconel is zero, and (3) the concen-
tration of Cr®* at the Inconel surface at any ¢t > 0
is governed by Eq. 1 and varies with time
according to the relationship

[Cr*Fz]

(5) (G, oo =Pul e _[E:E;T .

The simultaneous solution holds only when a
is constant with time. This will require that the
temperature and A/V be held constant for any
experiment to which Eq. 2 is applied.

Inconel containers have been fabricated for a
series of diffusion experiments based on these
considerations. The containers consist of large
cylinders which enclose small cylinders. The
cylinders are arranged so that the salt contacts
periphery and bottom of the small
cylinder and the inner periphery and bottom of
the large cylinder. For given-diameter cylinders
and given stirrer—thermocouple well arrangements
(all components Inconel), the A/V ratio becomes
independent of fluid height when the vertical
distance between the two cylinder bottoms is
properly adjusted. Salt samples may be removed
from the container to obtain the count data without
changing the value of A/V. The corrected counts
are plotted vs the square root of time, and this
curve is compared with a plot of the function

1 — e erfe(u) vs u, where u may be considered

to be @\/t. Values of a, and then D, are easily
calculated by this method.

the outer

Constant-Potential Method for Low-Temperature

Experiments. — The results of experiments based
on the depletion method indicate that the Cr*F,
content of molten salt will remain essentially

constant if the pot is first equilibrated, with

L e g e
respect to equation 1, at high temperatures for
several days and then subjected to lower temper-
atures. The same condition would exist if the
amounts of CrF, and Cr*F, in the container
exceeded the solubility of chromous fluoride at
the temperature of interest. In either case, the
Cr*F, concentration would remain constant under
small subsequent depletions, since large amounts
of Cr*F, and/or Cr°* are involved.

The surface of a ]/4-in. Inconel tube or thermo-
couple well subsequently immersed in such a salt
would pick up labeled chromium under conditions
of a constant surface potential, that is, the Cr°*
concentration at the surface of the immersed
specimen would be constant with time. The
corresponding Cr°* transfer equation is

1/2
D¢
(6) AM o, = 24C_ o, <—->

w
A rearrangement of Eq. 6 gives

2

1 [y [CrF2] 1
D = - '
167t (z (GO rbp >

b = height of the immersed specimen,

 

where

r = radius of the immersed specimen,

y = total counts of the entire specimen per min
at ¢,

z = counts of the salt per g-min ot ¢.

The measured variable y is an accurate indication
of the Cr°* gained by the specimen because of
the low penetrations or diffusional distances
involved {x < 5 microns).

From an experimental point of view, the
constant-potential method appears to be the
simplest of the two methods, and current plans
are to concentrate activity on this method and to
utilize depletion experiments only for a few high-
temperature studies. The depletion method does
not yield measurable depletions at low temper-
atures (T < 750°C) unless long exposure times
are used.

Results. — The results of initial experiments
with the two methods are shown on Fig. 2.3.15.
The 900°C value was obtained by means of the
depletion method and the 755°C value was
obtained by using the constant-potential method.

PERIOD ENDING JUNE 30, 1958

UNCLASSIFIED
ORNL-LR-DWG 31228

 

 

 

 

 

 

 

-

 

|
"ZK‘TI

[Cr°] TYPE OF EXPERIMENT

 

    
      
 
   
  
     

    

"3 T 020 [Cro*]DISTRIBUTION IN ALLOY |
(NO SALT), REF.15
-14 e 016 [Cr"F,] DEPLETION METHOO (WITH SALT)
o 0.6 Cro*/[Cr*F,] CONSTANT-PONTENTIAL
5 - METHOD (WITHSALT) —
| 7 |
-16

 

60 65 7.0 75 80 85 90 95 100 105
10%/°K

Fig. 2.3.15. Diffusion Coefficients for Chromium in
Nickel-Base Alloys,

The solvent was molten NaF-KF-LiF (11.5-42-46.5
mole %) for both cases. High-temperature data
which appear in the literature ' for a nickel-
chromium alloy similar to Inconel are also shown
for comparison.

