ORNL-3122.txt

/ MASTER

g iase / ' ORNL-3122
UC-80 — Reactor Technology

MOLTEN-SALT REACTOR PROGRAM
PROGRESS REPORT
FOR PERIOD FROM

AUGUST 1, 1960, TO FEBRUARY 28, 1961

OAK RIDGE NATIONAL LABORATORY
operated by

UNION CARBIDE CORPORATION
for the

U.S. ATOMIC ENERGY COMMISSION

DISCLAIMER

This report was prepared as an account of work sponsored by an
agency of the United States Government. Neither the United States
Government nor any agency Thereof, nor any of their employees,
makes any warranty, express or implied, or assumes any legal
liability or responsibility for the accuracy, completeness, or
usefulness of any information, apparatus, product, or process
disclosed, or represents that its use would not infringe privately
owned rights. Reference herein to any specific commercial product,
process, or service by trade name, trademark, manufacturer, or
otherwise does not necessarily constitute or imply its endorsement,
recommendation, or favoring by the United States Government or any
agency thereof. The views and opinions of authors expressed herein
do not necessarily state or reflect those of the United States
Government or any agency thereof.
DISCLAIMER

Portions of this document may be illegible In
electronic image products. Images are produced
from the best available original document.
- ORNL=-3122
UC-80 — Reactor Technology
TID-4500 (16th ed.)

Contract No. W-T4LO5-eng-26

MOLTEN-SALT REACTOR PROGRAM
PROGRESS REPORT

FOR PERIOD FROM AUGUST 1, 1960, TO FEBRUARY 28, 1961

R. B. Briggs, Progrém Director

Date Issued

JUN 27. 1961

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPPORATION
for the
U.S5. ATOMIC ENERGY COMMISSION
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SUMMARY

PART I. MSRE DESIGN, COMPONENT DEVELOPMENT, AND ENGINEERTNG ANALYSIS

1. MORE Deslgn

Design efforts on the MSRE continued on all fronts. Although findings of con-
tinuing studies necessitated some revisions, both the goals of the experiment and
the basic design of the plant remained unchanged. The reactor is to provide facili-
ties for demonstration of salety, dependability, and. serviceability of molten-salt
systems and is intended to produce information on the behavior of graphite under
irradiation and in contact with molten salts.

The reactor, fuel pump, and heat exchanger will be located in the reactor cell.
A component and process-piping layout was completed with a check of thermal stresses
in the piping. Layout of auxiliary connections is in progress. Checking of thermal
stresses in the coolant piping is in progress. The drain-tank area layout is under
revision following the decision to eliminate freeze flanges from the drain.lines in
the initial construction.

Detail drawings of the reactor were completed. A set of five molybdenum bands
was included in the midplane of the graphite core to restrain bowing of the -string-
ers caused by the fast-flux gradient. Iour "poison tubes" are intended to introduce
nuclear poison in the reactor to balance reactivity changes stemming from xenon
sorption or potential fuel abscorption in the moderator graphite. The fuel fraction
in the core was raised to 22.5%, minimizing the nuclear consequences of such even-
tualities. The design of the air-cooled radiator and drain tanks was completed.

The off-gas system piping is designed to accommodate either fresh or recycle
helium. The xenon decay was increased to 72 days minimum so that the radioactivity
of the stored helium should not exceed 15 curiles.

Salt-containing lines and components will be provided with heaters capable of
heating the system to 1250°F. Heating units were designed for the reactor, drain
tanks, and piping. ’

Specifications were prepared for the single-arm remote-maintenance manipulator.
Design of individual tools and racks is in progress.

Heat-balance and neutron-lcvel measurementc occrve as the indicators of reactor
power. Autumatic control of reactor power appears necessary because of the slug-
gishness of the system and uncertainties of graphite. The poison tubes provide a
total of 12% Ak/k for control purposes. Interlocking functions will control the
heat withdrawal at the radiator.

Building and site improvements will be accomplished in four steps; scheduled
for completion by October 1961. . Engineering is progrescing comsistent with thilsg
goal. Shielding requiremento were eslablished, and specification and design of
services and utilities are in progress.

aes
Construction work on the reactor will be divided between lump-sum and cost-
plus-fixed-fee contractors to accomplish the work most economically. The installa-
tion of the reactor is to be completed by July 1963.

2. Component Development

Design of an improved freeze flange for 6-in. molten-salt piping was completed,
and such flanges are being fabricated for testing. Freeze flanges of the original
design for 3- 1/2-in. and b-in. pipe passed a test consisting of 104 thermal cycles
without mechanical failure, although excessive gas leakage occurred.

Freeze valves opened by direct resistance heating, by induction heating, and
by Calrod heaters were tested successfully; the latter were adopted for the MSRE.

Prototype heaters for the MSRE piping and the core vessel were fabricated for
‘testing.

A sampler-enricher concept devised for the MSRE operates within a two-chambered
dry box located above and to one side of the reactor pit. A mockup of its drive
mechanism, latch, sample capsule, connecting tube, latch stop, and capsule guide was
built and tested. A solder freeze valve to isclate the sampler during maintenance
was developed.

Flow tests of the MSRE core were conducted in a one-fifth-scale model. Proper
Tlow distribution in the entrance volute, the vessel cooling annulus, and the bottom
plenum chamber was verified. :

A test unit was built to evaluate titanium, uranium, and other materials for
use as high-temperature oxygen getters in helium streams.

A maintenance philosophy including both remote and semidirect operations was
evolved for the MSRE. The remote-maintenance demonstration facility was operated
successfully with salt following extensive maintenance, thereby ‘completing the demon-
stration. The facility is ‘being converted to the study of sample MSRE problems, the
first of which is replacement of the fuel pump. A one-twelfth-scale model of the
MSRE is being constructed; the reactor cell was completed.

A program to develop a remotely brazed Jjoint was started. Several mechanical
disconnects for auxiliary lines were procured and tested. '

Operation of long-term forced-circulation corrosion test loops was concluded.
Two natural-convection salt loops were placed in service to investigate the compati-
bility of graphite, molybdenum, and INOR-8 in MSRE fuel.

A facility for water-testing the MSRE fuel pump was completed, and various
dynamlc and hydraulic characteristics were determined.

- The design of the hot-test prototype of the MSRE fuel pump was completed, and
. fabrication was started. Procurement was initiated for cast INOR- 8 hydraulic parts
and for the drive motors and the variable-frequency supply. An experiment was pre-
pared to determine the amount of gas back -diffusing from the pump bowl into the
lubrlcatlon system. :

A survey was made of thermal stresses in the MSRE primary-pump tank, and cor-
rective design action was taken in those areas where the stresses are excessive.
v

The design of the hot-test stand for the MSRE fuel pump was completed, and
fabrication was started. Design of the water test stand for the MSRE coolant pump
was essentially completed.

The pump containifig a salt-lubricated journal bearing continues to operate
satisfactorily after 7440 hr at 1225°F, including 72 stops and starts.

The small frozen-lead-sealed pump failed after 22,000 hr of operation, appar-
ently as a result of erosion of the shaft by metal oxides.

Design and fabrication of the MSRE engineering test loop were completed.

Several types of cnd seals are being developed for use as cold end seals for
mineral-insulated sheathed thermocouples in MSRE installations.

Kemote electrical, therfiocouple, and instrument disconnects have been developed
for use in the MSRE. Prototypes of the electrical and thermocouple disconnects were
fabricated and were tested in a remote-maintenance facility.

A float-type, electric transmitting continuous level indicator is being devel-
oped for measurement of salt levels in the MSRE pump bowls.

The feasibility of.using inductance probes for single-point measurement of
molten-salt levels is being investigated.

3. Reactor Engineering Analjsis

A number of survey-type calculations were performed to obtain information con-
_cerning various alternative designs for the MSRE core. Dividing the core into two
or three concentric cylindrical regions having different fuel volume fractions did
not lead to sufficient decrease in circulating-system inventory or sufficient im-
provement in the power-density shape to justily the additional complications arising
in fabrication of a multiregion core.

The reactivity worth of the proposed control poison tubes was estimated using
a two-group, two-dimensional model with group constants derived from multigroup,
one-dimensional calculations. The results ihdicate that the proposed four tubes
have a total worth of 11% Ak/k and that the tubes may be cut off 1 ft below the top
of the core without serious loss of reactivity worth.

Preliminary nuclear calculations were completed for an MSRE core having the
fuel separated from the graphite by INOR-8 tubes. For a core with 8 vol % fuel
salt, insertion of 30-mil-thick 2-in.-diam INOR-8 tubes increased the critical
circulating-system inventory by about TC%; the percentage of fissions caused by
thermal neutrons decreased from 92% to about 88%; the average number of fission
neutrons released per neutron absorbed in fuel was reduced from 2.02 to 2.00; and
the temperature coefficient of reactivity due Lo expansion of the core salt was re-
duced by about 1%.

Eetimates were made of the dose rates at various places outside the reactor
cell. Above the reactor-cell top plug, having a thickness of 7 ft of ordinary con-
crete, the dose rate during 10-Mw operation was estimated to be 90 mr/hr. Above
the drain-tank-cell top plug, having a thickness of 6 ft of ordinary concrete, the
dose rate immediately after draining the reactor core was estimated to be 62 mr/hr;
2 hr after draining the core this dose rate was 0..47 mr/hr, and the integrated dose
over a long period of time amounted to less than 25 mr.

R
'PART II. MATERTALS STUDIES .

L. Metallurgy

Examinations were completed on two INOR-8 forced-convection loops which oper-
ated with fluoride mixtures for 8,100 and 14,500 hr. Attack in the form of surface
roughening and pitting to a maximum depth of 1-1/2 mils was found in the former; the
latter showed no attack, except for the formation of a thin surface layer. Examina-
tions of two Inconel forced-convection loops, which operated with fluoride mixtures
for periods of 13,150 and 15,000 hr, revealed attack in the form of intergranular
void formation to maximum depths of 13 and 24 mils, respectively.

Two thermal-convection loop tests were started to determine the compatibility -
of the system salt-graphite-molybdenum-INOR-8, with the molybdenum in intimate con-
tact with both the graphite and the INOR-8. The purpose of these tests is to con-
firm the feasibility of utilizing molybdenum bands to prevent lateral distortion of
the graphite bundle in the MSRE core.

During routine welder-qualification tests on INOR-8, weld cracking was observed
in some heats of material. A study is therefore being made to determine the reasons
for the cracking and to develop methods for alleviating the condition. The addition
of 2 wt % Nb to INOR-8 appears to completely overcome weld-metal cracking on the
materials investigated. The Rensselaer Polytechnic Institute hot-ductility test is
being used to determine the base-metal cracking susceptibility of several heats of
INOR-8. Heats which have been shown to be subject to heat-affected-zone cracking
also have exhibited poor hot ductility in this test. '

The development of techniques for producing solidified-metal seals for molten-
salt reactor applications is continuing. An INOR-8 sump-type seal with gold-copper
alloy operated successfully for ten closures. A metallographic evaluation of a
Reactor Division sump-type seal was also conducted. A seal utilizing an impregnated--
metal-fiber compact was also successfully operated, and the experiment is described.

Investigations of Ni-Mo-Cr-Fe alloys had shown that an intermetallic compound
formed at compositions above the upper limits specified for INOR-8. The investiga-
-tion was continued to determine whether or not the intermetallic phase would form
in alloys within the upper compositional limits of INOR-8. Microstructural examina-
tions were made of a number of alloys with compositions within the range of interest
which had been heat treated for various times and temperatures. The results of this
investigation to date strongly indicate that the intermetallic phase will not pre--
cipitate from alloys within the compositional limits of INCR-8.

The allowable design stresses for INOR-8 were.altered as a result of the ac-
quirement of more complete properties data and the application of more critical
criteria. These criteria include a consideration of radiation damage effects.

Tensile data were obtained for cast INOR-8 and creep and stress-rupture testing
is projected.

A study of the effect of notches on INOR-8 stress-rupture characteristics was
started. Initial results indicate that a notch with a 0.005-in. radius does not
significantly weaken the material.

A radiographic technique was used to determine the distribution of LiF-BeFo-

- ThFL-UFY (67-18.5-14-0.5 mole %) salt in specimens of nine grades of low-permeability
graphite following their exposure to the salt for 100 hr at 150 psig at 1300°F. It
indicated that graphite can be made that will take up salt only in shallow (<50-mil)
surface 1mpregnat10n unider ‘the above conditions.
vii

The quantity of salt that can be forced into graphite may be increased %y
prior heating of the graphite in air, which probably enlarges the pore spaces of
the graphite.

When in contact with oxygen-contaminated graphite at l300°F, LiF-BeF,-ThFy-UF),
(65-30.5-4-0.5 mole %) precipitated uranium as UOp. Under similar conditions,
LiF-BeFs-ThF)-UF), (67-18.5-14-0.5 mole %) did not develop & precipitate that could
be detected radiographically, while LiF-BeF,-ZrF)-ThF)-UF) (70-23-5-1-1 mole %)
developed a precipitate of monoclinic ZrO,. No other oxides were identified in the
precipitate by powder x-ray diffraction analysis.

Oxygen was not removed effectively from graphite by 20- and 100-hr purges with
molten LiF-BeF,-UF), (62-37-1 mole %) or LiF-BeFp-ThF),-UF), (67-18.5-14-0.5 mole %).
Similar purges with NaF-ZrF)-UF) (50-46-U4 mole %) were more effective, although
this salt mixture does not remave all the uvaygen. .

Twenty-hour treatments of graphite by the thermal-decomposition products of
NI, F-HF at 1300°F (704°C) appeéar to effectively remove oxygen from graphite, so that
it can contain LiF-BeFp-UF) (62-37-1 mole %) for periods greater than 2000 hr at
1300°F (704°C) without a dectectable precipitate forming.

In 100-hr exposures of INOR-8 to the thermal-decomposition products aof NH), ¥« HF
crystals at 1300°F, a 0.0005-in.-thick reaction layer was produced on the INOR-8.

5. In-Pile Tests

The first two MSRE graphite-fuel capsule experiments (ORNL-MTR-47-1 and -2),
each containing four capsules, were examined in hot cells at the Battelle Memorial
JInstitute. In these capsules, delicate bhellows were used for encapsulating graphite
immersed in fused-salt fuel. Of the eight capsules examined, only one was found
intact; the bellows walls had ruptured in the other seven. It is bhelieved that ex-
pansion of the molten-salt fuel on melting (following a reactor scram) caused the
‘ruptures. fxamination of the graphite specimen from the intact capsule is discussed
below. '

A new capsule design has been made for the third 'experiment, based on somewhat
different criteria. One objective of this experiment, which also contains four cap-
sules, is to study samples of graphite that have been irradiated under conditions
more extreme than those expected in the MSRE reactor. For Lhis purpose two of the
four graphite samples will be impregnated with a modified MSRE fuel. A second ob-
jective is to detect changes in the wetting characteristics ot the fuel in contact
with graphite under high-temperature irradiation. For this purpose two capsules
will contain unimpregnated graphite of a type similar to that of the MSRE. A third
objective involves the inclusion of specimens of molybdenum, INOR-8, and pyrolytic
graphite in contact with both the graphite and the fuel to demonstrate the compati-
bility of the principal components of the MSRE system. The four capsules consist of
heavy-walled INOR-8 cylinders, each containing a l—l/h-in.—diam 3-in.-long graphite
cylinder. The upper half of each graphite cylinder is machined in the form of a
boat-like cavity to contain 5 cc of molten-salt fuel, 1n which the pyrolytic graph-
ite and the metal specimens will be cuspended. The remainder of the experiment
assembly is similar to that of the previous experiments.,

Results of the postirradiation examination of the SKA graphite specimen from
ORNL experiment 47-2 at Battelle Memorial Institute are given. Examination of the
- external surfaces of the graphite, up to 34X magnification, revealed that the sur-
faces generally appeared to be unaffected by the irradigtion. A study of & cross-
sectional face of the cylindrical graphite specimen by autoradiography and core-
drilling indicated that the major portion of the fission-product activity was on
viii

or very near to the exposed surfaces except for a plane of porosity that extended.
longitudinally through the cylinder. Intrusion of fuel salt occurred through this
.fault. The remainder of tfie interior graphite was relatively free of radiocactivity
with the exception of Cst Data derived by Battelle are not sufficient to
distinguish between mechanlsms which could explain the presence of cesium in these
internal areas, and additional radlochemlcal studies are being conducted on the
specimen at ORNL.

6. Chemistry

The MSRE fuel composition has been revised to include’ 5 mole % ZrFu and is now
LiF-BeFy-ZrF) -ThF),-UF), {70-23-5-1-1 mole %); the melting point is 442°C. The pur-
pose of the ZrF) is to serve as an oxide scavenger and to provide protection against
precipitation of UO,. When the Zer/UFu mole ratio in the liquid is high enough,
excess oxide under equilibrium conditions causes precipitation of ZrO, rather than
UOo. The protection due to ZrF) remains effective until the Zth/UFu mole ratio
fails to about 2. Also, at higher ZrF), concentrations there may be a benef1c1al
effect due to complexing of oxide ions by species such as Zrote. -

The density of the fuel, given by the eguation, p(g/ce) = 2 8L - 0. OOO56t( c),
is about 2. 5 at reactor temperatures; at these temperatures the vapor pressure is
negligible.

Slow freezing of the.fuel under conditions which favor segregation, glfies an
initial solidification of phases such as 6LiF- BeF2 Zth, thereby depletlng the Zth
concentration in the remaining unfrozen liquid. ,

Confusion was resolved in regard to the melting point of the MSRE coolant
(LiF-BeF,, 66-34k mole %); its melting point is L450°C.

The .solubility of BF3 in MSRE-type fuels is about 1000 times thal of noble
gases and is in a suitable range for nuclear control schemes involving a volatile
poison. Tests of the retention of BF3 on graphite gave a favorably low answer of
about 10 ppm.

Xenon poisoning due to diffusion of xenon into the.unclad graphite considered
for use in the MSRE is more serious than was previously recognized and is difficult
to remedy by merely decreasing the permeability of the graphite. The concentration
gradient controlling the diffusion of xenon in the graphite tends to steepen sharply
with lower gas permeabilities, giving an increased driving force which tends to
counteract the benefits of an increased resistance.

The capillary behavior of MSRE fuel in graphite corresponds satisfactorily to
the nonwetting performance expected from estimated surface tensions and measured
densities, at least in clean systems. Exposure of the salt and graphite to air
leads to the formation of oxide films at the interface and to superficial manlfesta—
tions of good wettlng

Proposed procedures for chemical analysis of the operating MSRE fuel are being'

actively developed. The problem is primarily one of modification of existing pro-
cedures to adapt them for use in the High-Radiation-Level Analytical Facility.

T. Engineering Research

The enthalpy of the MSRE fuel mixture (LiF- BeFo-ZrF), - ThFh UFL, 70-23-5-1-~1
mole %) was determined over the temperature range 100 to 800°C The mean value for
ix

the heat capacity of the liquid (between 550 and 800°C) was calculated to be 0.451
cal/g-°C; this agrees well with the value 0..458 cal/g-°C estimated from an empirical
relationship developed from data for similar salt systems. A preliminary value of
0.53 cal/g-°C was obtained for the heat capacity of the coolant mixture LiF—BeF2
(68-32 mole %) between LOO and 650°C.

Initial measurements of the viscosity of the fuel mixturé LiF-BeFo-ZrF)-ThF)-UF)
(70-23-5-1-1 mole %) indicate that, although the over-all results scatter widely,
kinematic viscosities based on the lower limits of three sets of data are in reason-
able agreement. Thus, at 650°C the data range from 2.05 to 3.15 centistokes and at
800°C, from 1.2 to 1.8 centistokes.

The study of heat transfer for the mixture LiF-BeF,-UF)-ThF}, (67-18.5-0.5-1L
mole %) flowing in heated Inconel and INOR-8 tubes was discontinued after 7050 hr of
circulation. A preliminary analvsis of the date shuwed no significant trend as a
funetion of The exposure time. The data lie on the average 25% below the general
heat-transfer correlation for other normal fluids in the range Nge = 10,000 to 20,000.

8. Tuel Processing

Rare earths, but not ThF), have a high solubility in solutions of SbFs in anhy-
drous hydrogen fluoride, which may have application to the problem of recovering
decontaminated ThF) from MSER blanket sall. The LiF compound formed with this rea-
gent, LiSbF,, was a monoclinic crystalline compound. It was found to be stable to
temperatures of 600°C or more, compared to a boiling point of ~150°C for SbFé.

MSRE fuel carrier salt (LiF-BeFs, 63-37 mole %) dissolved in boiling 90% HF -
10% water at an average rate of about 50 mils/hr for the first 10 min, but the rate
dropped to <10 mils/hr subsequently. The decrease in rate was probably due to a
layer of soft material on the surface of the salt, which dissolved more slowly than
the LiF.

A laboratory-model steady-state fluorinator was operated in four runs which
demonstrated that uranium could be recovered from a salt continuously. The primary
problem was corrosion; probably because the design of the apparatus required opera-
tion at high temperatures. With a 4O-min salt residence time and excess fluorine,
”98% of the uranium was removed in a one-stage fluorinator at about 650°C.
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CONTENTS

SUMMARY . ieietevioneonoensosssessasvsssseessssassssenensssoossnsennsosessnssss 1dl
PART I. MSRE DESIGN, COMPONENT DEVELOPMENT, AND ENGINEERING ANALYSIS

1. MSRE DESIGN ......... e trasvavasnertertaneens 1
Introduction ...... cesesens cesseveasen teses et s e et ettt an s enss 1
Reactor Corc and Vcabel o Cacveseretevseasann 2
Fuel Heat Exchanger ...c.iiieitentersenesressncssssocssssnanssas L
>
9

-

Radiator ....cceveeiienieneennennne seeseas Geeeesuescesaseneanne
Drain Tanks ..veccecieersnonncensntnsecenersstosecnans cereseaeas
Eguipment Arrangement ......cceeeersescrscrssvcccasresosssnssssess 12
Cover-Gas System ...veeeevevnrrresaserserscscsnsrsasrsseessssssces 15
System Heaters .ieieiercrtecrsnsesossnsnsessssscansoscsssnassnss 16
Design Status of Remote-Maintenance System .......ce000v000e0ss 20
Reactor Control Design ...eeereencierensoressvrsssonsosscnscsnes 20
Design Status of Building and Site .....eevieevvrressoscscesanes 20
Reactor Procurement and Installation seeciierieecrccrenerreneeee 23
1.12.1 Division of Work and Schedules ....eevevvervncrscneees 23
1.12.2 Procurement of Components .ieevevsnsecovorevscansensnss 24

HHEHPRFEHF PP =
2 HEEN0 0= OV W D

| ol &)

2. COMPONENT DEVELOPMENT ....00veessssocessesosesocscsassssssscssssnssnas 26
2.1 Freeze-Flange Development ..eveescesesesssssscssscsscnsesesecase 26
2.1.1 MSRE 6-in. FlaN@e5 ..eeceveescecssoneacsssscccsssssseses 26
2,1.2 Freeze-Flange Thermal-Cycle Tests ..cceevveveversccanes 27
Fréeze VAlvVeD ...ciereecrecoerosocasssssvorscnossssssesovencones 27
HeBteT TeSTE +eveevsrernneeoorensraceansoncesasassoersasancnnes 28
2.3.1 Pipe Heaters ....cieeeerecccerecscsesccsossnsssensacesees 28
2.3.2 Core Healer .iivivrieievernssrenncievecsorsaresscncsssene 31
2.4 Sample-Enricher Development ....cceeeveneeersveecssonsassassee 31
2.4.1 Sampler-Enricher Concept ....eeeeeesvereveroesesesneass 31

2.k.2 Sampler-Enricher MOCKUD ...coveverennrsnnenncssnevasans 33

2.4.3 Solder Freeze VALVE ....eitvevvneresoannnccnennns ceees. 33

2,5 MSRE Core Development .....cvueeersrresencasnsceccssassssncnanne 37
2.6 Helium Purification .....cc.cu.... S ereresreeieasteanns et eeaeen Lo
2,7 MSRE Maintenance Development ........ccveevirnecescnccs R Ty
2.7.1 Maintenance PhiloEoPhY .eeovseveearrnsvansvsrsnsconrese LO
2.7.2 Remote-Maintenance Development Facility .........c..... 4O
2.7.3 Semidirect and Remote Tools ....ceeveececenees cecssesaas Lo
2.7T.4 MSRE MOAEL tvvinvnverrenssassossnnssonoansnesnssansnass LD
Brazed-Joint Development ..ceeeeeccerersrssossoncrsccsnsssnsesese L2
Mechanical-Joint Development ....veevsvscssssresosnssvsssscanes U2
Forced-Circulation Corrosion LOOPS ceeeveocrocscossscnsorsesceess Ll
Graphite-Molybdenum Compatibility Test .uiveeeececeverscacseses LUl
Pump Development .....cieveereeeceereoceronsoacosossssnerasnoeses L7
12,1 MSRE Fuel PUmp ...coverecvenncanccscrsanoarrnsnanass R
12,2 MSRE Coolant PUmp ..eveveveveeorssnnsovenerrsanosssese 53
.12.3 Advanced Molten-Salt Pumps ...cceceevvescvecsrsvsncaess 53
.12.4 Frozen-Lead Pump Seal ...c.vvvevnvivsscscrsesasansaess 53
2.13 MSRE Engineering Test LoOOD ..iveevressvsrrtrosaaanaesiriioassa «es 53

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REACTOR ENGINEERING ANALYSIS .icesiienvococoonssonostococoonsns

IVIEI‘I‘ALIURGY- e 0 ¢ e e L] LA R B R SN R B A B R I A A A I ¢« s

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3.2
3.3

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xii

MSRE Instrument Development ......eeveveeen. e .