VAPOR PRESSURES FOR THE CsF--BeF2
SYSTEM

F. F. Blankenship

A study of the vapor pressures of the system
CsF-BeF, was made in order to obtain information
on the effect of composition on the thermodynamic
activities in fuel mixtures containing BeF,. The
deviations from ideal behavior in systems related
to the BeF,-containing fuel mixtures were found
to depend strongly on the size of the alkali cation.
Since vapor pressures for the Nc:l':-BeF2 system
17 it was of interest to
compare the effect of substituting Cs” ions (radius
1.69 B\) for Na® ions (radius 0.95 K) Another

reason for choosing the CsF-BeF, system for

S. Cantor

were measured previously,

study was that association in the vapor phase
was expected to be less pronounced, and hence
the system should be more readily amenable to
determinations of activities from vapor pressures.

 

7K. A. Sense and R. W. Stone, [. Phys. Chem. 82,
453 (1958).

97
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

The vapor pressure data were obtained by a
quasi-static method, developed by Rodebush,'®
which gives total pressures. The results, as
presented graphically in Fig. 2.3.16, show no
change with composition in the composition region
from 80 to 100 mole % BeF,. Since CsF probably
makes a negligible contribution to the total
pressure in this region, such behavior is sug-
gestive of a liguid-liquid immiscibility gap.
Attempts to confirm the existence of the immis-
cibility gap by examination of quenched samples
have, as yet, been inconclusive.

At 900°C and 30 mole % BeF,, the vapor
pressure of NaF-BeF, is 3 mm Hg compared with
7 mm Hg for ideal behavior; for CsF-BeF, the
vapor pressure is 1l mm Hg compared with
19 mm Hg for ideal behavior.
of the curves for vapor pressure as a function

The general shape

of composition in both systems probably implies
positive deviations in Ber-rich mixtures and
negative
Also, the existence of a vapor compound such
as CsBeF, probably obscures the extent of the
negative deviations in the middle of the CsF-BeF,

deviations at lower BeF2 contents.

system,

A transpiration method is being developed for
obtaining weight and composition data for the
saturated vapor. This information, combined with
the total pressure data, will yield partial pressures
of the vapor components and, hence, the activities.

 

8%, H. Rodebush and A. L. Dixon, Phys, Rev. 26,
851 (1925},

UNCLASSIFIED

ORNL -LR-0DWG 31229
200 T

 

 

 

   

 

 

 

 

 

 

oo —
00— U e N _
[ T T ’*_A‘i;’,, : 7 T Y
P ..Ln_&‘\ 1000°C ——= — i /.jk’
- = e e T
50 : e /'7 \

 

p— e, 9507C

 

 

Ny
Qo

 

VAPOR PRESSURE {mm Hg}

Q

 

 

 

 

CsF CONTENT OF CsF-Bef, MIXTURE (male Fo)

Figc 2.3-] 6-
Cs F-Ber.

Total Vaopor Pressures in the System

98

FUSED CHLORIDES AS HEAT TRANSFER
FLUIDS

R. E. Moore

Tests were run in order to determine the com-
patibility of Inconel and fused chlorides of
interest as heat ‘-ansfer fluids. In initial tests,
the eutectic mixtures 29 mole % KCI-71 mole %
ZnCl, and 23 mole % LiCl-77 mole % ZnCl, and
Inconel specimens were sealed in containers of
fused silica and held for 90 hr at 600°C. After
the test the melts were found to contain 330 and
300 ppm of chromium, respectively, and metallo-
graphic examinations of the inconel strips re-
vealed that there was no perceptible attack. After
a subsequent test carried out under similar
conditions with excess chromium metal instead
of Inconel, the KCI-ZnCl, melt was tound to
contain about 9.7 wt % CrCIZ, Another test with
an insufficient amount of chromium metal for
complete reaction produced a melt containing
6.3 wt % CrCi2 and a metallic phase that was
identified by x-ray diffraction as mainly zinc
metal containing some chromium metal probably
enclosed in pure zinc. It seems clear that the
attack on chromium by KCI-ZnCl, is so great that
corrosion of lInconel, in which the activity of
chromium is about 0.1, should be very serious.
The reaction with chromium in the experiments
with Inconel was probably far from completion
after 90 hr because of the slowness of diffusion
of chromium from within the alloy to the surface
layers.

The mixture 58.3 mole % LiCl-41.7 mole %
RbCl, which appears to be more attractive from
the standpoint of vapor pressure and corrosion,
was also tested in contact with chromium metal
in a sealed tube of fused silica. The chloride
melt was prepared by evacuation of the prepared
mixture at room temperature followed by continued
evacuation during slow heating until the sample
fused. After heating at 600°C for 72 hr in contact
with chromium the melt was still colorless, but
there was a thin, dark-green coating on the pieces
of chromium metal. The coating was probably
Cr,0;, but no x-ray pattern could be obtained.
Hydroxide ion resulting from hydrolysis during
fusion of the chlorides could account for the
formation of Cr203. The reaction

6LiOH + 2Cr°® — Cr203 + 3Li20 + 3H2

is strongly favored from the standpoint of free
energy. The melt contained 300 ppm of chromium,
that is, far in excess of the amount expected from
the reaction

Cro+ 2LiCl == 2Li°+ &Cl, .

The equilibrium constant calculated from values
of free energy of formation'® at 1000°K is 10-29,
The chromium content in the melt may represent
the solubility of Cr,O4 in LiCI-RbCl.

Procedures for purifying LiCI-RbCl to remove
hydroxides and oxides are being studied. An-
hydrous HCI should be satisfactory, but Laitinen
et al.?® report that the reaction is very slow in
LiCI-KCl. If this is due to a very low solubility
of HCl in the melt, the purification of large
batches by this method may be very difficult. One
proposed method is fusion with ZrCI4 under helium
pressure followed by distillation to remove excess
ZrCl, and filtration to remove ZrO,. The
reactions

ZrCl, + 2H,0 —> Zr0,, + 4HCl

and

ZrCl, + 2LiOH —> ZrO, + 2LiCl + 2HC

should serve to remove all water and hydroxide
ion. Experiments are being planned to determine
the rate of mass transfer of chromium in Inconel
in contact with LiCI-RbCl under a temperature
gradient.

PERMEABILITY OF GRAPHITE BY MOLTEN
FLUORIDE SALTS

G. J. Nessle J. E. Eorgan

There is evidence that molten salts of the type
now being considered for use in molten-salt
reactors will permeate graphite. Since advanced
designs now being considered include graphite

moderators, some method must be devised to

 

19A, Glassner, A Survey of the Free Energies of
Formation of the Fluorides, Chlorides, and Oxides of
the Elements to 2500°F, ANL-5107 (Aug. 1953).

20H. A, Laitinen, W. S. Fergusen, and R. A. Oster-
young, J. Electrochem. Soc. 104, 516 (1957).

PERIOD ENDING JUNE 30, 1958

prevent such impregnation or to reduce it con-
siderably, if the necessity of cladding the graphite
or protecting it in some manner from contact with
the molten-salt fuel is to be avoided.

A possible method for preventing the graphite
from absorbing the molten salt would be to
saturate the graphite with an inert salt whose
melting point is somewhat higher than the pro-
posed operating temperatures. In order to test
this method, samples of graphite were obtained
and machined into 3-in, rods 1, ]/2, and ]/4 in. in
diameter. A nickel rack was built to hold three
rods of each size in a vertical position inside
a nickel-lined flanged-topped receiver can. In
a separate reactor vessel a 3-kg batch of
LiF-MgF, (67.5-32.5 mole %) was purified by
successive treatments with hydrogen and HF in
the normal manner. The reactor vessel was then
cooled to room temperature and connected to the
receiver vessel containing the graphite rods by
means of a 3/8-in. nickel tubing transfer line. The
entire system was evacuated at room temperature
for about 1 hr,

The receiver vessel containing the graphite rods
was then heated to about 1500°F under vacuum
and maintained under these conditions for varying
lengths of time, After a predetermined time for
degassing the graphite, the reactor vessel con-
taining the salt was heated to near 1700°F under
When the salt had reached the desired
it was transferred to the receiver

vacuum.
temperature
vessel containing the graphite rods under vacuum.
Upon completion of the salt transfer, with the
graphite rods completely submerged, a pressure
of 15 psi of helium was applied to the receiver
vessel. After a specified length of time the
pressure in the receiver vessel was relieved and
the salt was transferred out. The graphite rods
were then cooled, removed from the receiver
and placed in a
desiccator. Visual observation indicated that
the rods were undamaged. The diameters of the
rods showed no changes from the original mi-
crometer measurements.

vessel, examined visually,

The rods were then weighed to determine
whether they had gained in weight. No visible
fluoride salt was adhering to the surface of any
rod. The weight gains of the rods in each of
three experiments are listed in Table 2.3.8. The
treatment of the graphite in experiment 1 consisted
of evacuation to a pressure of 200 g for 24 hr

99
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Table 2,3.8. Weight Gains of Graphite in Impregnation Tests