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2.14.1 Thermocouple End S€2L1S +vvveerenenoencenenseesensnonns
2.1

. 2:14.3 Pump-Bowl Level Indicator ............. ceretdereneren
2.14

Criticality Calculations ..... ceaeas teetrsesceacaccesenranaon e
Reactivity Worth of Control Tubes ......cceeeivvvvennensecoonss
Preliminary Calculations for Core Containing

INOR-8 TUDES tuvvvrsorosnonnnnaancnesen ceesstesracsreasrenans
MSRE Biological Shielding and Associated Dose Rates ......00...

PART II, MATERIALS STUDIES

Dynandc Corr051on Studles S e
1.1 Status of Test Program ....cieveeess e eresssssenns s e
.1.2 Examination of TNOR-8 LoOPS cvvvivrrnronesersnesonscnss
.1.3 Examination of Inconel LooPs .....e0:.. teestseeer s eana
.1.4 Molybdenum-Graphite Compatiblllty TeStS civvveeeroeanns
ding and Brazing Studies ......c.ivieiennnn ceesesesrenerranns
.2.1 Welding of INOR-8 .............- Ceretasreeeeieeresna .o
2,2 BSolidified-Metal.Seal Development ...... teesesensennana
.2.3 Sump-Type Seal Development ........... Ceviesacsaccriana
.2.4 Metal-Fiber Seal Development .............. Ceeeriteneas
OR-8 Development ...eoveeo.. ceee i teerecereereene b e e
%,3,1 Structural Stability of 18% Mo Alloys Ceebeneen

Mechanical Properties of INOR-8 ........... Cheteceseanas ceeenes
Impregnation of Graphite by Molten Salts ..... cesrerancnns o ees
4.5.1 Permeation Screening Tests on Large and
Small Pipe tvieevsveeesnsnas Ceresersenne cesserennnn
1.5.2 Oxidation and Permeation .........cevveeenes. Ceereei e
Precipitation of UOs from Fluoride Salts ....ecevevcovcenes .
Precipitation of Zr0O, from MSRE Fuel .........
Removal of Oxygen Contamination from Graphite ......... ceeseans
4.8.1 Purging with Molten Fluorides ............... ceereaeaas
4:8.2 Purging with Ammonium Bifluoride ........coiuveves Ceees
4.8.3 Reaction of INOR-8 with the Thermal-
Decomposition Products of NH4F-HF ........cc0veeeenes

II‘ETESTS * & 0 & 5 & % 0 S PP PSP S B SEEN A & 0 & 8 P PO b e P e PR RS Re S bR

Graphite-Fuel Capsule Experiments ........ccveveneeiinrenensan.
Examination of Graphite Specimen from Experiment
Om‘-u7-2 .I.II.O......I................l'l"'....C.'I......l.

Se Equilibrl‘mstudles ..... e 0 0 0 0 b B b .I....;.I....'II.I...
'Ihe SyStem LiF-Bng-ZI‘F4'~ThF4-UF4 ¢ s e P PRI ESIELIELEEIE PN
IIYhe SyS'temLiF-Bng-ZrFq. LI I R R N R B R R I I I I I I I B R A )

Phase Equilibria in the System NaF Bng-ThF4 cecsecoas
The System CsF-ZrFg ..iceseresitveersosessassnasonsnses
Crystal Structure of LiF- SbFS ciertesesserresaes oo s
Index to ORNL Work on Fused-Salt Phase Studies .......
Solubility of HF in Molten Fluorides .....cceevesecess
Solubility of BFs in MSRE-Type Fuels ..vevveveccrncenes
Freezing-Point Depressions in Sodium Fluoride ........

.1
.1
1
.1
.1.
1.
.1
.1
1
o1

o

L]

FJQ)EDLJO\v1¥?¢JhJP’

L,2 Instrument and Power Disconnects ........... Cerereenan ,

.4 Single-Point Level Device ..... chrsennene Ceirenaeas ces

'I'Ile SyStemLiF-BEFg e s 9t s s oe s LR I I A R N S S R A IR ) ‘
7.

8.

6.2

6.3

6.4

6.5

xiii

Oxide Behavior in Fuels ........
6.2.1 Identification of Oxides Precipitated from Fuels ......
6.2.2 Removal and Precipitation of Oxides from Fuels ........

Physical Properties .eeeeesess

6.3.1 Density of MSRE Fuel

® 0 & 0 & & & 6 0 b OSSR A S eSSt

LI I IR N B I R I B B N BN L I B B I I N LI I R ] » 0

6.3.2 Stochastic Correlation of Densities of Molten

Fluoride Fuels

e »

L N I N I BN B B DAY DAY BN DAY B L BN DN IR IR DN TN Y N I B I N L BN B B

6.3.3 Empirical Equation for Fluidity in LiF-BeFo-UF4
SVStem8 ..ceerevertovecerssesecscosrcassassasssesonsas
Graphite Compatibility .............
6.4,1 Xenon Diffusion in Graphite:
Absorption in Molten-Salt Reactors Containing
Graphite ........ )
6.4.2 Wetting of Graphite by MSRE Fused Sait ........ cersenas
b.4.3 Retention of BFs on Graphite in a Molten-
Salt ReBCtOr .vviuieerneorernveessoseescssarasseanssns

Analytical Chemistry ........

Effects of Xenon

2 8 & 8 & 0 ¥ ¢ 2 & RSP NSNS ESE SN

* o ¢ 0

LRI )

0 & 8 b 6 B & B BSOS ESEEN S eeos

6.5.1 Proposed Procedures for Analysis of the
MSRE Fuel ..........
6.5.2 Stripping Rate for CO, from Water in an
MSRE Pump-Demonstration System ........ cesessesarenen

mGImRmGRESE:ARCHI'...!-‘.'-I.

T.1

7.2

Physical-Property Measurements ..
T.1.1 Enthalpy and Heat Capacity

T.1l.2 Viscosity ..

L N BN I I B BN BN BRY BN BN NN BEY JNE O NN NN BN BN DN TR B BN BN IR I B BN N Y

o b & & & ¢ s e oo s bd e

Heat-Tra.nSfer S‘tU.d.ieS I R B N AR

® & & 0 & & & 2 e PO 0 &b e s hee e

® 0 & O ¥ PO GG E S S eeP e EEeEee

LI B B B BN I B B O N B N R N B R BN K RN N A

* 8 & 0 & & b e 0 S 8 b et s SESE S EEe

® ® 8 00 0 E 8PS PN P PSS SR s s de e

FUEL PROCEssmG LR R I A A R R R A R S R B R NI A A B A S I A B R R I A R RN A B I R R R Y R R R I A A B I ]
SbFS’HF‘ SOlU.tiOnS LI IR B BN I )

8.1
8.2
8.3

Dissolution Rates for MSRE Fuel Carrier Salt ...ceeeieescccenas

Steady-State Fluorinator

. o 800 00 ¢ & 0 s 8 b s b o e r b e s

I I A SR SR B R R NI N R R R R RN S )
PART |I. MSRE DESIGN, COMPONENT DEVELOPMENT,

AND ENGINEERING ANALYSIS

1. MSRE DESIGN

1.1 INTRODUCTTION

During the past six months, several revisions were made in the original MSRE
concept. The major concept did not change, however, and a large amount of design
work was accomplished. A very brief restatement of the original plan will be made
before taking up in some detail the features that have been modified.

The reactor consists of a graphite-moderated core encased in an INOR-8 con- S
tainer, through which is pumped a molten-salt mixture consisting of LiF, BeFo, S
ZrF),, ThFy, to which is added enough U232 in the form of UF), (approximately 0.25 ,
mole % UF)) to achieve criticality in the core. The reactor vessel, fuel circu-
lating pump, arid heat exchanger are located in the reactor containment cell.

The non-fuel-bearing coolant salt, LiF-BeF, (66-3k mole %) circulates through
the tube side of the heat exchanger and through the air-cooled radiator located out-
side the reactor cell. The coolant circulating pump is also located outside the
cell in the same general region as the radiator.

Drain tanks for the fuel salt are located in the drain-tank cell separate from
the reactor cell but having the same atmosphere and pressure requirements. One
addition was made to the drain-tank complex; a fourth tank for storage and subse-
quent batch-dispensing of spent fuel has been added. This fourth tank is located
in 4 separate but adjacent cell.

The essential changes in design have arisen from two insufficiently understood
properties of graphite: the extent of its permeation by fuel salt and the amount
of xenon holdup by the graphite.

Experiments have generally cstablished that the fuel will permeate only ~1
vol % of certain grades of graphite. This degree of permeation occurs, however,
in the absence of radiation and the possible complications imposed by fission--
fragment recoils in the graphite. At this stage there are not enough in-pile data
to warrant complete assurance that greater perméation will not occur in the reactor.

This lack of knowledge has brought about two modifications of design. One of
these is the "increase of fuel fraction within the core from 12% to 22.5%. This in-
crease minimizes the reactivity change associated with soaking of graphite by the
fuel. It has an additional beneficial effect in that it reduces the concentration
of uranium and therefore results in a lower fuel inventory. The only accompanying
adverse effect is an insignificant rise (~30°F) in hot-spot temperature in.the
core.
The second design change involves the addition of poison tubes within the
core. Four l-in.-diam inverted thimbles are now inserted in the core and provide
means for inserting or removing liquid poison in a controlled manner. The tubes
provide ample poison for overriding complete saturation of the graphite by fuel.

The absorption of xenon by the graphite has been studied theoretically and
experimentally; and with the gas permeability of the graphite expected to be used
in the MSRE, it is apparent that the xenon poison level will be 2 to L% Ak/k
The poison tubes can be used to compensate for the decay of xenon after reactor
shutdown. '

A third change in the design, also dictated by the indeterminant status of
the graphite problem, involves a better method of sampling the core graphite during
the course of the experiment. An additional penetration into the top of the re-
actor vessel has been provided. Through this penetration, with a minimum of incon-
venience, a sample of graphite can be removed from the core for study.

It is obvious from these design changes that, in addition to establishing
the safety, dependability and maintainability of the molten-salt reactor system,
a fourth objective has been added to the project. This objective is to establish
the feasibility of unclad graphite as a moderator for the molten-salt reactor.

Except for the changes referred to, the MSRE design proceeded in accordance
with the original plan. Component design, system layout, building-alteration de-
sign, and services provisions progressed satisfactorily. Detailed discussion of
. these design phases follows.

1.2 REACTOR CORE AND VESSEL

A number of significant changes to the reactor core and vessel were made
during the period covered by this report. Briefly they are: the flow rate was
reduced to 1200 gpm, the core inlet volute was changed to a flow distributor, the
fuel volume in the core was increased to 22- L/E%, a poiscn system was added, a
method of sampling graphite near the outer periphery will be added, .a rieutron
shield was incorporated-in the head, and other minor changes were made.

Detail drawings of the reactor were completed and were submitted for review.
Checking of all drawings is in progress.

The reactor receives the fuel salt from the cold end of the fuel heat ex-
changer. Once in the vessel the salt passes down in a l-in.-wide annulus in a
-helical path. The swirl is stopped at the bottom of the vessel, and the fuel
flows upward through the graphite-moderated core, through the holes of the neutron
shield, and finally leaves through a bellmouthed outlet pipe.

. i
The original core design was changed in several respects. These changes were
necessitated by:

additional nuclear information,

results of hydrodynamic data on the 1/5-scale model,
investigation into the behavior of graphite under 1rrad1at10n,
fabrication difficulties of the first design.

W

Significant design data are presented in Table 1.1.
)

Table 1.1. Reactor-Vessel Design Data

Structural material INOR-8

Core vessel
ID 58 in.
Wall thickness 9/16 in.
Design pressure 50 fisi
Design temperature 1300°F
Fuel inlet temperature 1175°F
Fuel outlet temperature 1225°F
Inlet Flow distributor
Annulue ID 56 in.
Annulus OD 58 in.
Over-all height of core tank ~8 ft
Head thickness 1l in.

6-in.-0D tubing, 0.203-in. wall

Tnlet pipe
Outlet pipe 8-in. sched-L0 pipe
Core container
ID 55-1/2 in.
Wall thickness 1/b in.
Length 68 in.
Graphite core
Diameter 55-1/L4 in.

Core blocks (rough cut)
Number of fuel channels
Fuel-channel size

Effective reactor length

2 x 2 x 70 in.

1064

1.2 x 0.400 x 63 in.
~65 in.

Fractional fucl volume 0. 400

Changes in the fuel composition resulting in an increased density and specific
heat made it possible to lower the flow rate to 1200 gpm and obtain the 50°F
desired At.

The fuel volume was increased to 22-1/2% by increasing the width of the chan-
nel to 0.4 in., maintaining the length at 1.2 in., but with semicircular ends.
The reducllon in flow ratée and increased volume lowered the velocity to approxi-
mately 0.72 fps and the Reynolds number to approximately 1200. The flow in the
annulus remgins turbulent, and maintaining the core wall at a reasonable tempera-
ture presents no problem.

The fuel tlow passages will be machined in the four faces of the graphite
stringers, but no cross-communicating channels will be provided. Five molybdenum
bands will cover the middle one-third of the matrix. It is possible that a lfull
cross-sectional break could occur in a graphite stringer. Should this happen,
the upper piece would tend to float upward approximately 3/h in., creating a fuel
pocket 2 x 2 x 3/4 in. This fuel, being relatively .stagnant, might reach a very

high temperature. To prevent this possibility, an INOR rod will be placed hori-

zontally through the upper ends of the graphite blocks, one rod in each row.

The coarse screen at the top will not be used, but a neutron shield will be
placed in the upper head. This shield will be a large INOR-8 casting of circular
shape but varying in thickness from 4-3/k in. near the outer edge to approximately
8-3/4 in. at the center. It is perforated by 1-1/8-in.-diam holes on a 2-in.
square spacing-.

The antiswirl vanes in the lower head extend radially 11 in. toward the center
. from the outer periphery. This reduces the radial pressure difference to approxi-
mately 0.3 in. of fuel salt. The pressure drop across the graphite lattice sup-
porting the stringers is approximately 4.5 in., which, being large in comparison to
the radial pressure difference, assures good flow distribution te the wvertical fuel
channels. '

A poison-tube system (shown in Fig. 1.1) was added for control during startup
and to compensate for xenon poisoning effects. Poisoning will be accomplished by
means of a molten salt composed of Li and BeF,. 1In the reactor the poison salt
will be contained in four l—in.-ODAO.O65-in.-wall tubes, each separately controlled.
The tubes enter the reactor vessel through the lower head on a U4-in. radius and ex-
tend through the graphite stringers to the upper. end where they are capped off.

The section of the tube which passes through the lower head is a 3/4—in. pipe and
joins the 1-in.-0D tube Jjust inside the vessel. OQutside the reactor vessel a

‘ l/2—in. pipe goes to a reservoir located just above the flow distributor, thus
forming a U-tube. A 1/4-in. tube enters the 1-in.-OD tube through an elbow outside
the vessel head and extends to the upper end where it terminates open-ended. Gas
pressure is required to keep the poison out of the reactor, thus providing a fail-
safe condition. :

A penetration will be provided in the top head near the outer periphery for
removing graphite samples without removing the pump. Details of the method of re-
moval have not yet been determined.

_ The fuel inlet volute was replaced by a flow distributor. The distributor is
composed of a perforated section of the vessel wall and a constant-area casing
around the outside of the vessel enclosing these holes. The holes are 3/& in. in
diameter and enter the vessel wall at a 30° angle to the tangent to the outer vessel
syrface. This arrangement provides the swirling flow desirable in the annulus. The
holes have a variable circumferential spacing to provide a more even flow distri-
bution to the annulus. The reactor with these changes is shown in Fig. 1.2.

1.3 FUEL HEAT EXCHANGER

The design of the heat exchanger was revised to compensate for changes in com-
position and properties of the fuel and coolant salts. Revised design data are
shown in Table 1.2.

Drafts of heat transfer and pressure-drop design reports were completed. A
summary of stress calculations was also drafted. Specifications were approved.
According to the specifications the fabricator will be responsible for detail de-
sign and completion of the stress analysis.
5

Table 1.2. Primary-Heat-Exchanger Design Data

Structural material INOR-8
Heat load ) 10 Mw
Shell-side fluid : _ Fuel salt
Tube-side fluid | Coolant salt
Layout ‘ | ’ 25% cut, cross-baffled shell
and U-tubes '
Baffle pitch 12 in.
Tube pitch 0.775 in.
Tube ‘
Outside diameter 0.500 in.
Wall thickness - 0.042 in.
Active shell length 6 ft
Average tube length 14 ft
Number of U-tubes 7 165
Shell diameter 16 in. -
Over-all length ~8 ft fi
Shell thickness 1/2 in.
" Tube-sheet thickness 1-1/2 in.
Design temperature 1300°F
Design pressure
Shell 50 psig
Tube _ 75 psig |
Terminal temperatures ,
Fuel salt Inlet 1225°F; outlet 1175°F
Coolant salt Inlet- 1025°F; outlet 1100°F
Exchanger geometry . Parallel-counter flow
Effective log mean temperature diffcrence . l33°F
Active heat-transfer surface area 259 ftg
ffuel-salt holdup 5.5 ft3
Pressure drop
Shell side 2L psig
Tube side 29 psig

1.4 RADIAYOR

The design of an air-cooled radiator coil, enclosure, and superstructure was
completed, and drawings were submitted for checking and review. A few minor
changes in the physical arrangement as previously reportedl were made, and Table
1.3 givee the currcnt rodiator=-coil design dala.
’
o e idmaes wc€e .. e x. el i

UNCL ASSIFIED
ORNL-LR~DWG 568714

THERMAL SHIELD-COVER

s |
o — ' wense 00020 "
e | e R

y : OPERATING
/4-in. SCHED. 40\

GAS
SUPPLY-

LOADING AND
SAMPLING RESERVOIR

ELECTRIC HEATER—¢

OFF GASw

: -
SUPPLY \

REACTOR
CORE

L 44-in. TUBE

Y4-in. TUBE-*T[

{-in. OD’
x.0.065 WALL-=~HH

THERMAL

" Ya-in. SCHED. 40

Fig. 1.1. MSRE Poison-Tube System.

Radiator-coil tube spacing (see Fig. 1.3) was changed from a 1-1/2-in. tri-

‘angular pitch to a diamond pitch arrangement. Spacing of tubes within a given row

was maintained at l-l~/2 in., but the spacing between tube rows was changed from
1.30 in. to l-.l/2 in. Alternate tube rows were staggered to give a transverse
spacing between tubes in adjacent rows of 3/14 in. The change in tube spacing de--

creased the air-side pressure drop from 11.6 in. H,0 to 9:9 in. H,0 (ref 2).

The subheaders. were redesigned to be forged from solid stock and bored to nom-
inal 2-1/2-in. sched-40 pipe dimensions. This change eliminated a difficult weld-
ing situation by providing integral forged and bored nipples on the subheaders to
which the radiator tubes will be welded. It also eliminated the need for a sepa-
rate line previously required for -draining the subheaders.
UNCLASSIFIED
ORNL-LR-DWG 52034R2

FUEL QUTLET

GRAPHITE SAMPLE BLOCK

TOP CORE SHIELD

FUEL INLET

MODERATOR STRINGER
POSITIONING GRID

:GRAPHITE-MODERATOR
STRINGER

REACTOR CORE CAN

REACTOR VESSEL® .

H

FLEL PASSAGE

—GRAPHITE-MODERATOR
STRINGER

MODERATOR STRINGER POSITIONING

MODERATOR STRINGER GRAPHITE BEAMS

SUPPORT GRID Z\I’ESSI;:L DRAIN LINE
ANTI-SWIRL VANES

Fig. 1.2. MSRE Reactor.

The main header was made 8 in. in inside diameter with a 0.5-in.-thick wall to
provide enough material for drawing nipples to which the subheaders will be welded.

The salt-side pressure drop was recalculated and was found to be 21.4 psi
(ref 3).

Design of an insulated structure for enclosing and supporting the rediator
coil was also completed (see Fig. 1.4). The radiator coil will be enclosed in the
insulated structure, which is equipped with vertically operating insulated doors.
During periocds when heat must be supplied to the coil to maintain the cocolant salt
in a liquid state, the doors will be closed to form a reasonably air-tight enclo-
sure. The radiator coll will be heated by means of panels containing resistance

heating elements. Heat transmission from the panels.to the coil will be primarily
Table 1.3.

Radiator-Coil Design Data

Structural material INOR-8
Duty 10 Mw
Temperature differentials ‘
Salt Inlet 1100°F; outlet 1025°F
Air Inlet 100°F; outlet 300°F
| Air flow . 167,000 cfm at 15 in. H,0 pressure
Salt flow 830-gpm at average temperature
Effective mean At '920°F
Over-all coefficient of heat transfer 53 Btu/hr—ft2 °F
Heat-transfer surface area 706 £t°
Tube diameter 0.750 in.
0.072 in.

Wall thickness
Tube matrix

Tube spacing

Subheaders per row

Main headers
Air-side AOp
Salt-side Ap

" Nominal 2-1/2 in.

12 tubes per row; 10 rows deep

Rows of tubes on 1—1/2 in.
in rows on 1-1/2 in.

centers.
centers.

Tabes

Center-

lines of tubes in alternate rows offset )

3/4 in.

forged and bored

sched-40 pipe size,

8-in. ID, 0.5-in. wall with drawn nipples

9.5 in. H,0
21.4 psi.

by radiation, with some convection caused by the air heated within the enclosure.
The doors are suspended from roller chains which run over sprockets to a single

counterweight, which weighs less than the combined weight of the two doors.
the doors are in the up (open) position, the counterweight is held down by three

magnets.

In event of power failure or reactor scram, the magnets release the
counterweight, and the doors fall freely, closing in about 3 sec.

When

" At other times

'‘the doors will be lowered or raised in about 1 min by an electric motor through a

magnetic clutch-brake arrangement.

Preliminary drawings were also made for modification of the existing air duct,
arrangement of the radiator-enclosure supporting structure, and design of the by-

pass .duct for control of air flow.

~ Because the two fans that provide the cooling air will run at constant speed,
the bypass duct will be used to control the amount of air passing over the radiator

tubes by short-circuiting part of the air around the radiator.

will also perform three other important functions:

The bypass duct

1. At low reactor power levels, the air leaving the radiator will be at very
During these periods, the bypass will
allow cooler air to mix with the high temperature air to keep the duct at a tem-

high temperatures as shown in Fig. 1.5.

perature below 300°F.

diator is at a.lower temperature, the bypass will be closed.

At higher reactor power levels when the air leav1ng the ra-
UNCLASSIFIED
ORNL-LR-DWG 54696A

10 ROWS OF TUBES

!
N}~
5
4_*.7
Nj~
=

R
/

(]
\J
(1
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(1
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(1
\/

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/
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(D
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I
/
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=

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2 N ) _ /{\\ R
y /,\_\J'_r;\ \J_(,\_\ij_
' \J_ /i\‘\“,/# f{\_ \J_Aj\' ‘_TEN gT
T O
AIR FLOW B C) - N CD_ B (i_’) mr
A
.H/ |
-

2in. 0D ~—2in. CLEARANCE

Flg. 1.3. MSKE Radiator-Tube Matrix.

2. The bypass duct will be used to reduce the wind force on the radiator
door in event of power failure or reactor scram. In either of these occurrences,
the radiator doors will be closed and the fans will be running down, still deliver-
ing air. The air will then be routed around the radiator through the open bypass,
reducing the air static pressure on the radiator.

3. During reactor down-periods when hcat is being supplied to the radiator
tubes in the enclosed radiator frame, the bypass duct will be open to reduce the
stack effect across the radiator.

1.5 DRAIN TANKS

Nesigns were completed for two duplicate fuel drain tanks, one flush-salt’
tank, and one coolant drain tank. Dimensional changes were made on all four tanks;
the new design data are listed in Table 1.4. The nozzle connections are subject to
change in accordance with the layout of the tanks in their cell.

Two minor additional changes were made on the fuel drain tanks. A guard was
added to protect the flexible steam tubes from incidental damage during maintenance,
and the size of the steam dome was reduced to facilitate maintenance of the tank
heaters. ' ' ‘
10

.

Table 1.4. Design Data for Fuel Drain Tanks, Coolant Drain Tank,
: ~and Flush-Salt Tank

Fuel Drain Tanks ‘ INOR-8
Height , '  81-1/2 in.
Total height with steam dome 133-3/4 1in.
Diameter 48 in.
Wall thickness |
Vessel 1/2 in.
Dished head 3/k in.
Design pressure ~ 50 psig'
Design temperature 1300°F
Volume ~67-1/2 £t3
Maximum operating temperature 1350°F
Cooling method Boiiing water in double-walled thimbles
.Cooling rate 100 kw" |
Coolant thimbles
Number Lo
Diameter , 2 in.

Steam dome

Height 13 in.
Diameter 30 in.
Wall thickness '1/2 in.
Coolant Drain Tank INOR-8
Height : 76 in. .
Diameter : 36 in.
Wall thickness ' :
~ Vessel ' 3/8 in.
Dished head 5/8 in.
Design pressure 50 psig
Design temperature 1300°F
Volume 39.5 ft3
Cooling method None ’
Flush-Salt Tank INOR-8
Height . 76 in.
Diameter : L8 in.
Wall thickness
Vessel - 1/2 in:
Dished head 3/4 in.
Design pressure ' 50 psig
Design temperature ' 1300°F
Volume | ~58-1/2 vft3

Cooling method None

DOOR CHAIN DRIVE SH
E
>|’x:

COUNTERWEIGHT SPRING |
SHOCK ABSORBER (EIGHT) —~T,

BLDG. 7503, FIRST FLOOR
(ELEV. 852 ft-0in.)

/

/
AIR INLET DUCT \/
/
/

/

COUNTERWEIGHT-

AIR BAFF%ES —

‘DO

AFT

~
o

N————————
\
> o s e e o

Fig. 1.4. MSRE Radiator Coil and Enclosure.

AIR FLOW

11

. DOOR, COUNTERWEIGHT, AND DRIVE
SHAFT SUPPORTING STEEL

MAGNETIC CLUTCH-
BRAKE ASSEMBLY

REDUCTION GEAR
DOOR DRIVE MOTOR

INLET DOOR DRIVE CHAIN :
OUTLET DOOR DRIVE CHAIN |

UNCLASSIFIED
ORNL-LR-DOWG 55841

/<NTHOUSE

- ouT
L il

TG *

LET DOOR
e N
| COUNTERWEIGHT HOLD-DOWN

/ A 4 &
7 K [l I &7 § t
.. & . - e M\ \
/ - e 4 B\
/ 4 AN } . Gur e B ! N
- S . AN ! bre, 4 % & \
A= & B B \Y
S — o S O E LV \
A =L . B * A A
4% AN A - Y /Dy X N\
Ao S D 9 AN
i , - % 75 T s et s S\ D % Y
o e VNS ? 229 B
AT = - B ovs il 2 s 1iliise. 9 b, 1) 5
- = -7 \ d \ D N\ ¢
=2 il ’ T = ¢ . pa . 4 \} N N
3 - o p ; \
% e \ 4 o \ s
L=t 3 5 A 4
- = = . J BYAV (e
A f .
- == - 8 = 7 4 & £y syl 3
=T a0 - cryarx a \l g - B =&Y -
N ” . . N\ A, o1 X
= . X . o o g

& | MAGNET (THREE)

T~AIR OUTLET DUCT

RADIATOR TUBES
12

UNCLASSIFIED
\ ’ ORNL-LR-DWG 54698A

§ L — . // — 1200

a 1000

3 : / 800

. :

. | . \ 200

0 0.2 . 0.4 0.6 . 0.8 1.0 -
FRACTION OF REACTOR DESIGN POWER

AIR MASS~-FLOW RATE THROUGH RADIATOR (Ib/hr x 10 %)

TEMPERATURE RISE THROUGH RADIATOR (°F)

Fig. 1.5. Air Mass-Flow Rate and Air Temp‘eroture Rise for MSRE Radiator.
1.6 EQUIPMENT ARRANGEMENT

The salt éystem located within the reactor-cell area and the coolant-salt
circuit outside the reactor cell, including the radiator, pump, and drain tank,
have been laid out and are shown in Figs. 1.6 and 1.7.

"Stresses in the salt lines within the reactor. containment cell were analyzed .
by using an Oracle codé. Maximum stresses formed were 7050 psi, with the piping

at 1200°F and isothermal. With the reactor running at a power of 10 Mw, the maxi-
mum stress was 3550 psi. These stresses are acceptably low.

The coolant-salt circuit, including radiator, pump, piping, and drain system,
was coded for analysis, but the analysis was not" completed. '
' The drain-tank area for the fuel system was revised. Freeze flanges, origi-
nally intended to be used on the drain lines, are cumbrous, make the layout diffi-
cult, and may never be required. The drain lines will be installed as an all-welded
system. At points where lines must be removed if replacement of pipe or components
UNCLASSIFIED
ORNL~-LR-DWG 58340

B :i:”fi;'! / 3°—in.D|A7
5 TEST-CELL
VENTILATION DUCT 4

EXISTING 24-in PIPE
INSTRUMENTATION . {3 .
AREA - 2

EXISTING 24-in. DIA
SPECTROMETER TUBE;
RESERVE FOR LUBE-

R T T 1T : —

J=———CONTAINMENT
4 VESSEL

FUEL TRANSFER CELL

- e o e e T .= -

HOT STORAGE
ARCA

FUEL TRANSFER

o FREEZE FLANGE !
d——p (TYPICAL)

STACK AND AIR-DUCT PLENUM

TCABLE ACCESS'
- OPENING —,

i
A LT o he e St

S~ —
CABLE ACCESS
. ~OPENING -

SRR

SAMPLER-ENRICHER
CELL 5
o~ FREEZE VALVE ]
S\ i
~ /l_ //,

COOLANT FILL AND [ |
SAMPLING LINE : DRAIN TANK ___gXISTING PUMP ll il
oL ) PENTHOUSE . I
- = AN "
——REACTOR i , W 7A(7 4 o~
N\ ——, SHIELDING 97 ‘ :
i . ¢

fefe— w -

PR
N

TR

FLUSH TANK | (

Q-

oL R

SRR

sy
R

N “INSULATION %

,_\_‘7 e | —— = L T — — __j
OFF-GAS

S\ HOLDUP LOOP

RN

o

s

EXISTING VENT HOUSE

YT

o

A I R A TR
RETENTION 7/
ANK
NS . s

’/210 REFERENCE LINE

EXISTING

e T et T A BN A TN S T A TR L NEW AIR-QUCT ADSORBER PIT
N
i .
i3
! i
.. ]
. - EXISTING FANS
. [} EN
J
|
CXISTING OLOWER 1MOUSE T
/
" - T. T.
* (e L
) . ‘ i 7
! ' . . ¢ ~
- Fig. 1.6. Arrangement of Equipment: Plan View. '
.
1 | *
;
T N ~ ,
' ) - N b
. , .
‘ .
!
14
Co
| : .
. f .
i
. |
|
i .
1
iy '
] -
!
, ' UNCLASSIFIED
1 ORNL-LR-DWG 58341
SHIPPING : .
CASK . ' TOP OF PENTHOUSE ) LA P
FUEL TRANSFER 2
HOT-SALT AREA !
SHIPPING = CONTAINMENT VESSEL
' CONTAINER . ] / ]
PR PO DT B S 4 % TN R 6 e i
o ;-fiifiz‘r fif@m %535"53 A 4?:«% %‘",‘é %
Sk it ‘ui‘?%‘@z COVER i FUEL PUMP, 3
i whEss 00 TR TUMR EEEER |
: OFF-GAS  [4% i
5 A\ —= "# HOLDUF, j%a TEST CELL ;
= \ —Rfi¥ _~OUTER LY ——| ,HOLD k
2 - ;; RETENTION TANK  WOOP ll *‘ig:—‘ HEAT EXCHANGER . 3;
b & 5 _ A= = © ;
AR — SOt = s i
> i AN ; ST et == =T % RADIATOR
Y 4 \ i £ ’ .',L — — s - o e * b
° %:3 : E! %' oo AT *J— p - e o ‘-'_:- = ( I : 7 ?;
R4 . | B ol ri 51 :“ = T o X —ijt 1T -—v;“‘
22| DECONTAMI- 2 : e g’f’f““% 4 - %2 ER | i v i
24 NATION AREA B FILL, DRAIN AND FLUSH - o [aPs - \
e LAl TANK PIT ol SHIELDING—{ | #15mea] | [~
}éz-'\ ‘:.‘?{ ,: {!?’é!?fl?"“‘"fi v 4 ] 3 ~airae —_
CE R . 30-in-DIA [/ YR — - I NN
‘ SLEEVE - —-IE
g'lc')gl-?;%EELAREA _— REACTOR VESSEL — — |3 ,-30-in. DIA
- ' = e
Sdl < A ta 9 "N}»» gk 32-in-DIA 2:.,5
'i-tjq .“%i. ! Q“'-‘:g.‘,‘z -WATER h— SLEEVE - '. y - T = =
' ' 5; ANNULUS — —\ — &
e e e L SR e R e e
FILL AND DRAIN s .
TANK NO. 1 b OUTER STEEL SHELL TEST-CELL
EILL AND DRAIN VENTILATION DUCT
TANK NO.2
)
SECTION A-
' H
. . . [ .
Fig. 1.7. Arrangement of Equipment: Elevation.
. - : - f :

15

becomes necessary, provision will be made for a brazed seal when the new pipe is put
in. Brazed joints are satisfactory for salt systems, particularly for small-
diameter pipes and locations which are only temporarily in contact with molten sait.
Both these conditions apply to the drain lines.

- The drain valve for the fuel system is now located directly under the reactor
vessel. When this valve is frozen, the drain line from the valve back to the drain-
tank manifold will be drained of .-salt. This is made possible by a vent line which
runs from the drain-tank side of the valve to the gas space within the pump bowl.
Drawings showing this drain system, including the drain-tank arrangement, are being
revised and are not ready for inclusion in this report.

The area immediately north of the drain-tank area has been allocated for spent-
fuel storage and transfer operations. It is assumed that fuel will be transferred
to another location for reprocessing. Both the capacity of existing processing
facilities and calculations of shielding requirements indicate that the size of in-
d1vidual shipments should be 50 liters, and thus each fuel charge will have to be
subdivided into approximately 30 batches for transfer to the processing plant. Re-
covery of the uranium from a fuel charge will require three to six months, and it
seems desirable to isolate the spent fuel from the operational storage tanks if
simultaneous power operation and spent-fuel processing are contemplated.

Lubricating-oil systems for the circulating ‘pumps have been located in the
spectrometer tunnel. Piping for the oil, gas, and other auxiliary lines to the pump
is 80% complete.

Detailed studies are being made of the awxiliary piping, power cable and dis-
connect layout, and thermocouple cable and disconnect layout in the reactor contain-
ment cell. Electrical leads for the heaters will terminate in receptacles, into
which plugs from each heater section can be inserted. These plug units can be in-
serted or removed remotely by means of the maintenance equipment provided. Design
of these plugs and receptacles was completed, and specimens were fabricated for
testing. Plugs and receptacles for thermocouple use were also fabricated and are
being tested. ‘

1.7 COVER-GAS SYSTEM

Over-all design criteria were established for the cover-gas sKstem, and pre-
liminary process and instrument flowsheets were issued for review. »2  Work was
started on the detailed design of the charcoal beds and other system components.

Tge following changes have been made since the original concepts were out-
lined:

1. The decay time of the recycle helium was increased to a minimum of 72 days
of xenon holdup. The maximum quantity of radioactivity permitted in the stored
helium was reduced to 15 curies. These changes were made in order to minimize the
hazard involved in storing radioactive gas. Under the original design a total
contained activity of 7000 curics was possible.

2. The dryer and oxygen-removal units in the fresh helium supply are heing
designed to operate at 250 psig. This change will eliminate the need for recom-
pressing the gas in this portion of the system, improving both the economy and
reliability of the operation.
16

3. System.pipihg is so arranged that the fuel-salt pump and storage tanks'may
use either recycle or fresh helium, while the remalnlng portlons of the system are
served only by the fresh helium supply. :

Y, A hlgh-flow, Jow-heat-load charcoal bed is provided to take care of venting
during salt transfer operations. '

Figures 1.8 and 1.9 present simplified flowsheets of the cover-gas system.

UNCLASSIFIED
ORNL-LR-DWG 57666

ACTIVITY |__,,
VYRR | MONITOR ;E
I B
| COOLANT SYSTEM

COOLANT-SALT VENT HEADER
PUMP ‘

. .  [acTivity] ___

[
VOLUME CHARCOAL BED
FUEL-SALT HOLDOuUP 72 doys’ Xe HOLDUP
PUMP
T
%fififififiLE — BLDG. 7503 ABSOLUTE STACK
T VENT HEADER" FILTER
PRESSURE
EQUALIZING LINE
I |
AUXlUARY-SYSTE@_J//I CHARCOAL BED:
VENT HEADER — 1y GH FLOW, LOW HEAT LOAD

Fig. 1.8. Off-Gas Disposal, MSRE quer-Gas System.

1.8 SYSTEM HEATERS

All salt-containing components and lines in the system must be equipped with
heaters that will satisfy two requirements. First, lines and components must be
preheated from room temperature to 1250°F before salt is introduced, and second,
heat losses must be counterbalanced during periods of zero power operation. Elec-
trical heaters of several types have been selected that will satisfy these require-
ments plus the remote-maintenance requirement of all components in.the primary con-
tainment cells. Power input into the system from the heaters will be regulated by
varying the voltage applied to the heating elements. Voltage regulation is to be
provided by three-phase induction regulators presently installed in Building 7503.
Metallic-sheathed, mineral-insulated cables will enter the cell and terminate in
disconnects for each heating unit. End seals on both ends will prevent gas leakage
through the cables. :

A typical line heating unit (see Fig. 2.5, Sec. 2.3.1) contains a ceramic
heating element of resistance heating coils (Nichrome V) embedded in fused alumina.
Embedding the coils in a ceramic material offers several advantages over an open.
coil design. Among these are a uniform release of heat and a more positive protec-
tion against a short to ground of the electrical conductor. The thermal insulation
shown is a light-weight refractory fiber in felted form. The principal constituents
of this fiber are alumina and silica. The heating element and the thermal insula-
tion are contained in a metallic housing. ' The housing will contain particulate
matter within the unit and will also provide a means by which tools can be attached
to the unit for removal or installation. The unit is split on the center line.
Attached to the clamshells are hinges, straps, and latches to secure the unit on
the pipe. A similar unit can be fabricated for the pipe bends.
UNCLASSIFIED
ORNL-LR-DWG 57667

-—] M Je——— oFF-GaS FrROM FUEL PUMP

OXYGEN DRYER RECYCLE SYSTEM
REMOVAL CHARCOAL BED
L
[___-'I
|
!
DIAPHRAGM RECYCLE -HELIUM : i
SIURAGE 1ANK :
|
ACTIVITY | _ | .
MONITOR — FUEL - PUMP
l——l\p—(l = FUEL-SYSTEM STORAGE TANKS

FILTER OXYGEN ——= COOLANT- PUMP
REMOVAL
[ FREEZE FLANGES,
PURGE AND BUFFER LINES
FRESH-HELIUM _ FRESH—HELIUM
STORAGE TANK SURGE TANK

Fig. 1.9. MSRE Cover-Gas System Showing Flow of Fresh Helium and Recycle Helium.

In the reactor-vessel heater, current is passed through 3/8-in.-0D, 0.035-in.-~
wall Inconel tubes. One hundred and twenty-six heater tubes are contained in sixty-
three container tubes located around the vessel as shown in Fig. 1.10. The heater
Lubes are 1nsulated from the containers by ceramic insulators. The heater tubes
can be replaced without removing the thermal-neutron cover shield. To minimize
heating in the cover-shield portion of the circuit, a transition is made from
.3/8-in. tubing to 3/8-in. rod. Fourteen heating tubes are connected in series to
form one leg of a three-phase Y electrical circuit. There are three of these cir-
cuits, with a combined capability of 30 kw. A 6-in. thickness of refractory-fiber
thermal insulation encased in Inconel sheets is attached to the cylindrical walls
and also to the top and bottom shielding.

Figure 1.11 shows the heater installation for the flush-salt tank. The cylin-
drical insulation iy sell-supporling from the cell tloor. 'l'he fiber insulation is
contained within Inconel sheets supported by the steel framework. The curved
ceramic heating elements are contained within a framework similar to a picture
frame. Each frame is a 100° segment that is removeble from above. The top cover
and the bottom contain insulation only.

The fuel fill and drain tanks and the hot-fuel storage tank will have heaters
similar to the flush-salt-tank heater. '
- \
18 '
) ) UNCLASSIFIED
i ORNL -LR-0OWG 57668
REACTOR ‘
- i .
, SHIELDING COVER ¢ ‘
| THERMAL INSULATION
SHIELOING

:

% . : 7 \9:“3- zes:

, CONTAINER TUBE

- . $
’
e
f
: : Fig. 1.10. Reactor-Vessel Heater Assembly.

. PR PR T D .
UNCLASSIFIED
ORNL—LR—-DWG 57669

HEATER ELEMENT ——

,_—
=
=]
=

HEATER FRAMING ——

\\\\\\\\\m\\\\\\\,\\\\\\\\\\\\\\\\\\\\\\\m\\\\\\\\ t

| — THERMAL INSULATION

oA

o,
b

“g——HEATER GUIDE FRAME

e
RESIDITEN,

22

prdo R A T S AL, LA, T,

£

I+ .

PO e e e
oy e emmrrmar -

e !
——————)
—_———|
—_—|

i
!

]

Fig. 1.11. Typical Tank Heater.

D

The fuel heat exchanger heater is shown in Fig. 1.12. The housing is hinged
on top to allow the unit to be opened. One half of the heater assembly is swung up
and rests on top of the other half to allow the heating elements to be replaced
from above. The ceramic heating elements are contained in frames for handling.

The drain line from the reactor will be heated by the passage of approximately
675 amp at 19 v through the pipe wall. This method of heating was selected because
the drain line is approximately 55 ft long, has limited access from above, and would
probably require 25 units if heated with individually removable heaters.

A1l piping and components of Llhe coolant-salt system that are located outside
the containment vessel will be heated and insulated in a conventional manner since
direct access for maintenance is possible. The radiator enclosure will be lined
with flat ceramic heating elements with a capability of 60 kw. These units will be
20 .

UNCLASSIFIED
ORNL-LR-DWG 57670

- AT : HINGE

e NN\ EEEREREREN

(0}
>
i
\

\
\
\
\
\

%
. LKL
- BRGNS
9 SRR § ¢%o%eYetetedele!

HEAT EXCHANGER*J '
HEATER ELEMENT THERMAL INSULATION

Fig. 1.12. Heater Assembly for Primary Heat Exchanger.

"energized at all times, because without this precaution salt could freeze very
quickly in the tubes in a loss-of-flow incident..

1.9 DESIGN STATUS OF REMOTE-MAINTENANCE SYSTEM

Studies were made of the arrangement of the operating area, the design of the
maintenance control room, the arrangement of the various concrete shielding blocks,
~‘and the coverage of the cranes. A result of this work is that certain decisions
were reached which determine the nature of the remote- malntenance tools that are
belng designed.

Specifications were completed for the singleé-arm manipulator, and others are.
being prepared for the shielding windows for the maintenance control room.

Drawings of the MSRE model were completed, and the model is under construction.

Necessary work was finished in support of those who were preparing the prelim-
inary proposal for the MSRE.

Ten drawings concerning remote-handling tools were released. An additional
nine drawings covering special torqueing tools for flange assembly and disassembly
are 95% complete.

‘ Efforts are now being concentrated on the layout of connection points for the
various remotely operated tools, the necessary racks for storage of these tools,
the arrangement of controls within the. maintenance control room, and the design of
jacks for separating the freeze flanges in the primary system.

1.10 REACTOR CONTROL DESIGN

The design of the control and safety system was altered substantially from the
original concept that the strong negative temperature coefficient in the fuel salt
21

would provide complete reactor safety and stability. These alterations stemmed
principally from hypothetical problems with the graphite,! namely: (1) fuel salt
soak-up and retention in the 7% connected void space and (2) xenon diffusion into
the graphite.

The reactivity changes expected from these mechanisms are gradual but will
necessitate shimming. Fuel penetration into the graphite creates, but to a lesser
degree, the xenon problem associated with solid-fuel reactors and also produces a
power coefficient of reactivity. The power coefficient arises from the fact that
the temperature coefficient of reactivity in the graphite is twice that of the fuel.
For these reasons the MSRE will be equipped with an external control and safety
system. '

Four poison tubes (see Fig. 1.1l) containing Li6F-BeF2 salt are being considered
as the means of providing shimming, safety, and control. This concept will be thor-
oughly tested before adoption. Three of the tubes will be used for shim-safety pur-
poses and will have an individual worth of h% Ak/k when all other tubes are fully
withdrawn. The combined worth, fully inserted, is 11% Ak/k. Withdrawal speed will
be limited to a safe value. The fourth tube will have a maximum worth of 1% Nk/x
and will be used for control purposes.