 

 

 

Experiment Red Rod Diameter Rod Weight (g) Net Weight Gain Average Net Weight
No. No. (in.) Before Test After Test g % Gain (%)

1 1 1/4 3.8698 4.0700 0.2002 5.17
2 ‘/4 4.0173 4.2966 0.2793  6.95 5.97
3 ‘/4 3.8709 4.0952 0.2243 5.79
4 ‘/2 15.6341 16.6759 1.0418  6.66
5 ‘/2 15.7552 16.9083 1.1531 7.32 7.12
6 ‘/2 15.6054 16.7557 1.1503  7.37
7 1 62.9228 68.0075 50847  8.08
8 1 64.1091 69.1408 5.0317  7.85 7.87
9 1 63.9181 68.8318 4.9137  7.69

2 1 ‘41 3.9140 4.2294 0.3146 8.04
2 '/4 3.8092 4.0628 0.2536 6.66 7.30
3 ‘41 3.9463 4.2303 0.2840 7.20
4 ‘/2 16.1923 17.2518 1.0595  6.54
5 ‘/2 15.7918 17.0468 1.2550  7.95 7.11
6 ‘/2 16.1700 17.2746 1.1046  6.82
7 1 63.7184 69.6872 5.9688 9.37
8 1 64.0042 69.3868 5.3826  8.41 8.80
9 1 63.8327 69.3364 5.5038  8.62

3 1 1/4 3.8981 4.1530 0.2549 6.54
2 ’/4 3.7425 4.0287 0.2862 7.65 7.1
3 ‘/2 15.8908 17.0235 1.1327 7.13
4 ‘/2 15.8184 17.0048 1.1864  7.50 8.23
> % 15.0504 172852 1.2348  8.20
6 1 63.6333 69.3552 5.7219 8.99
7 1 64.2699 69.2838 5.0139  7.80 8.51
8 1 62.6176 68.0962 5.4786  8.74

 

100
followed by exposure to the salt mixture for 1 hr,
in experiment 2, the graphite was evacuated at
a pressure of 400 u for 54 hr and then exposed
to the salt mixture for 48 hr. In experiment 3,
the graphite was evacuated at a pressure of 100 x
for 30 hr and exposed to the salt mixture for 48 hr,

One l/z-in.-dic rod and one l-in.-dia rod from
each experiment were used to determine the depth
of salt penetration by machining several ]/32-in.
cuts from each and submitting each cutting for
analysis of the lithium and magnesium contents,
One ]/4-in.-dia rod from each experiment was
completely ground and analyzed for lithium and
magnesium  content, The analytical results
obtained from the ]/4-in.-dia rods and the cuttings
from the ]/2- and 1-in.-dia rods are listed in Table
2.3.9. Analysis of the salt bath before the test
yielded the following results: Li, 16.5%;
Mg, 19.5%; F, 64.7%. According to makeup of
the salt composition the constituents should have
been present in the following amounts: Li, 14.32%;
Mg, 18.12%; F, 67.56%. For comparison with the
analytical results obtained for the
samples, the analytical values for the salt bath
will be used. After the first experiment only six
cuttings were made on the ]/2- and l-in.-dia rods,
but twelve cuttings were made on the rods used

graphite

in experiments 2 and 3, and every other cutting
was submitted for analysis.

The data of Table 2.3.8 show that the repro-
ducibility of the fluoride salt penetration of the
graphite was fairly good. The weight gains were
not as high as might have been expected, but the
consistency of the results
penetration under these conditions, The results
presented in Table 2.3.9 show that the salt
penetrated to the center of the rods with little
ot no change in salt composition and in a uniform
Samples of each size of rod that was
in the LiF-MgF, mixture have been
mounted on a rack in a receiver vessel and are
now soaking in LiF-Bez-FZ-UF:4 (62-37-1 mole %)
at 1200°F. The length of this test will be 1000 hr
if no complications occur. At the end of the test
the rods will be examined to determine the extent
of penetration of the beryllium-containing fuel

indicates maximum

manner.
soaked

mixture, in particular, the uranium and beryllium
compounds.

PERIOD ENDING JUNE 30, 1958

PREPARATION OF PURIFIED MATERIALS
J. P. Blakely F. F. Blankenship
Preparation of CrF,

B. J. Sturm

The usual procedure for preparing CrF, is to
reduce CrF, with H,, but, because of inherent
difficulties with gas-solid reactions, it is difficult
to completely reduce the CrF, without reducing
a small amount of CrF, to Cr° Since there is a
continuing need for pure CrF, uncontaminated
with Cr°in experimental work related to corrosion
problems
method of preparation has been sought.

The use of SnF, as both a fluorinating and
reducing agent appears to be promising. Chromium
fluoride has been prepared by heating anhydrous
CrCl, and SnF, to 1100°C in a graphite container,
in accordance with the reaction

in fluoride melts, a more convenient

2CrCl, + 25nF, —> 2CrF, + SnCl, T + SnCI, T .

A preliminary trial with stoichiometric proportions
of the reactants indicated that some of the SnF,
volatilized without reacting. A second attempt
with 20% excess SnF, and a starting mixture
compacted with a hydraulic press gave an im-
proved product. No CrF, could be found by
examination with a petrographic microscope;
chemical analysis showed 2.5% Cl| and no de-
tectable Sn.

Production-Scale Operations

Alteration of the production facility for use in
the large-scale processing of beryllium-containing
fluoride salt mixtures is in the final stages. The
facility should be ready for use before July 1,
1958.

Experimental-Scale Operations
C. R. Croft

The experimental facilities processed 30 batches
of mixed salts totaling some 380 kg during the
quarter. Twenty-one of these batches contained
beryllium and nine did not. Of the batches con-
taining beryllium, 16 were prepared for use in
the molten-salt reactor program, three were used
by the chemistry section, and two were prepared

101
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT

Table 2.3.9. Penetration Results Obtained from Graphite Impregnation Tests

 

. a
Material Balance

 

 

Graphite Rod Salt Components Found
Experiment  Rod Diameter Cut (wt %) Lithiom {q) Magnesium (g)
No. No. ) No.
(in.) Li Mg Foundb Expected® Foundb Expected”
1 ] % 1.52 1.40 0.062 0.028 0.057 0.036
5 % 1 1.46 1.14
2 .48 1.26
3 1.55 1.22 0.267 0.143 0.198 0.209
4 1.51 1.16
5 1.55 1.15
6 1.95 1.10
7 1 1 1.27 1.31
2 1.26 1.17
3 1.38 1.23
4 1.34 1.24 1.020 0.728 0.836 0.921
5 1.31 1.19
6 1.43 1.21
2 1 ‘/4 1.47 1.34 0.062 0.045 0.057 0.057
4 % 1 1.04 0.91
3 1.17 1.03
5 1.38 1.05
7 1.68 0.99 0.254 0.152 0.169 0.192
9 1.59 0.95
" 1.97 0.95
8 1 1 1.27 1.15
3 1.40 1.27
5 1.56 1.06 1.034 0.771 0.805 0.975
7 1.59 1.20
9 1.59 1.14
1 1.50 1.14
3 2 % 1.1 1.23 0.045 0.041 0.050 0.051
4 A 1 1.22 0.90
3 1.41 1.55
5 1.52 1.32 0.218 0.170 0.189 0.214
7 1.55 1.34
9 1.99 1.55
7 1 1 1.03 1.18
3 1.40 1.57
5 1.45 1.47 0.901 0.718 1.005 0.909
7 1.19 1.48
9 .45 1.51
n 1.27 1.47

 

aAveroge results for the cuttings used.
desed on percentage found analytically multiplied by gross weight of rod after experiment.

“Based on percentage found analytically for original batch multiplied by net gain in weight of rod after experi-
ment.

102
for physical properties study. Seven of the non-
beryllium-containing batches were used in the
molten-salt program, one is being used in fuel-
reprocessing studies, and one is being used by
the chemistry section.

Transfer and Service Operations

F. A. Doss

Thirty-nine filling and draining operations were
carried out during the quarter. These operations

PERIOD ENDING JUNE 30, 1958

involved the transfer of about 95 kg of liquid
metals and 200 kg of salt mixtures.

Since larger and larger amounts of beryllium-
containing salts are being handled by this facility,
the area is being enclosed and cut off from general
building traffic. It will, however, be directly
connected with the newly enclosed production
facility. The floors of this facility, together
with those in Building 9928, have been sealed
with a dust-proofing epoxy-resin compound.

103
 
WWMNMNNMNONMNPMNRDMNMNDNMNN - ad el e e et =
SOV DBNFIOEON SO NGTHEON =

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Reactors~Power
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105
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106