Nuclear instrument channels for measurement, control, and safety will be as
follows: (1) two fission counter channels for startup and low-level operation, (2)
two log N channels covering the upper six decades of reactor power, and (3) three
independent power-level channels used solely for safety.

Figure 1.13 is a diagram of the arrangement of ion chambers. The full range
of reactor operation, from source to above design power, will be monitored at all
times.

The primary measurement of reactor power is by a (flow x AT) measurement.
Neutron-sensitive chambers will also provide continuous power readings and will be
calibrated periodically by reference to the primary power instrumentation.

Experience gained on the ARE and subsequent analog simulator studies show that
automatic control of reactor power is a necessity in the low and medium power
region. This is because small temperature changes represent large percentage
changes in reactor power and because of the loose coupling between the radiator
and the reactor.

The automatic coulruller will utilize two inputs, heat power and the output
from a neutron-sensitive ion chamber. The heat-power input is used for accurate
power determinations and to reset periodically and sutomalically the ion chamber .
which will provide the necessary transient response. This system has been pro-
posed8’9 for the EGCR and the Pebble-Bed Reactor Experiment.

The radiator is the most sensitive component from the standpoint of system
integrity, control of reactor power, and operational continuity. Salt freezing
must not take place in the radiator. The vertically sliding doors which thermally
isolate the radiator will automatically drop and electrical heat will he supplied
if coclant=salt lemperature, measured at the radiator, approaches the freezing
point. This automatic door closure will be preceded by automatic blower shutoff
and by opening the air flow bypass around the radiator. These safety actions are
also automatic and will be produced by coolant-salt temperatures intermediate be-
tween the operating temperature and door-closure temperature.
B
[

(\v
'." .

22

UNCLASSIFIED
ORNL-LR-DWG 56873

{ - SAFETY CHAMBER 2—FISSION CHAMBERS
{ — COMPENSATED CHAMBER ' { —COMPENSATED CHAMBER
' 1—SAFETY CHAMBER

1— SAFETY CHAMBER —=( )\

WATER
ANNULUS ; -

OUTER STEEL SHELL

CONTAINMENT
VESSEL

REACTOR SHIELDING «-

Sn—-Be SOURCE TUBE

Fig. 1.13. Arrangement of {on Chambers.

1.11 DESIGN STATUS OF BUILDING AND SITE

Modification of the former ARE-ART Building 7503 to adapt it for MSRE use
involves fairly extensive changes to the building and services of various kinds
within the building. The design work on all modifications will be ‘done by ORNL.
This engineering attained a completely definitive and fairly advanced executional
status in the past six months. ‘ ' ‘

Design involves at least three different categories of work: (1) demolition
or removal of presently installed extraneous items, (2) large-scale structural
changes and additions to the building, and (3) changing and implementing present
building services. E ' ' '

Work is in progress, and some 36 items involving about 600 drawings are cur-
rently being worked on or are completed. The drawings are grouped into four pack-
ages associated with the following items:
23

Construction in reactor, radiator, and containment areas

Package A -

Package B - Drain-tank pits and maintenance control room

Package C - Charcoal pit, filter pits and stack, retention pond, etc.
Package D - Miscellaneous building modifications '

These packages will be completed in the order given above, with the last .one
(D) scheduled to be finished by October 1961. The status of design drawingec is
such as to be consistent with this completion date. Effort will be maintained at
the necessary level to assure meeting the design schedule.

All requirements for shielding were established, and the load-bearing walls,
which will also serve as biological shielding, were engineered. Loose aggregate,
cooled by flowing water, will be used between the reactor cell and the secondary
steel tank which encloses the reactor cell.

The remote-maintenanoc centiovl ruuvm was engineered and is being detailed.
This room must provide sufficient shielding to protect personnel from the high
level of radiation from components which have been used in the reactor.

Engineering of the ventilation system, which assures safe control of atmos-
phere during maintenance operations, was completed. This includes specifications
and sizing of fans, filters, ducting, etec.

All water serviées, such as cooling tower, demineralizer, and hcat exchangers,
were specified. Piping for the various systems is being laid out.

Electrical services were specified, and the actual distribution system is
partially drawn. Three separate power sources will be available for the building.
There will be two 13.8-kv service lines from separate 154-kv TVA substations. In
addition to these, there are three 300-kw diesel power units to provide emergency
power in case of power failure on both TVA lines.

There remains no area of uncertainty on the site preparation, and every phase
of building modification ic in design. All work is progressing satisfactorily and
-is essentially on schedule. '

1.12 REACTOR PROCUREMENT AND INSTALLATION
1.12.1 Division of Work and Schedules
~Work required to build the MSRE will be accomplished in three phases.

Phase I, which includes all demolition work and some minor alterations to
the building, will be accomplished by a cost-plus-fixed-fee contractor. This de-
cision is based on the fact that thie work is a small portion of the entlre job
and can be accomplished under field direction without extensive engineering draw-
ings. This phase is ready for release to a contractor.

Phase II work will be accomplished by lump-sum contractors. It includes
major modifications and additions to the reactor building: erection of additional
shielding walls, extension of the reactor primary containment vessel, modification
of drein tank and storage pits, modification of the radiator pit, and construction
of the remote-maintenance control room. Also included are the waste storage pond,
the off-gas system, the cooling tower for waste heat removal (other than reactor
heat), and miscellaneous utilities.

Phase III includes all the work to be performed by ORNL personnel. Only the
critical work associated with the installation and instrumentation of the reacter
24

complex is included in this phase. The installation of the INOR-8 fuel and coolant
systems will be made by Laboratory craftsmen experienced in working with the special
materials required.- ‘Individual components such as pumps, heat exchangers, the re-
actor vessel, and various fuel and coolant tanks will be ‘fabricated by outside ven-
dors who are experienced in fabricating reactor components of special alloys.

Table 1.5 shows the anticipated starting and completion dates for the three
phases of this work.

Table 1.5. Preliminary Construction Schedule for MSRE

Start Complete
Phase I¥ ' March 1961 June 1961
Demolltlon and minor alterations
Phase II** July 1961 July 1962
Major modifications and alterations ° .
Phase III*** July 1962 ' July 1963

Installation of reactor

* To be accomplished by CPFF contractor.
%% To be accomplished by lump-sum contractor(s).
_ ¥¥%¥ To be accomplished by ORNL personnel.

1.12.2 Procurement of Components

Specifications were written and engineering drawings were prepared for the
major MSRE components. These drawings and specifications are being checked pre-
paratory ‘to sending them to prospective fabricators for bidding.

In order to minimize the cost of qualifying welders and the cost of field in-
spection, an effort will be made to have all major components fabricated by a: single
- vendor. Bidders will be requested to. submit bids for the fabrication.of a com-:
plete package of components consisting of the heat exchanger, the radiator, the-
reactor vessel, and fuel and coolant drain and storage tanks. Vendors will also
be requested to submit bids on individual components. Under this option a vendor
will be permitted to eliminate from his proposal any component he does not wish
to fabricate.

REFERENCES

1. MSR Quar. Prog. Rep. July 31, 1960, ORNL—30lM, Table 1.3, p 12.

2. W. C. Ulr%ch, MSRE Radiator Design, ORNL CF-60-11-108 (Nov. 30, 1960), p 7-8.

/

3. Ibid., p 7.
L. A. N. Smith, MSRE Cover Gas System, MSR-60-11 (Oct. 21, 1960).
5. A. N. Smith, MSRE Cover Gas System, MSR-60-44 (Nov. 30, 1960).

6. MSR Quar. Prog. Rep. July 31, 1960, ORNL-301k4, p 18-19.

25

7. B. W. Kinyon, Effects of Graphite Shrinkage in MSRE Core, ORNL CF-60-9-10
(Sept. 2, 1960).

8. E. R. Mann, EGCR Reactor Control and Safety System, ORNL CF-60-10-106,
(Oct. 25, 1960).

9. Staff ORNL, Preliminary Design of a 10-Mw(t) Pebble-Bed Reactor Experiment,
ORNL CF-60-10-63 {Nov. 1, 19G0). '

2. COMPONENT DEVELOPMENT

2.1 FREEZE-FLANGE DEVELOPMENT
2.1.1 MSRE 6-in. Flanges

Design of a 6-in. freeze flange requiring no forced-air cooling and capable
of withstanding loads imposed by the piping and internal pressure was completed
(see Fig. 2.1). Photoelastic and mathematical analyses of the design indicate that
no significant stress concentrations exist and that the thermal stresses are not
excessive.l An Inconel flange is being fabricated in accordance with this design
for temperature distribution measurements and for a check of flange behavior under

UNCLASSIFIED
ORNL-LR-DWG 5767

{in. - :
Y
/ )
. 7e-in.R
_W@/ fi'YP)
. ] Y :

2-in. TAPER
0.100/in.

i
n.

/ Lot
1Y,-in. TAPER
0.050/in.({TYP)
Y

1in. (TYP)

DOUBLE MARMAN
CONOSEAL (BUFFERED)

A
~ |+—5/8 in.
8%-in. R / 0.050-in. GAP WIDTH o .
11%g-in.R
: N
1 Yo-in. R(TYP).
: - SLOPE 1:4
- {, (TYP)
CLAMP#CLAMP - v
CLOSED T OPEN : Y ina] + | L3gin, 4'%-in.
. T —, Y in .
FLANGE ASSEMBLY | 72 in. (TYP)
l , '
[| 6-in-0D,0.204 -in-WALL
|| TUBE SIZE
|
] ] il ] _ ] t—
SYM
SECTION A-A

Fig. 2.1. 6-in. Molten-Salt Freeze Flange.

26
27

thermal-cycling conditions. A request for quotation for furnishing two flanges
fabricated from INOR-8 to this design was issued. These flanges will have double
conical-gasket gas seals with a buffer annulus between the gaskets.

2.1.2 FYreeze-Flange Thermal-Cycle Tests

The freeze-flange thermal-cycle facility2 was operated for 104 thermal cycles
between 250 and 1350YF with a 3 1/2- and a 4-in. flange in the system. No evidence
of thermal-fatigue cracking was found at the end of the test, although there was
some flange distortion.

Gas lcakage was excessive during the cooldown portion of each thermal cycle in
which the standard clamp ring was used with either a nickel-plated copper gasket
or a gold-plated Inconel gasket. The seal was greatly improved by using & ncwly
designed resilient clamp ring with eithor type ul gasket. The lowest leakage rates
were obbtaiued DY using the resilient clamp ring with a nickel-plated copper gasket.
However, indications are that the copper gasket 1s forced deeper into the flange
" grooves with each thermal cycle and that the seal would eventually fail from this
action.

No cooling air was supplied to the 3—1/2—in. flange during one of the thermal
cycles. The frozen-salt seal formed at a larger radius than during normal cycles,
but was still sufficiently smaller than the gas-seal radius to ensure a reliable
salt seal and to prevent contamination of the gas-seal gasket surfaces.

A report was issued describing the 104-cycle operation and presenting the
results obtained.3

A seal-test facility is being constructed to permit dry thermal-cycling of
test flanges under more closely controlled conditions and with better leak
detection than in the flowing-salt facility. Figure 2.2 shows the arrangement of
a test flange in the facility. Preheating is accomplished by clamshell heatere on
the pipe stubs adjacent to the flange. The bore ol Lhe t'lange is heated by a sili-
con carbidc bar heater located at the center line. These heaters will be time-
controlled to permit completely automatic ecycling. Leak detection will be accom-
pliched by means of a mass~spectrometer type of leak detector connected to the
buffer annulus of the gas seal. By flooding either the inside or the outside of
the flange with helium, either of the gasket seals may be checked. Initially, two
test flanges will be accommodated, with space provided for two additional stations
10 be added at a later date if the need arises.

2.2 IFREEZE VALVES

Two valves were fabricated and tested successfully for use in molten-salt
drain systems. These valves are built of l-l/2-in. sched-40 INOR-8 pipe cEimped
into flats, 1/2 in. thick by 2 in. long, in which a frozen plug is formed.

Testing of the direct resistance-heated valve, shown in Fig. £.3, was termi-
nated after completion of 100 freeze-thaw c¢ycles, and the valve was removed for
examination. Frcezing was carried out at zero flow through the pipe. The average
freeze time was 10 to 15 min, with 55 c¢fm of alr directed across the valve flat.
Ten cfm was required to maintain the plug after it was formed. The size of the
plug was determined by the temperature of the salt above and below the flat and by
the flow rate of cooling air; the average length of plug was 3 in. Thaw time
varied from 2.5 min with a heat input of 3.5 kw to 6 min at 1.75 kw.
28

UNCLASSIFIED
ORNL-LR-DWG 57672

CLAMSHELL HEATER

TEST STAND

- SILICON CARBIDE HEATER

PURGE LINE

. AT

AN e "-
\ . N . ANY o A
WL TN

lgi\x:

flu

)

TEST FLANGE

I

Fig. 2.2. Test Flange Installed in the Seal-Test Facility.

Testing of the induction-heated valve (Fig. 2.4) was terminated after 60
cycles. ~ Power input to the valve was 12 kw at 450 kc. Melt time averaged 35. sec,
and freeze time averaged 10 min; 10 cfm of air was required to keep the plug
frozen

At the completion of the inductioin-heated test, the induction coil was removed
and replaced by two l-kw, 2h- -in.-long Calrods rated at 1500°F sheath temperature.
The Calrods were bent into a 6-in.-long W-shaped element and clamped onto the valve-
flat. The valve has been cycled 80 times without difficulty. TFreeze time was
again 10 to 15 min, with 10 cfm of air required to malntaln the plug. Melt time
varied from 2. 5 min at 2 kw to 6 min at 0.8 kw.

The . Calrod-heated valve was selected for use in the MSRE because of 1ts
s1mple electrlcal system. : '
f2.3 - HEATER TESTS
2.3.1 Pipe Heaters

A prototype clamshell heater . for usé on the. MSRE piping has been fabricated
and is ready for testing. Figure 2.5 is a cutaway drawing of this type of heater
29

UNCLASSIFIED
PHOTO 36193

T

Fig. 2.3. Resistance-Heated Freeze Valve.
30

' UNCLASSIFIED
PHOTO 36123

20

Fig. 2.4. Induction-Heated Freeze Valve, Which Can Also Be Heated with Calrod Elements.
31

UNCLASSIFIED
ORNL—LR—DWG 57673

24-GAGE STAINLESS STEEL
PROTECTIVE COVER

4-in-THICK THERMAL INSULATOR

CERAMIC-EMBEDDED
HEATER ELEMENT

S

o

Fig. 2.5. Clamshell Pipe Heater.

and its associated insulation. The heaters will be formed for the particular size
and shape of pipe to be heated, with the requirement that they can be replaced with
a remote manipulator.

2.3.2 Core Heater

The MSRE core-vessel heaters, required to bring the vessel to operating tem-
perature, are divided into replaceable sections each containing seven heaters.?
A full-scale mockup of one such section was built for testing. The setup includes
4-in. of Cerafelt (a trade name of Johns-Manville Corporation) insulation covered
with a 0.031-in. Inconel reflector sheet between the heaters and the insulation.

Both the functional and replacement characteristics of the heaters will be
investigated.

2.4 SAMPLER-ENRICHER DEVELOPMENT
2.4.1 Sampler-Enricher Concept

The sampler-enricher system mechanism operates within a two-chambered,
shielded dry box located on top of a shielding slab above and to one side of the
rcactor cell (Fig. 2.6). A connecting tube joins the gas space of the fuel pump
bowl to the dry box. The tube penetrates the containment vessel at a 45° angle.
An operational valve and a metal freeze valve (for use when replacing the opera-
tional valve) are used to isolate the sampler-enricher cell from the pump bowl.
The drive mechanism for raising and lowering a capsule is located inside the inner
compartment of the dry box.

An outline of the procedure required to obtain a sample is as follows: A pre-
trealed sample capsule, scaled in a gas-tight container, is lowered into the outer
compartment of the dry box. Air is purged from the cell and replaced with helium.
UNCLASSIFIED
ORNL-LR—DWG 57674

/" SAMPLE CARRIER
SAMPLER-ENRICHER CELL
\\\\\\
Pi{§42727§§56// //MANlPULATOR

CABLE-DRIVE MECHANISM

7

> o]
LATCH A J D“\M
CAPSULE_——EZZA—% /] GAS-TIGHT CAN
OPERATIONAL VALVE k%a \\\\\\\Eif — DOOR
MAINTENANCE VALVE J—A"% A
I %

LEAD SHIELDING

SHIELDING I

|

l: = STATIONARY

I SHIELDING SLAB
SHIELDING

~——— CONNECTING TUBE
4
FLANGE

EXPANSION
JOINT

FUEL PUMP

\\\\\ CONTAINMENT VESSEL

Fig. 2.6. MSRE Fuel Sampler-Enricher System“Concept.

The door between the two compartments is opened, and the capsule is removed irom
the container and connected to a latch on the drive mechanism. With the door
closed and operational and maintenance valves opened, the capsule is lowered into
the pump tank to obtain the sample. The sample is raised into the dry box, and the
operational valve is closed. Fission-product gases are purged from the inner com-
partment before the door is opened. After purging, the capsule is resealed in the
gas-tight can and the door is closed. The can is then drawn into a shielded
carrier,

The addition of enriching salt follows a procedure similar to sampling.

All operations performed inside the dry box are done with a one-armed manipu-
lator. A periscope is used for viewing.
33

2.4.2 Sampler-Enricher Mockup

A mockup of a drive mechanism, latch, sample capsule, connecting tube, latch
stop, and capsule guide was constructed and tested (see Fig. 2.7). Some difficulty
was encountered in preventing the drive cable, which lowers and raises the sample
capsule, from jumping off the storage reel. Several alternative designs are under
consideration.

Two types of latches, shown in Figs. 2.8 and 2.9, were given preliminary test-
ing. The ball-type latch is easier to operate with a one-armed manipulator. The
shaft-type latch has a more positive locking action. Further testing on a more
complete mockup will be required to fully compare the two latches.

Equipment ic becing assembled four delermining the optimum shape, material of
construction, and pretreatment of the sample capsule.

Since the latch will be operated remotely, it is highly desirable to keep
frozen salt away from the parts which must be disengaged. Therefore a latch stop
was designed to be located on top of the pump tank above the maximum operating
level of the fuel salt. To reach the bottom of the pump tank, the capsule must
extend 10 in. below the latch. This long assembly must have a flexible connection
to move through the two 45° offsets in the connecting tube. To prevent the capsule
from being washed around in Lhe pump bowl by the turbulent flow of the salt, a
guide cage was designed to fit inside the pump.6 The cage consists of five l/H-in.
INOR-8 rods with a stiffener ring to hold them in place. The estimated maximum
tempera%ure of the rods due to beta-gamma heating in the gas space is approximately
1700CF. [,

2.4.3 Solder Freeze Valve

One design criterion for the samplcr-enricher system requires that the com-
ponents that are located outside the reactor cell be replaceable. A reliable
valve must thereforc be provided to prevent the escape of fission-product gases
during such maintenance operations. A metal freeze valve is being developed for
this application. It is expected that this valve would be closed only a few times
during its life.

An apparatus was constructed to test the feasibility of a metal freeze valve
which operates by meclting and refreezing a pool of solder. The wetted parts of a
simulated 1-in. valve were fabricated from INOR-8. The other parts were fabricated
from stainless steel. Figure 2.10 shows the test assembly.

A 72% Ag - 28% Cu solder (melting point 1434CF) was believed to have the
desired wetting characteristics and thermal-expansion properties when in contact
with INOR-8. Since the valve was designed for service in a gas region, the com-
patibility of the colder with fluoride salt was not considered to be an important
factor. Diffusion of constituents between the molten solder and INOR-8 may be a
problem.

A test cycle for the valve consisted of melting the solder, opening and closing
the valve, and allowing the solder to refreeze. Oxygen was purged from the system
before each cycle by evacuating the air and pressurizing the system with argon to
about 1 psig. Even with repeated purgings, traces of oxygen remained in the system
as evidenced by an adherent oxide film on all metal parts. No flux was added to the
solder.
34

s A "
%
.
“ 2
:
¢ “ i :
%

]

Fig. 2.7. Mockup of Parts of the MSRE Sampler-Enricher System.

et

Vi

I ’]m‘nm.m.‘ i

ATt

[

UNCL ASSIFIED
PHOTO 36462

&

i

)
g )

i

UNCLASSIFIED
ORNL-LR-DWG 57675

LATCH POSITIONER

3
)

a2

BALL LATCH

SAMPLE CAPSULE

Fig. 2.8. Ball-Type Latch with Sample
Capsule for MSRE Sampler-Enricher.

35

UNCLASSIFIED
ORNL-LR—-DWG 57676

SHAFT LATCH

~— SAMPLE CAPSULE

Fig. 2.9. Shaft-Type Latch with Sample Capsule for
MSRE Sampler-Enricher.
UNCLASSIFIED
ORNL-LR-DWG 54491A

TO HELIUM SUPPLY

i

NE NN
2 7
Z 7
TO LEAK DETECTOR / ’
% 7
g e #0 /
s ——- g
—— __ 7
= — === —A\‘\“\‘% g
g ’
: 4 /
/ /
Yy -,
) | ]
P
Al 7 HEATER

‘7

/ 3

A
r S N
N AR \

THERMOCOUPLE ? 28 N / \ SOLDER
b\\\{ \\\§ (72 % Ag—28 % Cu)
D 3
INOR-8 N

Fig. 2.10. Metallic Freeze-Valve Assembly.

The valve was operated successfully for 37 of the first 41 cycles. The leak
rate through the seal was < 10-8 cc of helium per second at a pressure differential
of 30 psi across the valve. Failures occurred during cycles 1, 2, 12, and 17.
During the first two cycles, when the maximum solder temperature was 1455 and
14710F, respectively, the gas seal was not tight. By increasing the temperature to
15000F and holding it there for 15 min or longer, a gas-tight seal could be formed
reproducibly in the subsequent cycles. The other two leaks resulted from failure
to lower the valve cap deep enough into the pool of solder. This will be corrected
in the MSRE valve by a positioning device. During the 41 cycles, the solder was at
1500°F or above for 73 hr. When a gas-tight seal failed to form during the next
two cycles, the test was terminated and the valve was opened for inspection.

Figure 2.11 shows the valve cap as it appeared after testing. The ridges at
the top of the solder region are primarily metal oxide, resulting from incomplete
removal of oxygen from the system. This difficulty should not occur in the reactor.
37

UNCLASSIFIED
PHOTO 36383

Fig. 2.11. INOR-8 Solder Freeze-Valve Cap After Testing.

Chemical analysis of the solder before and after testing showed that considerable
nickel and some chromium and iron had been picked up during the test. However, even
with thc changes in cowpuslilion, no change in the melting point was detected.

Failure of the valve resulted from insufficient solder in Lhe cup. Some of
the solder was lost because of improper procedures during the initial melting, and
come was lost by oxidation.

Since the valve was conceived as a maintenance valve, the anticipated service
time would be much less than the time this valve was operated successfully.

A similar valve was fabricated of Inconel. The 72% Ag - 28% Cu solder failed
to wet the Inconel al 1500°F. Subsequent tests confirmed that INOR-8 will be wect
but Inconel will not be wet by Ag-Cu solder at 1500°F in a hydrogen atmosphere.

2.5 MSRE CORE DEVELOPMENT?

The one-fifth-scale model (Fig. 2.12) of the MSRE core was used to determine
the proper ciivumlerenllial orifieée dastribution between the conslant-flow-area
L5
Fig. 2.12. Cne-Fifth-Scale P astic Model of the MSRE Core.

U~CLASSIFIED
FP40TO 36£32

-

8t
39

volute and the core-wall cooling annulus. The principal objective of the orifice
distributor is to attain uniformity of flow to the annulus. The average velocity
distribution in the volute with the final orifice design is shown in Fig. 2.13.
Measurements were made at design flow rate and at 40% of the design flow rate.

UNCI ASSIFIED
= ORNL—-LR-DWG 57677
24 l
/ INLET |
20 fif —_— =
/ \%mw RATE '
N
g 18 ™~
b
L (! INLET
5 /
il
S /
> 12 —0" f
= /
3 P /
=3}
g - ’\‘5\ ‘\40% DESIGN FLOW RATC ]
o £
SINLET /
4 I~ ,/
0
0 45 90 135 180 225 270 315 360

ANGLE (68 AS DEFINED ABOVE) (degrees)

Fig. 2.13. Velocity of Fluid in the Inlet Volute of the MSRE Core.

Note that, at the design flow rate, the velocity suddenly increases from approxi-
mately 14 fps in the inlet pipe to approximately 23 fps in the volute head. This is
because of recirculation of fluid around the volute. Flow distribution from the
volute to the annulus is prouportional to the velocity gradient at any point around
the volute, and it can be seen that within the limits of experimental error it is
constant in the model. As the fluid leaves the volute region and moves down the
core-wall cooling annulus, the swirl component of velocity decreases; therefore the
resultant total velocity drops. The lowest total velocity is at the bottom of the
annulus and was measured to be 4.2 fps. Further, it was found to be constant within
the limits of experimental error all the way around the reactor. This velocity
gives a Reynolds number of approximately 20,000 and a resultant heat transfer co-
efficient of over 1000 Btu/hr-ftz-oF, which is more than adequate for core-wall
cooling.

Residence times were measured in the lower head of the one-fifth-scale model
by a salt-conductivity technique.lo It appears that the highest fluid temperature
in the lower head, which is next to the wall and at the vessel center line, will be
less than 20°F above the mixed-mean temperature in the lower head.

Design work is continuing on a full-scale flow model of the core; however, con-
struction has not yet started.
40

2.6 HELIUM PURIFICATION

A test unit was designed and constructed to evaluate titanium, uranium, and
other materials for use as high-temperature oxygen getters. The data will be used
as a guide in designing the oxygen removal units for the cover-gas system. The test
unit will be operated in conjunction with an oxygen analyzer currently under develop-
ment. If successful, the analyzer will read oxygen concentrations in the range of
0O to 1 ppm.

2.7 MSRE MAINTENANCE DEVELOPMENT
2.7.1 Maintenance Philosophy

Because the proposed modifications to Building 7503 will not provide complete
biological shielding when the roof plugs are removed from atop the reactor cell or
drain-tank cell, it has been decided to adopt a dual maintenance philosophy that
will permit working directly through the shielding when possible or remotely con-
trolling the manipulator from a shielded maintenance control room whenever removal
of all the cell shielding is required for an operation.

The lower roof plugs will be subdivided into units 2 ft wide by a length
determined by the width of each of the five bays. A portable, steel maintenance
shield (Fig. 2.14) that can cover the opening left by the removal of two adjacent
plugs will be provided for semidirect work.

Maintenance that requires a larger opening (e.g., the removal of the heat
exchanger, the reactor vessel, or a drain tank) will be accomplished by the removal
of all the cell shielding and the use of the manipulator-television complex. At
such time the shielded maintenance control room will probably be the only portion
of the building suited to long-term occupancy.

Designs of the cell shielding and the layout of components are being made in
accordance with the requirements of the dual philosophy. Designs for electrical
and instrument disconnects, auxiliary piping disconnects, and smaller components
such as the pipe heaters, vessel heaters, and the pump and motor are taking advan-
tage of the ability to use several "hands" in the semidirect maintenance approach.

2.7.2 Remote-Maintenance Development Facility

As previously reported,ll:12 the remote-maintenance development facility was
operated at 1200°F with molten salt and then disassembled and reassembled to demon-
strate fully remote maintenance.

Before a second salt run could be made to complete the demonstration, it was
necessary to improve the heating of the fill-and-drain line, particularly near a
pair of flanges and at a freeze valve. This involved adding and rearranging heaters,
and channeling the cooling air to the freeze valve to localize the freeze plug.
After these changes were made, the system was heated to 1200-1250°F in 73 hr, and
salt was circulated for 4-1/2 hr. The system was then drained and permitted to cool.
These operations were conducted with no difficulty.

Remote disassembly and reassembly were performed on the pump, the pump motor,
the reactor assembly, and the heat exchanger. Part of this work was accomplished
by new operators in training, and with new techniques. Although a completely remote
operation could have been performed by an experienced operator, the actual operation
included some direct vision as an aid to the trainees.
41

UNCLASSIFIED
ORNL-LR-DWG 46005R
ECCENTRIC PLUG

LIGHT HOLES

MOTOR DRIVE ROLLER

Fig. 2.14. |sometric View of Portable Maintenance Shield.

The television cameras required a considerable amount of repair, as original
parts wore out. Moving parts of future cameras should be of higher quality, if
possible.

After the facility was operated and remotely disassembled, it was dismantled
and removed from the area. Various component and assembly mockups of the MSRE have
been designed, are in various stages of construction, and will be installed in the
area.

A portable-shield mockup is being constructed for installation on the manipu-
lator bridge. Tooling suited to semidirect work will be developed and will be
tested in use with this shield.

A mockup of the primary pump with holddown-bolt and Jjack-bolt extensions and
awdliary pipc stubs has been consliucled aud will be sel 1n Lhe area.

A mockup of a pipe run with freeze flange and pipe supports has been designed
and is being constructed. It will be used primarily for the development of
equipment for remotely separating and closing the freeze flanges while maintaining
the required precise alignment, and secondarily for the development of closures
and tools for replaceable pipe heaters.

The total mockup ultimately will be completed in the MSRE building.
42

2.7.3 Semidirect and Remote Tools

The detail design of several tools for use with the portable maintenance
shield has been accomplished and construction has begun. The tools will be proved
in the remote-maintenance facility and will become a part of the MSRE tool inven-
tory. Detail design of tools for use with the manipulator is also in progress.

2.7.4 MSRE Model

A one-twelfth-scale model of Building 7503, col. 3 to col. 9, sections A, B,
and C has been designed and is being constructed. It will be a functional model to
aid in design, construction, and maintenance of the MSRE. The reactor cell con-
taining the reactor with its associated heaters and shield, the pump, the heat
exchanger, and fuel and coolant piping has been completed (see Fig. 2.15.)

2.8 BRAZED-JOINT DEVELOPMENT

It has been decided to eliminate freeze flanges from the salt system of the
drain tank cell. The salt system will be of all-welded construction. In the event
a component must be replaced, the l/Z—in. and/or l-l/Z—in. salt piping will be cut
and the replacement component joined to the system by a brazed joint. It is neces-
sary to develop mechanical equipment for remotely cutting the pipe, cleaning, align-
ing the joint, heating in the proper atmosphere, and inspecting the final product.

2.9 MECHANICAL-JOINT DEVELOPMENT

Many pipe and tube disconnects are required on MSRE auxiliary lines to permit
replacement of components. Included in the auxiliary lines are lubricating oil,
helium, air, etec. Generally, specifications require that leak rates for the mechan-
ical joints be less than 1 x 10-6 std cc of helium per second as tested by a mass
spectrometer. System pressures will vary between 20 and 200 psia, and temperatures
between 75 and 1000°F. Materials used in disconnects must be able to withstand
radiation doses of 1 x 1010 rads.

Initially, disconnects were desired which could easily be operated remotely by
a one-armed manipulator. Since the adoption of the dual-maintenance concept, it
has been decided to use helium-buffered ring-joint flanges as disconnects. These
can very easily be broken and remade through the portable shield, and can if neces-
sary be maintained with the manipulator.

Prior to the decision to use ring-joint flanges, which are not expected to
require any development, several promising disconnects were evolved..l3:ll‘L These
types were subjected to pressure-cycling in steam, temperature cycling between 200
and 1000°F in air, and make-break cycles. The most promising disconnect for use on
radioactive lines is the spring-ring joint shown in Fig. 2.16. This is essentially
a tongue-and-groove joint which seals at two concentric line contacts with a helium
buffer or leak-detector region between the two sealing surfaces. For nonradiocactive
lines the cone-seat disconnect may be used (also shown in Fig. 2.16). Here the male
cone is clamped into the female recess, compressing the cone sufficiently to form a
seal but limiting the stress below the yleld point of the metal.

In addition to the single disconnects, multiple cone-seat disconnects (up to
seven per block) were tested successfully. Another multiple disconnect which
appeared promising was a ring-joint flange with manifold parts and a number of con-
centric Flexitallic gasket seals.
43

UNCLASSIFIED
PHOTO 36531

A

':'ux Mu_; N‘""’m

Fig. 2.15. Model of MSRE Reactor Cell.
44

UNCLASSIFIED
PHOTO 36401

3/8-in. CONE-SEAT
DISCONNECT

1/2-in. CONE-SEAT
DISCONNECT

TNGHES
III}I'I?I[I?[I[?III?IAIIJ'l‘{lll?l'l?l'l?l]l‘l'll

TWO-BOLT SPRING-RING CLAMP JOINT

Fig. 2.16. Cone-Seat and Spring-Ring Types of Disconnects.
2.10 FORCED-CIRCULATION CORROSION LOOPS

The operation of long-term forced-circulation corrosion test loops, eight of
INOR-8 and two of Inconel, was concluded during this period. Table 2.1 gives a
summary of the loops operated during this period and termination dates. Where loops
had sections removed for metallurgical examination prior to final termination, this
is noted in the table. All loops operated normally during this period until termina-
tion.

The examination of loops MSRP-l2, MSRP-13, 9377-5, and 9377-6 are reported in
Secs. 4.1.2 and 4.1.3. The termination of MSRP-12 and -13 was discussed
previously.12,L1

At the conclusion of the forced-circulation corrosion test program for the
MSRE, the operating experience of the 25 pump loops (9 of Inconel and 16 of INOR-8)
is summarized as follows: The 25 loops operated for a total of 289,483 hr at test
conditions, with 20 of the 25 operating for at least one year. The 9 loops of
Inconel operated an average of one year each; the 16 loops of INOR-8 accumulated a
total of 24 years of operation.

Of the 25 loops, 18 were shut down for examination as scheduled, and 7 were
shut down prematurely because of mechanical failures. The seven breakdowns may be
classified as follows: four cases of improper operator procedure in thawing frozen-
salt lines, causing pipe breaks; one case of improper maintenance procedure, causing
a tubing failure during a maintenance operation; one case of continual malfunction
of the pump clutch, leading to excessive maintenance cost; and one case of material
failure in a machined INOR-8 fitting.

2.11 GRAPHITE-MOLYBDENUM COMPATIBILITY TEST
The graphite stringers of the MSRE core are to be restrained by molybdenum

bands as described in Sec. 1l.2. It was desirable to demonstrate experimentally
that graphite and molybdenum are compatible in a fuel-salt environment.
Table 2.1. Molten-Salt Fcrced-Circulation Corrosion Loop Jperations:
Summary as of February 28, 1961

High Wall Hours of
Loop Composition (mole %) Temp. AT Operation at Termination
No. Material LiF NaF BeF, UF) ThF, (°F) (°F) Conditions Date Comments
9377-5 Inconel 62 36.5 0.5 1 1300 200 15,038 8-22-60
-6  Inconel i ! 16 13 1300 200 13,155 9-16-60
9354-3 INOR-8 1200 100 19,942 10-10-60 Loop section removed at
4314 hr and operation
resumed
MSRP-7 INOR-8 71 16 13 1300 200 20,001 12-7-60 Loop section removed at
- 8750 hr and operation
resumed
-6 INOE-3 62 36.5 0.5 1 1300 200 20,002 12-1=-60
-10 INOR-3 53 k5.5 0.5 J 1300 200 20,015 1-11-61 Loop section removed at
8772 hr and operation
resumed
-11 INOR-8 53 46 1 1300 200 20,01k 1-23-61
~-15 INOR-8 67 18.5 0.5 14 1400 200 10,878 1-30-61 Insert removed at 8768 hr
and operation resumed
-14 INOR-8 67 8.5 0.5 1k 1300 200 10,571 1-31-51 Inserts removed at 486 and
8465 hr and operation
resumed
-16 INOR-8 67 18.5 05 1k 1500 200 7,240 2-1-61 Inserts removed at 2081

and 5254 hr and operation
resumed

Sy
46

Two test sections, incorporating graphite, molybdenum, and INOR-8, were
designed and fabricated for testing in molten-salt thermal-convection loops. The
test sections were installed in two INOR-8 thermal-convection loops fabricated of
3/8-in. sched-10 INOR-8 as shown in Fig. 2.17. The test sections consist of a
number of graphite units 6 in. long encased in a 1-1/k-in. sched-LO INOR-8 pipe
section and mounted in the hot leg of the thermal-convection loops. The first test
section placed in operation consisted of two graphite units, and the second con-
sisted of four units. The salt-to-graphite volume ratios of the two loops is
approximately 2:1 and 1:1, respectively.

UNCLASSIFIED
ORNL—LR-DWG 57678

TEST UNIT WELDED INTO THERMAL-CONVECTION LOOP

P
1%4-in. SCHED~40 INOR-8 PIPE AN
il
L il
RO025 GRAPHITE 77N 2222?

MOLYBDENUM
STRIPS

INTERNAL DETAIL SHOWING ONE GRAPHITE
UNIT AND COMPONENTS

Fig. 2.17. Graphite-Molybdenum Compatibility Test Unit.

The graphite units were machined from type ROO25 graphite, a low-permeability
grade produced by National Carbon Company. The molybdenum was cut into strips 1/16 in.
thick, 1/8 in. wide, and 6 in. long. The strips were bent into the shape shown in
the drawing to provide spring-force contact against the graphite in the slots and
against the INOR-8 spacer sleeves inserted at the ends. Six molybdenum strips were
placed in each graphite unit. The flow passage through the center of the graphite
has the same diameter as the loop piping.

The first loop test, with two graphite units, began operation in October 1960,
and the second loop, with four units, began operation in December 1960. The salt
mixture used was the designated MSRE fuel salt (LiF—BeFZ-Zth-ThFu-UFu, 70-23-5-1-1
mole %). In preparation for filling with salt, the loops were evacuated at room
temperature and slowly brought up to 1140°F. They were held at this temperature
for 16 hr and brought up to 1300°F (operating temperature) while the vacuum was
maintained. The system was held at 1300°F for another 16 hr before being filled
with a salt. The first fill was circulated for 2 hr and then replaced with the
operating salt.

The metallurgical aspects of these tests are discussed in Sec. L.1.L.
47
2.1l2 PUMP DEVELOPMENT

Design, water testing, and thermal analysis of the MSRE fuel pump received
emphasis during the report period. Contracts were signed for obtaining pump
impeller-and-volute combinations of cast INOR-8. Procurement was initiated for
pump drive motors and variable-frequency supply.

2.12.1 MSRE Fuel Pump

(a) Fuel-Pump Water Tests.--Fabrication of equipment for the fuel-pump water
testl( was completed, and tests were performed on priming, fountain (up-the-shaft)
flow, coastdown characteristics, head-flow-speed characteristics, and the effective-
ness of the xenon-removal stripper.

The test facility (see Fig. 2.18) consists of a closed loop constructed of
6-in. pipe. The loop containg a wventuri [luwmeter, a throttle valve, and an orifice
flow restrictor (not shown). Provisions are available for measuring head, flow,
speed, pump input power, and tempcrature. The hydraulic design of the test pump and
that of the reactor fuel pump are identical.

Priming teste wcre conducted in which the pump was accelerated from O to
1030 rpm in approximately 30 sec, while noting changes in pump-tank liquid level
and observing attainment of normal pump-head and flow performance. The following
operating levels were noted for the listed starting levels (the levels are with
reference to the center line of Lhe volute; minus is below the center line):

Starting Liquid Level Operating Liquid Level
(in.) (in.)
2 l.2
2% 0.6
1.0 - 0.9
0.5 - 2.1

Normal hydraulic performance was achieved by the end of pump acceleration for all
runs except the one with a 0.5-in. starting level, which required approximately an
additional minute. These data should not be used for reactor-system computations
unless differences in volumes of system trapped gas are accounted for.

The behavior of the "fountain flow" was observed over a wide range of operat-
ing conditions and found to be in the desired direction, outward from Lhe shaft
anmilus into the pump tank. In particular, the resistance lines through 45 and
55 ft of head at 1450 gpm were traversed satisfactorily over the speed range of
700 to 1300 rpm.

Coastdown tests were performed to determine the time required for the pump
motor to stop after the drive motor was disconnected from the electric supply.
Tests were performed on the same flow-resistance line for operating speeds of
1150, 1033, and 860 rpm at respective flow rates of 1630, 1450, and 1210 gpm.
Coastdown times ranged from 10.1 to 10.4 sec; elapsed times for flow reduction to
540 gpm ranged from 1.5 to 2.0 sec.

Performance data were obtained over a widc range of operating conditions. The
resulls are shown in Fig. 2.19. These data cover the range of 27 to (0O ft of head
and 750 to 1700 gpm of flow. The bypass flow into the pump tank for these data was
50 gpm measured at 1450 gpm and 1030 rpm.
48

UNCL ASSIFIED
PHOTO 36279

Fig. 2.18. Water Test Stand for MSRE Fuel Pump.
49

UNCLASSIFIED
ORNL-LR-DWG 57679

80
l | I
WATER-TEST DATA 1Z-in. LIQUID LEVEL
BY-PASS FLOW, 50 gpm AT 1450 gpm AND
20 1030 rpm, WATER TEMP, 60°F ~o
\.Q..\
[ 1
60
- -
\
T T\ °- 1030 rpm
5 , . - ' . . ’_ ‘\
T ‘0\
40 ~—r——
.‘\
860 rpm
\.
~
30
._~. -
° ~0-o__ 700
o—o rpm
20
800 900 1000 1100 1200 {300 1400 ° 1500 1600 {700

FLOW (gpm)

Fig. 2.19. Head-Flow-Speed Characteristics for MSRE Fuel Pump.

- The effectiveness of various xenon-removal stripper configurations was investi-
gatcd for throughputs up to 35 gpm. These investigations were conducted with dis-
tilled water initially supersaturated with COp. By monitoring the pH of the water,
the time to reduce the concentration of dissolved COp gas by a factor of 2 was
measured. The time required for an ideal 100% effective stripper was computed to be
1.9 min, and times of 2.2 min were measured for several stripper configurations at
35-gpm throughputs.

A stripper configuration has been placed in the water stand for testing at
throughputs from 50 to 70 gpm.

(b) Hot Test of Fuel-Pump Prototype.--The design of the hot-test prototype of
the fuel pumplS (Fig. 2.20), except for the xenon-removal stripper was completed
and the pump is approximately 50% fabricated.

Contracts were signed for the production of impeller-volute combinations of
cast. TNOR-8. Specimens of the first casting or from the first-casting melt were
.requested for weldability tests to be conducted at ORNL. .

Pump-tank heads of INOR-8 were fabricatéd by hot spinning. Dye-penetrant in-
spection revealed so many imperfections that it was necessary to redesign prototype
pump tank to make use of acceptable portions of the formed heads.

Specifications for the drive motor for the fuel pump and for the coolant pump
were written; those for the fuel-pump drive motor were revised to incorporate water
50

UNCLASSIFIED
ORNL-LR-DWG-56043

. —MOTOR
A A '
/1 A
SHAFT SEAL
LEAK DETECTOR v i ; , UPPER SEAL LEAKAGE
LUBE OIL - IN MOTOR GAS CONNECTION
LUBE OIL BREATHER
SHAFT SEAL .
GAS IN
: ~ LUBE oIL ouT N
f ' LEVEL - INDICATOR
t.—~ - LEAK DETECTOR GAS IN B
| ____GAS VENT
25 o 0
i
. / g
/
BUOYANCY . ,
LEVEL T e=- ; =——
INDICATOR— NN _ c—
L

Fig. 2.20. MSRE Fuel Pump.
51

cooling; and both specifications were submitted for procurement. Specifications for
the variable-frequency power supply for the fuel-pump drive motor were written, and
the low bidder was awarded the contract.

The conversion of an existing NaK pump test facilityl8 to circulate molten salt
was completed. The pump has been in operation continuously for 3720 hr, pumping
LiF-BeFp-ThFy -UF), (65-30-4-1 mole %) at 12259F, 1950 rpm, and 510 gpm. The upper-
and lower-seal oil-leakage rates have averaged 10 and 13.5 cc/day, respectively. An
experiment, utilizing radiocactive Kr85 as a tracer gas, has been planned for this
pump, to determine the extent of back diffusion of fission gases in the pump-shaft
annulus from the gas space in the pump tank to the region of the lower oil seal. A
radiation-safety review of the experiment was completed, and the design was revised
to provide additional safety. The experiment fabrication is complete. After cali-
bration of the instruments, the experiment will be connected to the PK-P loop for
testing.

(c) Thermal-Stress Analysis for Fuel Pump.--The most severe thermal stresses
in the pump tanklO have been found to occur in the cylindrical neck at the junctions
with the torispherical shell and the pump-tank flange. The high stresses are caused
by the high over-all temperature differences between the hot end (~ 1300°F) and the
cold end (200°F) of the cylinder, high thermal gradients, and rapid changes in
thermal gradient. Temperature distributions in the pump tank and shield plug were
determined for several operating conditions by the GHT (Generalized Heat Transfer)
code.l9 From these temperature distributions, it was concluded that controlled
cooling on the torispherical head would be reguired to maintain the pump tank within
acceptable temperature limits and to aid in reducing thermal stresses in the cylin-
drical portion of the pump tank.

Temperature profiles were estimated for the cylindrical portion of the pump
tank and the volute support for several degrees of cooling of the torispherical
shell. A metal temperature of 1000CF at the junction of the spherical head and the
cylinder and an assumed reasonable temperature profile produced the following
thermal bending stresses: at the volute, 5308 psi; at the sphere-neck junction,
16,900 psi; at the top flange, 24,000 psi. These stresses must be combined with
the pressure stresses to obtain the maximum principal stresses.

Revised GHT code input data have been calculated and machine calculations are
currently in progress to determine the temperature distributions more precisely
under several operating conditions. The GHT temperature distributions in the
cylindrical neck and volute support and in the torispherical head are shown in
Figs. 2.21 and 2.22, respectively. These distributions are for the operating con-
ditions of 10 Mw-reactor power, starting liquid level in the pump tank, and forced-
air cooling of the pump-tank wall. The peak temperature of 1370YF occurring between
the liquid level and the spherical shell is caused by the presence of windows in the
lower portion of the volute support, which increases the thermal resistance to
removal of the nuclear deposited heat in the volute support. This temperature is
believed to be excessive, and further refinements in the geometry and calculation
procedures are being made to reduce the peak temperature and gradients.

(d) Hot-Test Stand for Fuel Pump.--Mechanical, electrical, and instrumentation
designs for the fuel-pump hot-test standld were completed, and fabrication has
started. The test facility consists of a closed loop constructed of 6-in. Inconel
pipe. The loop contains a venturi flowmeter, a flow straightener, and an orifice
flow restrictor. The orifice is located in a freeze flange and is replaceable for
purposes of varying the system flow resistance. The system will be equipped with
provisions for measuring head, flow, pump speed, pump input power, and system tem-
peratures. The loop is equipped with a cooler section for assistance in temperature
control. The loop will contain the necessary preheaters and thermal insulation.

TEMPERATURE (°F)

(°F)

. 4300
CASE 2000
STARTING LEVEL
POWER, 10 Mw ' ' P
1200 FORCED - AIR COOLING

OUTSIDE SURFACE TEMPERATURE

52

UNCLASSIFIED |
ORNL—LR—-DWG 57680 .

1400

/’

LIQUID LEVEL

\

800 SPHERICAL
SHELL

600 : :
CASE 200 : \
STARTING LEVEL _
POWER, 10 Mw
FORCED -AIR COOLING |
400 ‘

200 .
0 2 4 6 8 10 12 14 16

AXIAL POSITION (in)

Fig. 2.21. Axial Temperotufe Distribution for MSRE. Fuel Pump; Volute Support.and Bowl Neck:. -

UNCLASSIFIED.
ORNL-LR-DWG" 57681

1400 -

/ LIQUID LEVEL

1000

900 - -
0] 2 4 6 8 10 42 4 16

RADIAL POSITION (in.)

Fig. 2.22. Radial Temperature Distribution for MSRE Fuel Pump; Sphericgl Head.
33

The pump tank and all the pump wetted parts will be constructed of INOR-8.
The pump support will simuiate the support proposed for use in the reactor. The
punp tank has provisions for testing the liquid-level indicators proposed for
reactor use. '

2.12.2 MSRE Coolant Pump

The design of the coolant-pump water test is essentially complete. The tests
will be used to determine the hydraulic performance of the pump and to determine
the feasibility of conducting the coolant-pump hot tests in the fuel-pump volute.
Drawings have been released and fabrication has started for the first series of
tests, which will evaluate the operating characteristics of the coolant-pump im-
peller in the fuel-pump volute.

2.12.3 Advanced Molten-Salt Pumps

Pump Equipped with One Molten-Salt Lubricated Bearing.--The pump containing
one salt-lubricated journal bearinglO continued in operation circwlating
LiF~-BeFp-UF), (62-37-1 mole %) at 1225°F, 1200 rpm, and 75 gpm. The pump has now
operated for 7440 hr and has been stopped and started 72 times. There has been no
discernible leakage of oil from the shaft seal.

2.12.4 Frozen-Lead Pump Seal

The small frozen-lead-sealed pump (Fig. 2.23) failed December 20, 1960, after
22,000 hr of operation.zoA It had operated continuously until the shaft jammed and
stalled the small fractional-horsepower motor drive. The 3/16-in. shaft parted
during attempts to free it, and the test was terminated.

The pump was removed from the loop and cut open as shown in Fig. 2.2k. The
left half is shown as-cut; the right half shows the barrel and shaft with the lead
and salt removed. The lower half of the pump barrel, as shown by area 6 in
Fig. 2.24, consisted of a porous, reddish-browh mixture of nickel and lead oxide.
The corrosion of the Inconel barrel (area 5 in Fig. 2.2k) is visible to the eye to
be over 1/8 in. deep at the point indicated. Approximately l/l6 in. of the 3/16-in.
shaft remains at this level.

The shaft had also been eroded, apparently by the abrasiveness of the oxide at
the frozen-seal area illustrated in the sectional view of Fig. 2.23 (see also
area 7 in Fig. 2.24k). The shaft parted at this point when attempts were made to
free it before shutdown. Further investigation will be made of the materials in

the pump.

2.13 MSRE ENGINEERING TEST LOOP

Design and fabrication of the l-l/2-in. Inconel system for the engineering
test loop2l were carried out during the report period. The piping is complete, and
instruments are being installed. ‘

The loop is situated above and below the track floor of Building 9201-3 as
shown in Fig. 2.25. Although this figure shows freeze flanges and a graphite con-
tainer, the first test will be operated with l-l/2-in. pipe throughout. The drain
line, dual freeze valves, and drain tanks are shown in Fig. 2.26 below the floor
level. The reservoirs shown on the vertical risers from each freeze valve are for
the purpose ol retaining sufficlent salt on draining of the loop to form a frozen
54

UNCLASSIFIED
ORNL—-LR-DWG 57682

i SALT INLET

y/A
Q

X~

=van

A\ SN\ sALT OUTLET
7 1210°F

MOLTEN-LEAD LEVEL

FROZEN-LEAD
SEAL AREA

810°F
AIR COOLING
TUBES — |

~— FRACTIONAL
H.P. MOTOR

Fig. 2.23. Sectional View of Lead-Sealed Pump Showing Temperatures of
Operation. :
55

UNCLASS;:;EqD
. HOTO
(1) Lead globules in the salt area of the volute 4%

(2) Impeller blade
(3) Salt globules in the lead area
(4) Un-oxidized lead

(5) Gross corrosion of pump body wall and shaft at the apparent interface

(6) Porous mass in lower pump body next to seal opening, consisting of
lead and nickel oxides

(7)

Seal area (not visible) where shaft was eroded in two.

Fig. 2.24. Frozen-Lead-Sealed Pump Body. Left half shown as terminated. Right half shown after
salt and lead were removed.
56

UNCLASSIFIED
ORNL-LR-DWG 54492

—-—SAMPLER-ENRICHER
CONNECTION

DANA PUMP
GAS

AIR LOCK
o
CONNECTION

CONTROLLED-
" ATMOSPHERE BOX

1%-in. FREEZE
1 FLANGES

|_—— REMOVABLE SECTION

|_——6-in. FREEZE FLANGES

"//,fé;T.mps(GRAN«TECONTAWER)

i
1
|
|
|

e

MANUAL
SAMPLER

- FUEL TANK

FREEZE VALVES

Fig. 2:25. MSRE Engineering Test Loop.

plug. The upper portion of the circulating loop is shown in Fig. 2.27. The DANA
pump bowl and support are shown to the right, with the shielding and inlet-pipe

support to the left. In the right foreground is the access to the drain-tank area
below the floor level.

In conjunction with the sampler-enricher system (to be tested on the engineer-
ing test loop later in the program), a capsule guide and retaineré2 was fabricated
and installed on the 1id of the DANA pump bowl 1lid as shown in Fig. 2.28. The
sampler-enricher system will be mocked up and attached to the tube (shown capped)
at a later date. Provisions are also included on the pump bowl for a manual salt
sampler23 and a differential-pressure type of level indicator.

The first hot tests to be conducted with the loop will determine the time
required to reach equilibrium oxygen concentration during the flushing operation

with salt (LiF-BeFp, 66-34 mole %) and will determine the operational characteris-
tics of the dual freeze valves.
Fig. 2.26. Drain-Tank and Freeze-Valve Area of Engineering Test Loop.

. UNCL ASSIFIED
PHZTO 36445

LS
Fig. 2.27.

=ncineering Test Loop Piping, Upper Section.

UNCLASSIFIED
PHOTO 35445

85
59

UNCLASSIFIED
36380

F'ig. 2.28. DANA Pump Bowl and Lid, Showing Sampler-Enricher Capsule Guide.
2.14 MSRE INSTRUMENT DEVELOPMENT

2.14.1 Thermocouple End Seals

End Seal for l/8-in.-OD Oxide-Insulated, Two-Conductor, Sheathed Thermocouple Wire

The insulation for most sheathed thermocouple wire is chemically hygroscopic;
consequently, the sheath must be made hermetically tight. A liquid sealant for
sheathed thermocouple wire is commercially available, but will not meet the specifi-
cations for a reactor-grade product. The liquid sealant is unsatisfactory because
of the effects of radiation and high-temperature damage. Moreover, this sealant
cannot be applied in the field since a furnace is needed to heat the sheathed
thermocouple before the sealant is applied. This process takes approximately 8 hr.

Figures 2.29 and 2.30 show an end-seal assembly for a l/8—in.-OD oxide-
insulated sheathed thermocouple, which can be installed in the field. The seal con-
sists of two parts: the seal holder and the cable end bell. The seal is hermetic
when installed according to the following procedures:

1. Remove the sheath material and oxide insulation to expose the elec-
trodes, making certain the sheath material and electrode have not
been electrically shorted.
60

2. Attach the cable end bell to the sheath by soldering, brazing, or
welding, whichever suits the intended application.

3. Insert the electrode through the seals.

4. Engage the electrode holder with the attached end bell and seal
this joint as the end bell was attached to the sheath.

5. ©Seal the electrode to the seals by soldering.
The seal should then be tested by the following technique:

l. Jig the thermocouple in an assembly that can be pressurized
with helium to 100 psi.

2. Leave the seal in the pressurized jig for 1 hr.

3. Remove the pressure.

k., Pour acetone in the jig. Bubbles in the acetone indicate a leak
in the seal. (A helium leak detector could be used instead of
the acetone.)

5. Repair all leaks and proceed from step 1.

UNCLASSIFIED
PHOTO 36573

CABLE END BELL

FEEDTHROUGH
ELECTRODE HOLDER

Ye-in-0D INCONEL-SHEATHED
CHROMEL-ALUMEL THERMOCOUPLE

Fig. 2.29. End Seal for ‘/8-in.-OD Sheathed Thermocouple.

UNCLASSIFIED
PHOTO 36578
FEEDTHROUGH CABLE END BELL
ELECTRODE HOLDER

Yg-in.-0OD INCONEL-SHEATHED
CHROMEL-ALUMEL THERMOCOUPLE

Fig. 2.30. End Seal for ‘/B-in.-OD Sheathed Thermocouple with End Bell Assembled.
61

End Seal for l/l6-in.-OD Oxide-Insulated, Single-Conductor, Sheathed
Thermocouple Wire

The use of two-wire single-conductor sheathed thermocouples is being proposed
for the MSRE. The problems associated with two-conductor cables also apply to
single-conductor cables. Figure 2.31 shows a typical thermocouple jack and two
single-conductor sheathed cables with end seals attached. The attachment of a
single=-conductor end seal is similar to that of a two-conductor end seal. The
sheath material for the thermocouple wire can be stainless steel, aluminum, copper,
iron, Inconel, INOR-8, or some other material. Aluminum-sheathed cables are more
difficult to seal than other materials. Inconel or INOR-8 will be used for the
sheath material in most MSORE applications. The plug and Jjack are fabricated from
Electrobestos, which can be purchased from numerous manufacturers. This assembly
is designed for use in a nuclear environment at elevated temperatures.

UNULASSIFIED
ORNL—LR—DWG 56084A

STANDARD THERMOCOUPLE JACK——__ : |

ZZ il
2

\

J«Q;;// //C;;;;;¢%/ Y46-in. OD INCONEL SHEATH
Géész駢¢K\\WMNERALINSULAflON
/ NO.22 GAGE WIRE

Fig. 2.31. End Seal for l/lé-in.-OD Mineral-Insulated Thermocouple and
Typical Thermocouple Niscannect.

2.14.2 Instrument and Power Disconnects

Ganged-Thermocouple Disconnect for Remote Operation

Figure 2.32 shows a multithermocouple disconnect for remote operation. The
unit is operated by a straight-line push-pull motion, with the alignment being
62

UNCLASSIFIED
PHOTO 36397

Fig. 2.32. Multithermocouple Disconnect.

accomplished by the aid of guide pins of different diameters. The guide pins also
act as a polarizing key for the disconnect. The unit has been operated very satis-
factorily in the remote-maintenance demonstration facility. The basic unit as
shown can be modified to accommodate as many thermocouples as the user might re-
quire. For the MSRE, the basic four-thermocouple unit will be used and will be
modified for two and eight thermocouples. This unit is fabricated from Unistrut,
Unistrut fittings, and Electrobestos plugs and jacks for an elevated-temperature,
nuclear environment. Approximately 50 1b of force is required to operate the
basic four-thermocouple unit.

Power Disconnects for Three-Phase Operation

Figure 2.33 shows an electrical disconnect for three-phase power operation.
This unit has an electrical rating of 4O amp, three phase, 600 v ac; a similar unit
has a rating of 70 amp, three phase, 600 v ac. Both were designed for remote
handling and a nuclear radiation environmment. Inorganic materials are used in their
fabrication; the metallic parts are aluminum and silicon bronze, and the insulating
materials are porcelain and lava. The units are operated by a straight-line push-
pull motion with a force of less than 25 1b required for operation. Intended for
nonpolarized balanced-load service, they can be polarized by the addition of a
guide key. Both units have been tested in the remote-maintenance demonstration
facility and are easily operated. The units will be fabricated by an outside manu-
facturer in accordance with ORNL designs and specifications.

A 150-amp, three-phase, 600-v ac disconnect of construction and design similar
to the two above is being considered for use in the MSRE for pumps and tank heaters.

Four-Pole Instrument Disconnect

Figure 2.34 shows a prototype instrument disconnect, with a maximum rating of
100 v ac or dc, which is being developed for use in the MSRE. The unit is fabri-
cated from inorganic components; the insulating material is fired lava and the
housing is aluminum. The unit is self-polarizing due to the conical electrode
holder. It is not designed to be disconnected while energized. A four-conductor
sheathed electrical cable will be used for the power to this disconnect.
63

UNCLASSIFIED
PHOTO 36577

Fig. 2.33. Power Disconnect for Three-Phase Operation.
64

UNCLASSIFIED
ORNL-LR-DWG 56159A

PULL BAIL —=

ESSS N

j=—HOUSING
RECEPTACLE ——

ELECTRODE
L AVA INSULATOR

LOCK RING

LAVA INSULATOR
ELECTRODE

—

LOCK PIN rO

HOUSING
©

b
il

A NEANEERLALALA AU RRNRRANRR AR

MOUNTING
BRACKET

PLUG

THREADED GLAND
SHEATHED CABLE

Fig. 2.34. Instrument Disconnect, Four Pole.

A power disconnect of similar construction is being fabricated for 90-amp,
three-phase, 600-v ac service and should be ready for testing within 60 days.

2.14.3 Pump-Bowl Level Indicator

MSRE instrumentation requirements include the measurement of the level of the
molten salt, over a 5-in. range, in the fuel- and coolant-salt pump bowls. In
order to make this measurement, a level-sensing device is required that is capable
of extended operation at 1200°F in an intense radiation field (y + n + B). Some
considerations affecting the selection of the level-sénsing element are: contain-
ment of the molten salt and cover gas, corrosion, temperature and pressure effects
on the zero and range of the instrument, long-term material damage from temperature
and radiation, possible effects of ZrF) snow accumulations, gamma and beta heating
effects on materials in the gas space, simplicity of construction, and compatibility
with commercially available receiving instruments. A further restriction is the
relatively low electrical conductivity of the fuel and coolant salts.

A number of level-measuring techniques were considered; however, no presently
available device was suitable for MSRE use without some development. While several
types of level devices offered possibilities of successful development, the ball-
float type used on the ARE and further developed by Southernzh offered the greatest
possibility of success with the least development. The major problems associated
with development of the ball-float sensor are the design of the float and the design
of the position-sensing coil and core assembly.
65

The design of the float was the first problem considered. Two types of floats
are being designed. The float used in the fuel system must withstand the high tem-
perature of the fuel salt and the still higher temperature caused by beta heating.
A solid graphite float is being designed for this service. The submerged portion
of the float will be cylindrical, and the unsubmerged portion will be spherical.
Calculations show that the density difference between graphite and the fuel salt is
sufficient to produce the desired excess buoyancy when allowance is made for absorp-
tion of fuel salt in the graphite. The solid graphite float cannot be used in the
coolant-salt system because the density of the coolant salt is too low. The float
in the coolant salt will not be subjected to beta heating, however, and a hollow
INOR-8 ball float can be used.

The position of the float will be detecled by measuring the position of an
iron core, suspended below the float inside a contaimment tube, as shown in
Fig. 2.35. A differential transformer located outside the contaimment tube will
detect the position of the core through the nommagnetic wall of the containment
tube. The core will be clad with INOR-8 for corrosion protection.

The differcntial transformer was selected for the position-sensing device in
preference {0 the taper-core - variable-inductance sensors previously used because
it is less affected by temperature, it is compatible with commercially available
receiving instruments, the core weight is lower, and the cylindrical configuration
of the c¢ore eliminates the nccd for guide bushings.

The high temperature at which the differential transformer operates presents
several problems. The conducting and insulating materials in the transformer wind-
ings must maintain their physical and electrical characteristics at the exposure
temperatures and radiation levels for periods longer than the longest expected
period of reactor operation. Also, the inherent decrease in insulation resistivity
with temperature must not affect the transmitted signal.

Silver was originally considered for the wire material, but because of grain
growth of silver at the operating temperature, other materials, including platinum,
are now being considered. The insulation problem is being attacked from two direc-
tions. First, it was decided that low-impedance windings would be used. This will
permit the use of heavy-gage ceramic-coated wire and will reduce the effect of
changes in the resistivity of the insulation on the transmitted signal.2> Second,
the available insulating materials are being investigated to determine which will
best meet our needs. The usual ceramics are available, and a newer material,
Flourmica, may be applicable. One Flourmica sample has been submlitted to the Solid
State Division for evaluation. Another sample of a difterent type will be sub-
mitted as soon as it arrives.

A low~temperature mockup of the differential transformer with the desired 5-in.
range has been constructed and tested. A sensitivity of approximately 40 mv/v/in.
was obtained. Deviation of the transmitted signal from linearity was less than 2%.
The transformer consists of two single-layer windings. Each winding has 436 turns
of No. 24 AWG wire and is 10 in. in length. The secondary has the conventional
differential arrangement except that half the turns are wound in one direction and
half in the other, with no breaks or connections in the wire. The primary is one
continuous coil wound in one direction. This transformer will be developed further
to incorporate high-temperature wires and insulating materials, and additional tests
will be made at temperature.

A test stand for testing prototypes of the level device at high temperature
and in molten ocalt is now lu Lhe design stage.
VENT TO

66

PUMP —a—

BOWL

FROM

PUMP —e {

BOWL -

FOXBORO
‘DYNALOG
READOUT

Fig. 2.35.

A

A S SN SS SN

|

.

UNCLASSIFIED
ORNL—LR—-DWG 57683

DIFFERENTIAL
TRANSFORMER CORE

DIFFERENTIAL
TRANSFORMER

%”//////,/—DRAINLJNE

Continuous Level-Indicating Device for Pump Bowl.

P S
67

2.14.4 Single-Point Level Device

The fuel salt and the coolant salt for the MSRE are contained in four tanks:
two for the fuel used during normal operation, one for fuel which is used to flush
the system, and one for the coolant salt. The weight of fluid in these tanks is
measured by individual weighing systems, which provide a continuous indication of
the amount and level of fluid in each tank. For a given weight of fluid in a tank
the depth of the fluid can be calculated, and this calculation can be checked by
measurement, if the measurement is made prior to the time the reactor starts normal
operation. With this level information available, the calibration of the weighing
system can be checked if one or more fluid levels can be determined accurately. In
the past the spark-plug probe has been used to give positive indication of a single
fluid level. This device is very accurate and will provide a level indication that
can he nsed to calibrato o weighing 3ysleu.

Unfortunately the spark-plug probe has an insulator, to support and insulate
the center electrode, which becomes a part of the contaimment. vessel. The seal
between the metal housing and the ceramic insulator of the spark plug is not leak-
tight and will not contain the gases in the system. Also, the insulator inside the
container accumulates material, from the vapor in the container, which short-circuits
it.2 For these reasons the use of a spark-plug probe, or any device that requires
an insulator as a part of the containment vessel, cannot be tolerated. As there is
no similar device which will provide single-point level indication, development of
a single-point probe or level device became necessary.

During the ANP program congsiderable work was done on the evaluation of liquid-
metal level transducers.Za Paragraph 3.2 of ref 26 describes a probe that was
designed to replace the spark plug. This basic design is the one around which the
development is centered. The probe described in ref 26 was designed to be used in
NaK, which has a high elcctrical conductivity. The molten tiuorides, on the other
hand, have a very low conductivity, which makes changes in the design necessary.
The original design used the change in voltage across the probe, when NaK came in
contact with the end of the probe, to operate a relay. As NaK is a good conductor,
Lhere 1s a large current change, which results in a voltage change sufficiently
large to operate a relay. With the MSRE molten fluoride mixture, the change in
current flow is relatively small; so a signal transformer is inserted in series with
the excitation transformer and the probe. The output of the signal transformer is
used as the desired signal (see Fig. 2.36). By keeping the voltage drop in the
circuit external to the probe a very small fraction of the voltage drop across the
probc, the small resistance change that takes place when molten salt touches the
probe results in a usable cignal from the signal transformer. With the present
probe the maximum signal change is about 5%. Although detectable, this signal
change is not large enough for normal use. The probe now being used consists of a
5/8-in. Inconel tube with two ceramic-bead-insulated No. 18 silver wires run down
the center and welded to the contact end of. the probe.

The molten-salt ¢ontainer in which the first tests were made with this device
is being modified to make it more useful. The probe design is also to be changed.
With these changes the development program for the device will have three major
phases. One will be to continue work on the probe and the associated electrical
excitation and readout system. Another will be a continuation of the work done by
Greene2l on the conductivity of molten fluorides. The third will be a program to
determine the cause of what seems to be the polarization of electrical probes in-
serted in the salt even when excited with high-frequency a-c voltages. This polar-
ization was mentioned by GreeneZT ang by Hyland2 in his work on molten-salt level
devices.
68

UNCLASSIFIED
ORNL— LR—DWG 57684

AC EXCITATION READOUT

(S

AC CURRENT -z | By
PATH —

AL RA AL ARA SR A A WA\ ¥

MR ALY

SALT BELOW SALT ABOVE
PROBE TIP PROBE TIP

Fig. 2.36. Single-Point Level Device.

REFERENCES < -

Sturm-Krouse, Inc., Analyses and Design Suggestions for Freeze Flange Assemblies
for MSRE, Nov. 30, 1960.

MSR Quar. Prog. Rep. July 31, 1960, ORNL-301k, p 24-25.

J. C. Moyers, Thermal Cycling Test of 3 1/2 in. and 4 in. Freeze Flafigeg,
ORNL CF-61-2-38 (Feb. 2, 1961).

MSR Quar. Prog. Rep. July 31, 1960, ORNL-301L,.p 25.

See Sec. 1.8 of this report.

R. B. Gallaher, Modification of MSRE Sampler-Enrlcher Entry into Pump Bowl,
MSR-60-33 (Nov. Ik, 1960).

C. H. Gabbard, Maximum Temperature in MSRE Sampler-Enrlcher Guide Rods,
memorandum to R. B. Gallaher (Dec. 7, 1960}).

G H. Llewellyn, MSRE Maxirmum Temperature in Sample Capsule Guide Rods of Fuel
Sampler-Enricher, ORNL CF-61-1-58 (January 1961).

24,

25.

69

MSR Quar. Prog. Rep. July 31, 1960, ORNL-301k4, pp 4, 27.

Tbid., p 27.

MSR Quar. Prog. Rep. Apr. 30, 1960, ORNL-2973, p 8.

MSR Quar. Prog. Rep. July 31, 1960, ORNL-301k, p 27-28.

P. P. Holz, Status of Small Pipe Disconnects for MSRE Auxiliary Lines,
ORNL CF-60-9-102 (Sept. 27, 1960).

P. P. Holz, Interim Report, Status of Small Pipe and Tube Disconnects for MSRE
Auxiliary Lines, ORNL CF-61-2-58 (Feb. 21, 1961).

MSR Quar. Prog. Rep. .mly 31, 1960, ORNL-301k, p £0.

MSR Quar. Prog. Rep. Apr. 30, 1960, ORNL-2973, p 1lZ2.

MSR Quar. Prog. Rep. July 31, 1960, ORNL-3014, p 2L.

Tbid., p 30.

T. B. Fowler and E. R. Volk, Generalized Heat Conduction Code for the I.B.M.
704 Computer, ORNL-2734.

MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL-2626, p 23.

MSR Quar. Prog. Rep. July 31, 1960, ORNL-301L4, p 31.

R. B. Gallaher, Modification of MSRE Sampler-Enricher Entry into Pump Bowl,
MSR-60-33 (Nov. 1L, 1960).

J. L. Crowley and W. B, McDonald, A Sampling Device fur Molten-salt Systems,
ORNL-2688. '

A. L. Southern, Closed Loop Level Indicator for Corrosive Liquids Operating at
High Temperatures, ORNL-2093.

R. L. Moore, "The Differential Transformer in High-Temperature Nuclear-
Radiation Enviromments,'" Proceedings of the Instrument Society of America
Second Biannual National Nuclear Instrumentation Symposium, vol 2, p 107.

R. G. Affel et al., Fluld Metal Level Transducer, ORNL CF 58-8-19 (August 1958).

N. D. Greene, Measurements of the Flectrical Conductivity of Molten Fluorides,
ORNL CF-54-8-6L4 (August 1954).

R. F. Hyland, unpublished report.
3. REACTOR ENGINEERING ANALYSIS

3.1 CRITICALITY CALCULATIONS

‘Three sets of survey calculations were performed, using the current MSRE fuel-
salt composition (LiF-BeFpo-ZrF4-ThF4-UF4, T0=-23-5-1-1 mole %) and basic core design
(right-circular cylinder, 27.7 in. in radius by 63 in. high). In the first set,
the core was assumed to be a single region consisting of a homogeneous mixture of
graphite and fuel salt, with various volume fractions of salt. In the second, the
core was divided into two or three regions having different volume fractions of
‘fuel salt and graphite. In the third set, the core was made up of fuel salt and
a reflector of moderator material (graphlte, beryllium, or beryllium oxide), and
in some of these cases a central moderator region was also considered. All calcu-
lations were done with the IBM-704 multigroup diffusion-theory program GNU-II.

Calculated system inventories of U=35 and critical mole percentages of uranium
are plotted for the first set in Fig. 3.1; some typical two- and three-region re-
sults are listed in Table 3.1l. It should be noted that while some reduction in
circulating system inventory was obtained, the effect of subdividing the core was
slight; it is possible, however, to obtaln a wide range of powver shapes, as indi-
cated in Fig. 3.2.

Some typical results obtained for the third set of calculations (reflected
reactors) are listed in Table 3.2. In order to achieve a critical uranium concen-
tration in the salt as low as 0.2 to 0.3 mole %, it was necessary to utilize an
island reflecter of beryliium or beryllium oxide.

Table 3.1l. Results of GNU Calculations for Two- and Three-Region
Reactors Having Different Fuel Volume Fractions
in the Various Regions

. Fraction of Core Circu. .System Critical

No. of Fuel Volume Fraction Cross=-Sectional Area Inventory Conc.,
Regions Inside Middle Outside Inside Middle Outside (kg U-235) (mole % U)

2 0.2k - 0.8 0.50 - 0.50 46.5 0.217

2 0.2k - 0.18 0.70 - 0.30 48.4 0.221

3 0.25  0.13 0.07 1/3  1/3 1/3 4,2 0.243

3 0.0 0.13  0.07 S 53.8 0.272

3 0.0 0.13  0.07 54.0 0.324

3 0.25  0.13 0.0k | ‘ 43.3 0.243

3 0.25 0.06  0.04 , b4 ,2 0.257

70
71

Table 3.2. Results of GNU Calculations for
Homogeneous Molten-Salt Reactors

Per cent Median
Reflector Mole % Core Critical Thermal Fissioning
Material Uranium Mass (kg U-235) Fiecions Energy (ev)
5-in. Reflector Thickness, No Island
c 1.04 250 4.6 150-400
Be 0.72 175 20.9 50-65
BeO 0.76 186 16.0 80-90
10-in. Reflector Thickness, l-ft-diam. Island
+*
C 0.67 3 33.2 20=25
*
Be 0.25 3k 67.3 Thermal
*
BeO 0.28 39 62.0 Thermal
*Island and reflector material.
UNCLASSIFIED
ORNL—-LR—- DWG 57685
70
60 ‘\\
W
2 \
> /O
U 50 e EEa—— =
p \\O____()—/O’
p
(0 ®
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=2
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= 30 - 0.3
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= €
3 20 0.2 =
3 s
o 2
© &
2
10 0.1
0 0
5 10 15 20 25 30

FUCL SALT (vel %)

Fig. 3.1. GNU Results for One-Region Reactors.
72

UNCLASSIFIED
ORNL—LR~— DWG 55954A

25
- | |
{0.10, 0.13, 0.07)
— \
= |
é 2.0 T~
I T $ UNIFORM
Z .
@
Y \\\\
=15 (0.25, 0.13, 0.07) (0.40, 0.13, 0.07) —
: >
<I
=
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X 1.0 \
2 AN\
(0.25, 0.13, 0.07) \
-
2 \
a
L 05 UNIFORM N\
- |
(0.10, 0.13, 0.07)
o | !
0 5 10 (5 20 25 30 35

SPACE POINT NUMBER

Fig. 3.2. Comparison of Thermal-Flux Shapes in Three-Region Reactors.
Numbers in parentheses refer to fuel volume fraction in inner, middle, and outer
region of reactor, respectively.

3.2 REACTIVITY WORTH OF CONTROL TUBES

The control-tube=worth calculations were done.in x-y geometry using the- IBMp?OM
program . PDQ® with two groups. For the purposes of these calculations the.actual :
cylindrical tubes were replaced by square-tubes of cross-sectional :area equal to

the cylindrical tubes. Two=group constants derived from multigroup calculations
were used for the core and empty~tube regions; the full tubes were assumed to be

. black to thermal neutrons, but to have the same fast-group cross sections as the
empty tubes. A cross-sectional view of a control-tube cell is shown in Fig. 3.3.
(It should be noted that an "empty" tube contains fuel salt and graphite, as well

as INOR-8.) A total worth of 11% 6k/k was obtained for four tubes, when these .

© tubes were symmetrically placed 4 in. from the vertical axis of the core.

Additional calculations were performed using the two-group r-z geometry
IBM-T090. program BQUIPOISE-2 (ref 3) to obtain estimates of the reactivity worth
of the tubes as a function of the height of the poison fluid. Since r-z geometry
was used in this calculation, the four tubes were replaced by a ring; this ring
had the same surface-to-volume ratio as did the tubes. The calculational results
were normalized to the 11% result obtained from PDQ. The results of these calcu-
lations are shown in Fig. 3.4, with fraction of total tube worth plotted versus
the fraction of the tubes filled with poison salt. A curve of the approximate
tube worth obtained from elementary perturbation theory is included for comparison.
UNCLASSIFIED
ORNL—LR—DWG 57686

— FUEL SALT

INOR-8

— POISON-SALT ZONE

[~ vOID

—— GRAPRKITE

DIMENSIONS ARE IN INCHES

Fig. 3.3. MSRE Control Tube (Dimensions as Shown).

73

UNCLASSIFIED
ORNL—LR-DWG 57687

1.0 —

0.8
= EQUIPOISE-2
& /
3
T, 086
g /
-
(o]
P~ ®
(o]
Z 0.4
Z o.
5 4
é . INTEGRAL. SINE. SQUIARFD

0.2 9/4,

7
o}
o} 0.2 04 06 0.8 1.0

FRACTION OF TUBES FILLED

Fig. 3.4. Reactivity Worth of Partially Filled Control

Tubes.

3.3 PRELIMINARY CALCULATIONS FOR CORE CONTAINING INOR-8 TUBES

As an alternative to the present MSRE design it has been proposed4 to can
the core graphite by means of an INOR-8 calandria, with fuel salt flowing through

tubes,

Preliminary nuclear calculations have been made, assuming the core of the

‘reactor to consist of a homogeneous mixture of fuel, graphite, and INOR-8, with a
nominal fuel vnlume fraction of 0.08.
different tube thicknesses,

Table 3.3.

GNU Results for Reactors with Tubes

Results are listed in Table 3.3 for three

Fission Neu-

Neutrons Ab-

. trons Pro- Leakage per sorbed in Th

Critical Circ. System duced per Neu= Neutron Ab- per Neutron

Tube Thick- Conc. Inventory  tron Absorbed sorbed in Absorbed in

ness (in.) (Mole % U) (kg U=-235) in Fuel Fuel Fuel

0 0.39 61.8 2.023 0.851 0.076
0.02 0.56 89.5 2.008 | 0.733 0.063
0.03 0.66 104 2.000 0.695 0,058
0.04 0.75 119 1.993 0.665 0.055

3.4 MSRE BIOLOGICAL SHIELDING AND ASSOCIATED DOSE RATES

Estimates were made® of the gamma-ray source strengths and some of the result-

ing biological shield requirements for the MSRE.

A summary of the calculational

results is presented in Table 3.4; the dose rate for a given shield thickness and
74

Table 3.4, Summary of MSRE Shielding Calculations

Attenu~ Shield Thickness Change in Thick-

Power Level ation and Associated ness for Tenfold
or Cooling Distance Dose Rate? Change in Dose
Shield Time (£t) (£t and mrad/hr) Rate (ft)
Reactor-cell top plug 10 Mw 16 7.0 (ord) 92 1.29
Drain-tank-cell top plug 0 hr 27 6.0 (ord) 62 1.13
2 hr 27 6.0 (ord) 0.47 0.86
Side wall® 2 hr 16 3.0 (bary) 25 0.52
24 hr 16 3.0 (bary) 5.6 0.49
Storage-tank top plug 106 sec 8 5.0 (ord) 4.2 0.83
Fuel-carrier lead shield 120 days - 10.0 in. Pb 1.8 1.6 in.
Maintenance control room
Activated INOR-8° 30 days 15 2.5 (bary) 18 0.40
1% fission products® 30 days 15 2.5 (bary) 3 0.50
10% fission products® 1 day 30 2.5 (bary) 50 0.53
Piping~disconnect cell
Activated INOR-8¢ 30 days 5 5 in. Pb 6.5 -
Radiator-pit penthouse 10 Mw - 2.0 (ord) 2.3 -

aComposi’cion in parentheses: ord, ordinary concrete, p = 2.3 g/cc
bary, barytes concrete, p = 3.5 g/cc.
This considers that particles strike the wall at an oblique angle.

Reactor vessel, one year of irradiation at 10 Mw.
Assumed holdup in reactor vessel.
Assumed source from off-gas holdup tank.

o 2 0 o

composition, the power level or cooling time, and the distance from the source to
the dose point are given. In addition, the change in shield thickness required for
a tenfold change in the dose rate is specified and is applicable over the range of
at least one-tenth to ten times the given dose rate (too large an extrapolation in
either direction will give a low estimate). In order to decrease the dose rate
above the top plugs to permissible levels, a thicker shield or better shield ma-
terial is required.

An estimate of the residual activity in the INOR-8 piping of the coolant-salt
system due to activation by photoneutrons (after one year's operation at 10 Mw)
gave less than 1 mrad/hr at a distance of 6 in. from the piping. The photoneutrons
are produced by the Be9(7,n)Be8 reaction in the coolant salt.
75

REFERENCES

C. L. Davis, J. M. Bookston, and B. E, Smith, GNU=-II - A Multigroup One-
Dimensional Diffusion Program for the IBM=-704, General Motors Report GMR=-10l

G. G. Bilodeau et al., PDQ - An IBM-704 Code to Solve the Two-Dimensional Few-
Group Neutron-Diffusion Equations, WAPD=-TM=TO (1957).

T. B. Fowler and Melvin Tobias, EQUIPOISE-2: A Two-Dimensional, Two~Group
Neutron-Diffusion Code for the IBM=-7090 Computer, ORNL CF-60-11-67 (Nov. 1960).

E. S. Bettis, personal communication.

D. W. Vroom, MSRE Gammé Ray Source Strengths and Biological Shielding Re=-
quirements, (ORNL CF memorandum, to be issued).

- THIS PAGE
WAS INTENTIONALLY
LEFT BLANK
PART Il. MATERIALS STUDIES

4. METALLURGY
4.1 DYNAMIC-CORROSION STUDIES

4.1.1 Status of Test Program

As discussed in Sec. 2.10, the 24 forced-convection lobps progra.mmedl to
evaluate the corrosion rates of INOR-8 or Inconel in fused fluoride mixtures
have now completed operation. Five operated for more than two years and 19 for
one to two years. ' .

Eight loops are now being examined. A summary of the test findings for the
16 loops that have been examined is presented in Table L.1. Results for the last
four loops listed in Table 4.1 have not been previously reported and are discussed
below.

4.1.2 Examination of INOR-8 lLoops

Metallographic examination of INOR-8 forced-convection loop MSRP-12, which
circulated LiF-BeFo-UR,-ThF). (62-36.5-0.5-1 mole %) for 14,498 hr at a maximum
temperature of 1300°F, revealed no manifestation of attack except the development
of a thin surface layer and slight depletion of second-phase material below the
exposed surface. The extent of this layer and depletion zone can be seen in
Fig. b4.1. ‘

The salt was sampled frequently during circulation in this loop so that
chromium buildup could be monitored as a function of operating time.2 As shown
in Table 4.2, the samples revealed a slight increase in chromium concentration
from an initial value of about 450 ppm to a level of 500 ppm after 1200 hr.
Within a spread of #60 ppm, this level remained constant throughout the test.
The concentrations of nickel and iron remained at 60 + 20 ppm and 210 % 70 ppm,
respectively, during the test. ’

A somewhat higher rate of corrosion was incurred by INOR-8 loop MSRP-13,
wvhich circulated a fluoride mixtwre containing a large uranium concentration
(LiF-BeF,-UF),, 70-10-20 mole %). As discussed previously,3 operation of this
loop was terminated after 8085 hr when a pump failure and subsequent startup
caused a rupture of the cooling coil. Metallographic examination of specimens
removed from both heater legs and the unheated bend comnnecting the two legs
revealed concentrated attack in the form of surface roughening and pitting to a
maximum depth of 1-1/2 mils. The cold-leg sections revealed only moderate surface
roughening; however, examination of the cold-leg specimens did reveal a small

77
Table 4.1.° Results of Metallographic Examinations of Forced-Convection Loops

130 LiF-BeFy-UF), 62-37-1 mole %
131 LiF-BeF5-UFy, 60-36-4 mole %

133 LiF-BeFp-ThFy, 71-16-13 mole %

134 LiF-BeF,-UF),~ThF),, 62-36.5-0.5-1 mole %

136 LiF-BeF,-UF), 70-10-20 mole %.

Maximum .
Test Fluid-Metal ' Flow
Loop , Period Salt Interface AT Reynolds Rate
No. Material (hr) Mixture* Ternp (OF) (oF) No. (gpm) Metallographic Examination
9075-1 - Inconel 8,801 122 1250 100 5000 2.0 Intergranular penetrations to 1% mils@
93kLh-1 Inconel 8,892 123 1300 200 3250 2.0 Heavy intergranular voids to 38 milsb
93u4h-2 Inconel 8,735 12 1200 100 8200 2.5 Intergranular voids to 8 milsb
9377-1 Inconel 3,390 126 1300 200 1600 2.0 Intergranular voids to 7 milsa
9377-2 Inconel 3,046 130 1300 200 . 3000 2.0 Intergranular voids to 8 milse
9377-3 Inconel 8,76k 131 1300 200 3400 2.0 Intergranular and gen. voids to 14 milsb
- 9377-k Inconel 9,57& : 130 1300 200 2600 1.75 Intergranular and gen. voids to lh:mils?
9354-1 INOR-8 1k, 563 126 1300 200 2000 2.5 Heavy surface roughening and pitting -
- ‘ . to 13 milsC
9354-4  INOR-8 15,140 130 1300 200 3000 2.5  No attackb,d
9354-5 INOR-8 lh,SOB 130 1300 200 3000 2.5 No attackb
MSRP-8 INOR-8 9,633 124 1300 200 - 4000 2.0 No attackC
MSRP-9 INOR-8 9,687 134 1300 200 2300 1.8 No attackC
9377-5 . Inconel 15,038 134 1300 200 2300 1.8 Intergranular voids to 24 mils-
9377-6 Inconel 13,155 133 - 1300 200 3100 1.8 Intergranular voids to 13 mils
MSRP-12  INOR-8 14,498 134 1300 1200 2300 1.8  No attack
MSRP-13  INOR-8 8,085 136 1300 200 - 3900 2.0 Heavy surfacé reoughening and -pitting
to 1% mils :
* .
"Salt Compositions:
12 NaF-LiF-XKF, 11.5-46.5-42 mole % _
122 NaF-ZrF)-UFy, 57-42-1 mole % 8MSR Quar. Prog. Rep. Oct. 31, 1958, ORNL-2799, p 54-55.
123 NaF-BeF5-UFy, 53-46-1 mole % Dure
12k NaF-BeF,- " 58-35-7 mole % MSR Quar. Prog. Rep. Oct. 31, 1959, ORNL-2890, p 35-L2.
126 LiF-BeF5-UF), 53-46-1 mole % CMSR Quar. Prog. Rep. Jan. 31 and Apr. 30, 1960, ORNL-2973,

p 36-308.
dMsR Quar. Prog.

Rep.

July 31, 1960, ORNL-301Lk, p 55-58.

8L
79

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Fig. 4.1. Appearance of Specimen Removed from Point of Maximum Salt-Metal

Interface Temperature (1300°F) of INOR-8 Forced-Convection Loop MSRP-12.

loop was operated for 14,498 hr with LiF-Ber-UF4 (62-36.5-0.5-1 mole %). Etchant:

3 parts HCI, 2 parts H,0, 1 part 10% chromic acid.

Table 4.2. Chemical Analyses of LiF-BeF,-UF)-ThF), (62-36.5-0.

mole %) Forced-Convection Loop MSRP-12

The

5-1

Major Minor
Time Constituent (wt %) Constituent (ppm)

Semple (hr) ] Th NL

Cr

Fe

2.79 6.1k 50

Operating fill 2.62 6.04 5
Operating 2k 2.72 6.04 65

L8 2.56 5.95 60
120 2.55 5.90 45
290 273 5.62 75
450 2.80 5.98 80
790 2.83 6.10 80
1,800 R0 6.30 55
2,298 3.07 5.61 L0
3 TLT 2.98 5.52 15(7)
13,700 2.64 10.4 40

Termination 14,498 2.72 8.67 45

325
490
460
4h5
455
510
475
505
4hs5
555
555
560
415

80

number of magnetic metal crystals loosely adherent to the cold-leg wall.
Results of chemical analyses of these deposits are not yet available.

A sampling probe was used in this loop experiment to determine the buildup
of chromium in the salt as a function of operating time.2 Samples of the salt
showed an increase in chromium concentration from an initial level of 160 ppm to
a final level of 1730 ppm (Table 4.3). The concentration of iron increased from
an initial level of 175 ppm to a final level of 475 ppm. The nickel concentration,
which was abnormally high at the start of the test, decreased to a level of 100
ppm. The uranium concentration remained relatively constant during the test.

The chromium concentration when plotted as a function of test time was
found to closely approximate a function of the type ekt. Such a time depend-
ence is typical of reactions whose over-all rate increases as the concentration
of reaction products increases. It is believed that the high initial nickel
concentration in the salt contributed to the buildup of chromium. However, it
has not been possible to propose a mechanism which satisfactorily explains the
ekt dependence.

Despite the initially high nickel concentration and the relatively large
percentage of UF) (20 mole %) incorporated in the salt systems, the maximum
attack of INOR- 8 was limited in all specimens to 1-1/2 mils. Thus the test
lends encouragement to the use of salt mixtures containing the larger concentra-
tions of uranium required for one-region power-convertor reactors. The indica-
tion of mass transfer, however, will require further study.

Teble 4.3. Chemical Analyses of LiF-BeF -UF), (70-10-20 mole %)
Circulated in INOR-8 Forced-Convecglon Loop MSRP-13

Time Uranium Minor Constituents (ppm)
Sample (hr) (wt %) Ni Cr Fe
Batch 55.3
Operating fill 52.8 110 210
Operating 1 54.1 660 160 175
120 54.0 685 165 170
825 53.8 100 215 160
1550 54.1 100 230 165
2290 54.5 100 305 210
3990 54.1 100 535 230
5275 54.3 100 820 270
6950 54.5 100 1095 300
Termination 8085 50.9 100 1730 475

4.1.3 Examination of Inconel Loops

Metallographic examination of Inconel forced-convection loop 9377-5,
which circulated LiF-BeFo-UF)-ThF), (62-36.5-0.5-1 mole %) for 15,038 hr at a
maximum temperature of 1300CF, revealed heavy intergranular and general sub-
81

surface voids ranging in depth from 12 to 22 mils in the first heater leg and
from 18 to 24 mils in the second heater leg. Examination of Inconel loop
9377-6, which circulated LiF-BeFo-ThF), (71-16-13 mole %) for 13,155 hr at a
maximum temperature of 1300°F, revealed moderate general and intergranular
voids ranging from 1 to 8 mils deep in the first heater leg and from 3 to 13
mils in the second leg. The unheated bend connecting the two heater legs and
the cold-leg sections of each loop showed some pitting to a depth of 1 mil and
a relatively light concentration of subsurface voids up to 3 mils deep.

Of particular interest is the difference in depths of corrosion in the two
loops. The depth of void formation in loop 9377-6, which circulated a non-
uranium-bearing breeder salt with 13 mole % ThF), was approximately half that
in loop 9377-5, which circulated a salt containing a small percentage of UFh
and ThFh. Since the temperature and operating conditions for both loops were
approximately the same, the results support the prediction that salts contain-
ing thorium, even in relatively large amounts, should cause less corrosion
than salts which carry uranium. (This prediction is based on the finding that
long-term corrosion in fuel-bearing fluoride mixtures results from the reaction
2UF), + Cr <=2UFy + CrF,. A like reaction does not occur in the case of ThF),
which is reduciblé only to thorium metal.)

4.1.4 Molybdenum-Graphite Compatibility Tests

Graphite in the MSRE core is to be banded to prevent lateral distortion
during reactor operation.h A suitable banding material must meet the following
requirements:

1. withstand corrosion by the fluoride fuel,
2. retain its strength at reactor operating temperatures,

3. have approximately the same coefficient of expansion as the
graphlle,

i, not be carburized as a result of intimate contact with the
graphite,

5. be compatible with INOR-8 (e.g., no dissimilar-metal mass
transfer or serious interdiffusions).

Molybdenum appears to be highly suited to this application on the basis
of the first three requirements. To evaluate its properties with respect to
the last two requirements, two thermal-convection loops (No. 1250 and 1251)
have been placed in operation with the MSRE fuel composition LiF—BeFe-ZrF -
UF),-ThF), (70-23-5-1-1 mole %). Each of the loops is constructed of INOR-g and
contains a molybdenum insert in the hottest section in contact with graphite
and INOR-8. Details concerning the design and operation of these tests are
discussed in Sec. 2.11 of this report.

4.2 WELDING AND BRAZING STUDIES
4.2.1 Welding of INOR-8

During routine qualification of welders on INOR-8 material, the presence
of a possible weld-cracking problem was detected in one heat of the alloy.
82

Both bend tests and metallographic examination revealed the incidence of
cracking, and a cursory examination of the welding procedures indicated no
obvious remedy. Because of the importance of this problem, an investigation
was immediately started to determine its seriousness and to develop preventive
methods.

Parent plate and welding wire of the same types as those used when crack-
ing was observed were studied in the laboratory. Highly restrained welds were
prepared and the incidence of cracking was verified, indicating that minor
modifications in welding procedure would not materially improve the situation.
A photograph of a typical side-bend specimen cut from a weld in the l-in.-
thick plate is shown in Fig. 4.2a. A photomicrograph of a representative crack
in such a weld is shown in Fig. 4.2b. Further examination of these welds
revealed prevalent microporosity and occasional microfissures in the heat-
affected zone of the weld (in the base metal). Figure 4.3 illustrates the
undesirable fusion-line porosity and microfissures in this suspect heat of
base material, as well as the weld-metal cracks. As a result of this
preliminary evaluation, study was concentrated on both the weld metal and
the base metal.

In an effort to overcome the weld-metal cracking difficulties, an
experimental INOR-8 weld-metal composition containing 2 wt % Nb was
investigated. This filler metal was developed in the course of the INOR-8
welding program previously conducted by the Welding and Brazing Laboratory
and was considered exceptionally promising in view of its excellent elevated-
temperature mechanical properties.5 Welds made on the suspect heat of base
plate exhibited no evidence of weld-metal cracking, either in bend tests or
in metallographic sections. A typical side-bend specimen is illustrated in
Fig. 4.4a; a metallographic section is shown in Fig. L.4b. Wire from two
laboratory melts of this alloy was deposited with no evidence of defects in
the weld metal, indicating that use of this alloy will prevent weld-metal
cracking.

In addition to this work, approximately 175 metallographic specimens of
various INOR-8 welds made over the past few years were carefully re-examined
to determine whether cracks were present which had been overlooked during
routine metallography. A rapid means of screening was developed (involving
a sensitive fluorescent penetrant and ultraviolet light) which accurately
detected the presence of minute defects without the laborious process of
high-magnification examination of each sample. Any cracks that were found
in this screening could subsequently be examined with a conventional
microscope. As was expected and previously reported,6 most of the welds
made on the high-boron heat of INOR-8 (Haynes SP-16) exhibited many cracks.
However, welds made with approximately 10 different heats of weld wire (both
commercial and experimental) which where considered to possess good welda-
bility were resistant to cracking. A high degree of restraint was used in
making the welds (parent plates were welded to a 2-in.-thick backup plate),
and only one of these heats exhibited microfissures in the weld metal. Even
in this case, however, the incidence of microfissuring did not appear notably
more pronounced than that observed in similar welds of other nickel-base
alloys, such as Hastelloy-B, Hastelloy-W, and Inconel, which are usually
considered to have excellent weldability.

Studies to evaluate the microfissuring tendencies of INOR-8 parent
material were concentrated on the hot-ductility test developed and carried
out at Rensselaer Polytechnic Institute.’ The results of tests on Haynes
heats SP-16 and SP-19 have been reported previously. To date, it has been
demonstrated that the particular heat of parent material exhibiting the
83

UNCLASSIFIED
Y-39487

i
|

Fig. 4.2. (a) Side-Bend Specimen Showing Crack in Weld. (b) Photomicrograph of Representative Weld-
Metal Crack. Etchant: 3 parts HCl, 2 parts H,O, 1 part 10% chromic acid.
84

5 ' L R 1 e L8 . .
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Fig. 4.3. Photomicrograph of Welded Joint Showing Fusion-Line Porosity and Microfissures and Weld-
Metal Crack. Etchant: 3 parts HCI, 2 parts H,0, 1 part 10% chromic acid.

cracking difficulties in the welder qualification work and the subsequent
laboratory work possesses poor ductility at high temperatures. The
Rensselaer data for this heat of material and for a heat of material
possessing good hot ductility (Westinghouse heat M-1566) are plotted in
Figs. 4.5 and 4.6, respectively. It can be seen that samples of Westing-
house heat M-1566 exhibited 75% reduction in area when pulled at 2300°F,
while the suspect heat of material exhibited essentially 0% reduction in
area, even though both materials have approximately the same melting points.
Furthermore, the suspect heat of material appeared to be permanently damaged
after heating to 24OO0CF since the ductility remained nil upon cooling down
to at least 2000°F. Heat M-1566 material, on the other hand, provided a 48%
reduction in area after being cooled to and tested at 2000°F.

Although the results to date are inconclusive, it appears that the melt-
ing practice affects the hot ductility and weld-cracking tendency to a large
extent. The investigations are continuing in an effort to determine the
influence of material chemistry and melting practice upon the hot ductilities
and cracking or microfissuring susceptibilities of various heats of INOR-8.

4.2.2 Solidified-Metal Seal Development

Studies are continuing in an effort to develop suitable techniques for
producing solidified-metal seals for MSRP applications.
85

UNCLASSIFIED
Y-39488

1 f s A gza
N N ‘;’ J b H 7 ‘fuk',.’. ;"g]’ ) »
gl b Fass L i il 3 b v RN o SRy WE (] f9.47 e 174 ez - O—
ool WA Iy o O U S ' R LR E N i, L e
7 N0 A L Y / ¥ i =N A Y VT /‘,1,,..1.7‘!‘, -_,_‘j_.‘,//) L — —
AT e L T e L At [V 2 EVERF 5 A AL

Fig. 4.4. (a) Side-Bend Specimen Made with Niobium-Modified Weld Metal Showing Good Weld-Metal
Soundness. (b) Photomicrograph Showing Excellent Weld-Metal Quality. Etchant: 3 parts HCI, 2 parts H,0,
1 part 10% chromic acid.
REDUCTION IN AREA (%)

REDUCTION IN AREA (%)

86

UNCLASSIFIED
ORNL-LR-DWG 57705

80 80 0.4
60 J 60 ‘ 0.3
2l Y
T TESTED ON
| | % .
20 x 20 £ 0.
\ 2 k z $
0 & = 8 &
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80 2 80 _, 0.4
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40 ¥ 40 — 0.2 N o———8 P300°F
o———-0 2400°F
20 20 o\j}\ = 0.4 \
\‘OE . \\:-'OY
0 ' - 0 0
1800 2000 2200 2400 1800 2000 2200 2400 1800 2000 2200 2400
TEST TEMPERATURE (°F) TEST TEMPERATURE (°F) TEST TEMPERATURE (°F)
Fig. 4.5. Results of Mechanical-Property Tests on Suspect Heat of INOR-8.
UNCLASSIFIED
ORNL-LR-DWG 57706
100 100 0.5 ~N
80 I/,A\ 80 0.4
60 ( 60 0.3
I }TESTED ON
- - y 2 HEATING
o A\ T A
40 \ 7 40 \ 0.2 ~ &
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40 W 40 N 0.2 EE\T{ COOLING
e \ a, \ 2400°F
20 20 \I 0.4 ;
| | \
0 b 0 | g 0 | :
1800 2000 2200 2400 1800 2000 2200 2400 1800 2000 2200 2400
TEST TEMPERATURE (°F) TEST TEMPERATURE (°F) TEST TEMPERATURE (°F)

Fig. 4.6. Results of Mechanical-Property Tests on Westinghouse INOR-8 Heat M-1566.
87

4.2.3 Sump-Type Seal Development

The sump-type solidified-metal seal, which was described previously,8
failed to seal on the eleventh attempt after 10 successful prior sealings
(20 individual heating and cooling cycles). The base metal of this seal was
INOR-8, and the sealing alloy was 80 Au-20 Cu (wt %). Opening and closing the
seal was accomplished by direct induction heating. The required sealing
temperature with this alloy was about 1700CF.

Failure was caused by a fracture in the INOR-8 base material which allowed
the sealing alloy to leak out. The failure occurred through an intergranular
crack which probably resulted from general intergranular attack of the base
metal by the sealing alloy and from the severe thermal stresses set up by
direct induction heating. Figure 4.7a shows the nature of the intergranular
attack.

The matching member of this seal is shown in cross section in Fig. 4.7b.
Although some alloying can be seen, this part of the seal could probably be
subjected to many more sealing cycles without difficulty.

4.2.4 Metal-Fiber Seal Development

Studies are continuing with molybdenum-fiber compacts impregnated with
brazing alloy as the sealing medium. A previously constructed assembly8 was
tested as outlined in Table h.k.

Table 4.4. History of Seal Made Between a Gold-Copper Impregnated
Molybdenum-Fiber Compact and Two Molybdenum Plates

Sealing

Cycle No. Atmosphere Remarks
1-10 Helium Helium leaktight after each ccal
11 Helium Seal not leaktight; one molybdenum
plate appeared oxidized and it
was refaced
12-13 Helium Seals not leaktight
14 Hydrogen Plates were rewet in hydrogen and a
successful seal was made
L5 Argon Leaktight seal
16-20 Helium Seals not leaktight
21 Hydrogen Seal leaktight

It thus appears that impregnated metal-fiber compacts of this type can be
operated successfully for several sealing operations. However, slight oxidation
of the mating components will prevent adequate sealing, and these components must
then be cleaned of oxide or replaced.

A seal of this type was attempted on two INOR-8 plates. However, after
only one sealing, the alloying of base metal and gold-copper sealing alloy was
so great that the seal would not separate cleanly. Because of the requirement
to use INOR-8 for MSRP applications, two INOR-8 plates have been coated with
molybdenum by vapor deposition and are being prepared for use in another fiber-
seal experiment.
* %~ UNCLASSIFIED
j ' Y-39484
5 EXY l‘ 0.02
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ORIGINAL BRAZE METAL-

INOR—-8 INTERFACE

UNCLASSIFIED §
Y—-36739

Fig. 4.7. (a) Failure Area of Sump-Type Seal Experiment, Showing Intergranular Penetration of Base
Material by the Seal Alloy. Etchant: 3 parts HCl, 2 parts H,0, 1 part 10% chromic acid. (b) Matching
Component of Sump-Type Seal Experiment. Some alloying of base metal and seal alloy is evident. Etch-
ant: 3 parts HCI, 2 parts H,O, 1 part 10% chromic acid.
89
4.3 INOR-8 DEVELOPMENT

4.3.1 Structural Stability of 18% Mc Alloys

Past work was done to determine whether INOR-8 is embrittled by
precipitation of intermetallic compounds. No embrittling precipitates were
found in material which contained alloying constituents in concentrations
from about the minimum to the median of the range specified for INOR-8. 1In
studies of nickel-base alloys contsining 10, 15, and 20 wt % Mo; 5, 7, and
10 wt % Cr; 4 and 10 wt 4% Fe, there was evidence that intermetallic compounds
formed within the investigated temperature range of 1500 to 2000CF in the 20%
Mo alloys when they contained 10% Fe. Precipitate was not found in alloys
containing h% Fe, implying that the solubility of iron is between 4 and 10%
in nickel-base alloys which contain 15 to 20% Mo, 5 to 10% Cr, and 0.04 to

0.08% cC.

The specification for INOR-8 permits a maximum of 5% Fe and when other
constituents within the allowable ranges are also present, it remains to be
confirmed that no intermetallic¢ compound will occur. Also of interest are the
compositional limits of the phase boundary in question to establish what
deviations from specifications might be permitted.

Microstructures were studied of a series of 18% Mo alloys encompassing the
maximum compositional limits of INOR-8. Samples of these alloys were heat
treated for 100 and 1000 hr at temperatures ranging from 900 to 2000CF.

In general, metallographic examination of the specimens showed no inter-
metallic compound, but did reveal the presence of a fine-grain boundary
precipitate occurring in the temperature range of 1100 to 1650°F. This
precipitate, thought to be a carbide, went back into solution at 1800°9F. Two
alloys (VT-150 and VT-153) were observed to have precipitated particles, as yet
unidentified, that did not redissolve at 1800°F. Both the iron and chromium
content of these alloys was in the range of 8 to 10%, indicating that this may
be the region of composition where intermetallic compound formation begins.

Test specimens of additional alloys were prepared that bracketed the
composltions where the alleged intermetallic compounds were found in an attempt
to confirm the proposed phase boundary. The alloys were decarburized by a
hydrogen firing treatment prior to testing in an effort to eliminate any
precipitate caused by residual carbon. Aging tests for 100- and 1000-hr
duration have been started on these materials.

L, .4 MECHANICAL PROPERTIES OF INOR-8

The allowable design stresses for INOR-8 reported previously9 have been
revised as follows: (1) the minimum specified yield strength was increased
from 35,000 psi to 40,000 psi, (2) additional creep data enabled minimum creep .
rates to be established, (3) the criterion for establishing high-temperature
stresses was changed. Several factors led to these decisions. For instance,
previous values were based on a Larsen-Miller plot of the stress to produce 1%
creep strain. This criterion was replaced by four-fifths of the stress to
produce rupture in 100,000 hr, or two-thirds of the stress to produce a :
minimum creep rate of 0.1 CRU (10-5 %/hr). The fractions four-fifths and two-
thirds were used to avoid possible problems associated with the effect of
irradiation on rupture life or creep rate. A plot of the four criteria used to
establish maximum allowable stresses is shown in Fig. 4.8. The maximum allow-
able ctress is based on the eriterion which produces the lowest stress at each
temperature. Values at key temperatures are given in Tuble 4.5.
90

UGCLASSIFIED
. 4 - ORNL-LR-DWG 57707
4x10 -
T
S s S .
2 ' e e
Y&o—-:h\__._
b, TS A\ N
4, _—-14 2/3 YS - - ..‘-.' .
104 —— 4/5 RS |- \—"-
~ \
a \
- | ———1, TENSILE STRENGTH \
w 5 |—*—" 2/; YIELD STRENGTH
a Rk
= [ eeeneeee 4/5 RUPTURE STRENGTH (10% hr)
v — 2/3 0.4 CREEP RATE UNIT \ .
— MAXIMUM ALLOWABLE STRESS \
: N\
~—2/; 0.1 CRU —=
10° |

0 200 400 600 800 1000 1200 1400
TEMPERATURE (°F)

Fig. 4.8. Criteria for Establishing Static Design Stresses for INOR-8.

Table 4.5. Values of the Static Design Stress
for Wrought and Annealed INOR-8

Temperature Maximum Allowable
~ (°F) " Stress (psi)
0-200 24,000
300 22,800
400 21,700
500 20,800
600 20,000
700 19, 300
800 18, 700
900 18,150
1000 . 16,000
1050 13,250
1100 9, 600
1150 6,800
1200 ' "~ 4,950
1250 3,600
1300 2,750
1350 - 2,050
1400 1,600

Tensile tests on cast INOR-8 have been performed by the Haynes Stellite
Company and by the Mechanical Properties Group at ORNL. These tests. were
performed. on 0.25- and 0.505-in.-diam bars of three heats of cast metal.
Tensile data are shown in Figs. 4.9 and 4.10. Creep and stress-rupture tests
will also be performed on cast metal within the next six months.
91

UNCLASSIFIED
ORNL-LR-DWG 57708

0
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: ‘TW%%

TENSILE STRENGTH—T"

80

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>4

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= 50 A

x %

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T s YIELD STRENGTH
/ 427’;§/// /%éggfif/; 7
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4

30

20
SYMBOLS DENOTE. VARIOUS HEATS

0 200 400 600 800 1000 1200 1400

TEMPERATURE (°F)

Fig. 4.9. Tensile and Yield Strength vs Temperature for Cast INOR-8.

A program is in progress to determine the effect of notches on the stress-
rupture properties of INOR-8. Rod specimens of heat SP-19 are being tested at
1100 and 1300°F in air for times up to several thousand hours. Results of the
_ first few tests, shown in Fig. 4,11, indicate that specimens with a notch radius

of 0.005 in. are not significantly weaker than smooth specimens.

4.5 IMPREGNATION OF GRAPHITE BY MOLTEN SALTS

The use of unclad graphite in a molten-salt reactor requires that not more
than a small percentage of the graphite pore space be filled with salt. To
ensure this condition, a test program has been in progress to study the
impregnation of various potential grades of graphite by LiF-BeF,-ThF),-UF), (67~
18.5-14-0.5 mole %) at 1300°F (704OC) for various times and pressures.

A radiographic techniqpe* has been used to determine fhe location and con-

figuration of the salt impregnated into specimens of nine grades of low-
permeability graphite. All had less than 1% of their bulk volumes impregnated

*The technique is an adaptation of that used by C. E. Grieder at the
National Carbon Research Laboratories in his studies on the location and
configuration of pores in graphite by impregnating them with mercury prior
to the radiographic cxamination.
STRESS (psi}

80

60

40

REDUCTION IN AREA (%)

20

60

ELONGATION (%)

20

92

UNCLASSIHFIED
ORNL-LR-DWG 57709

SYMBOLS DENOTE VARIOUS HEATS

| /W%fl/

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% % % ////%///.//////V///}

i

A <&

| E T
SYMBOLS DENOQTE VARIOUS HEATS

40-

7 f

/4.9

Q\Y

e v//Q
&

o} 200 - 400 600 800: 1000 1200 1460

TEMPERATURE™ (°F)

Fig. 4.10. Elongation and Reduction in Area vs Temperature for Cast INOR-8.

UNCLASSIFIED
ORNL-LR-DOWG 57710

S
10 . I I
T ]
— . O SMOOTH
. - [ — : ® NOTCHED
5 ) e o}
[
- s 1100°F
[
P ——— .
7 .-~~ ' \\-—-__
» —— | ""‘\~l__~
I~ 1300°F
—
9 \~~~\..
104 -
10 2 5 10° 2 5 10> 2 5 10°

RUPTURE LIFE (hr)

~ Fig. 4.11. Stress-Rupture Data for Notched and Smooth Rod Specimens of INOR-8.
93

by the salt in the standard permeation screening test. The resultslO of these
screening tests are summarized in the tabulation on Fig. 4.12.*%

Transverse and longitudinal thin-section samples from salt-impregnated
specimens of each of the grades of graphite were taken as shown in Fig. 4.12.
Radiographs of the samples are shown in Fig. 4.13. The white phase is the salt,
and the gray is the graphite. These radiographs are indicative of what has
occurred in each graphite specimen, but examination of more specimens would be
necessary to determine whether these salt distributions are typical. The AGOT
graphite was included to illustrate the differences between it and the low-
permeability grades of graphite.

The skin-like permeation of some grades of graphite suggests that machining
tends to open up some of the closed pores of the graphite.

In general, the pipe specimens were impregnated through the cut ends rather
than transversely through the walls. Presumedly, little or no salt would
impregnate the pipe if the ends were suitably sealed by some method such as
brazing. Tests are planned to determine whether this is true.

The B-1 graphite, which had the smallest amount of impregnating salt, had
the deep longitudinal crack shown at the left in its radiograph prior to the
impregnation test. The small crack above it was not detected until after the
radiograph was taken; so it is not possible to determine whether salt impregna-
tion caused it. However, the over-all appearance of the radiograph suggests
that the small crack was in the as-received graphite. The longitudinal thin
section for the B-1l graphite was broken during its preparation, apparently due
to the deep longitudinal crack.

These radiographic examinations seem to show that graphite can be obtained
that will not take up molten salt except for shallow (<- 50-mil) surface
impregnations under the conditions of the tests. The 150-psig salt pressure
used 1in Lhe tests is approximately three times the maximum expected in the
MSRE.

4.5.1 Permeation Screening Tests on Large and Small Pipe

Specimens from the wall of a large graphite pipe, 3-11/16-in. OD, 2-5/8-in.
ID, 6-in. length, were submerged for 100 hr in LiF-BeF,ThF),-UF), (67-18.5-14-0.5
mole %) at 1300°F under 150 psig--the standard permeation screening test. This
graphite, which was identified as a CEY typc, had 5.5% of its bulk volume per-
meated by the salt, which is approximately five times the maximum found for the
standard size of 1-1/l.in. OD and 7/8-in. ID for CEY graphite. Additional
samples of the large pipe are being sought for testing to determine whether the
increase in porosity resulted simply from the larger fabrication size.

Additional standard salt permeation tests were made on 1-1/4-in.-OD, 7/8-in.-
ID CEY-grade pipe. The specimens were taken from the same lot of graphite as
were those for previous testslO but from a different pipe. The bulk volume of
the graphite permested by LiF-BeF,-ThF)-UF) (67-18.5-14-0.5 mole %) was 0.3%,
compared with 1.0% on the specimens in the earlier tests.lO0 The CEY specimens
in the earlier tests had some microscopically visible cracks; those in the
current tests did not, which probably explains the permeation difference.

**The test-specimen sizes and shapes were not standardized because of
the sizes and shapes of the original pieces of graphite plus the demands of
other experiments.
| ondCAIES, . | UNCLASSIFIED
| ORNL—LR—DWG 56248

PERMEATION OF VARIOUS GRADES OF GRAPHITE
BY LiF-BeF ,-ThF ,-UF, (67-18.5-14-0.5 MOLE %) SALT

TEST CONDITIONS: TEMPERATURE:  1300°F {704°C)
PRESSURE: 150 psig
EXPOSURE PERIOD: 100 hr

BULK VOLUME SPECIMEN DIMENSIONS

- . GRAPHITE ‘OF GRAPHITE  SPECIMENS Ay
'Gifl‘&PH'TE APPARENT DENSITY  PERMEATED WITH.  PREPARED D]MET(EN:M'NAL) inches
RADE . glect SALT . FROM oROD 1D " LENGTH(S)
”n
B s 00, . ROD 050 - 0.50. A~/ .
: e @ 00025 in. A
S-4LB - 1.90 0.24 BAR 0.50 0.75 THICK
GT-123.82 1L9Ne) 0.3 *ROD - 0.50 1.50 g
cs-12s. . 1894 0.5¢) ROD 0.90 0.50 \
RH1 T ONE .06 ROD 0.90 0.50 ’
. CEY-1350 1.90. 0.7 ° . PIPE 1.24 0.8  0.25 '
" CEY-G 1.88 0.9 PIPE 1.25 0.8  0.50 . N
s4LA . - 1894 1.04) BAR 0.50 0.75 THIN SECTION
CEY . 1.916) 1.06) PIPE 124 ¢85  0.50-0.25 SAMPLES
. AGOT 168200 12.9(20) BAR 0.50 1.50 MOUNTED FOR

*ALL VALUES ARE AVERAGES OF THREE WITH THE EXCEPTION OF THOSE WITH SUPERSCRIPTS OF ‘ WITH SALT SECTION SAMPLES RADIOGRAPHY

NUMBERS IN PARENTHESES; THE NUMBERS IN PARENTHESES ARE THE NUMBER OF VALUES AVERAGED.
| | Cylindrical Specimen of Graphite.

UNCLASSIFIED
ORNL-LR=DWG 56249

- - - ~ THIN SECTION

IMPREGNATED ORIENTATION OF THIN SECTION SAMPLES S SAMPLES
WITH SALT | - MOUNTED FOR .
' RADIOGRAPHY

Graphite Pipe Specimen.

Fig. 4.12. Location and Orientation of Transverse and Longitudinal Thin Sections Taken from Salt-Impregnated Graphite Specimens for Radiographic
Determination of Salt Distribution in the Graphite Specimens. ' -

v6
UNCLASSIFIED
PHOTO 52997

Gli—425—82 CS—112—S8 RH—1

CEY—1350 CEY-G S—4—LA

AGOT

Fig. 4.13. Salt Distribution in Various Grades of Graphite Following the Test Outlined in the Tabulation on Fig. 4.12, as Shown by Full-Scale Radio-
graphs of Their Thin Sections.

g6
96
4.5.2 Oxidation and Permeation

The brazing of graphite-to-graphite and graphite-to-metal joints in air
would be expected to increase the pore sizes of the graphite because of its low
oxidation resistance. Since the brazing cycle is very short, however, it was
thought that only a slight effect might be produced, despite the relatively high
temperatures used.

To determine the degree of change, the Welding and Brazing Laboratory
supplied control and "oxidized" specimens of CEY graphite for the standard salt
permeation test. The "oxidized" specimens had been exposed to air for 20 sec
at ~ 23720F. The percentages of the bulk volume of the graphite permeated by
salt were 0.3 and 0.5, respectively, for the control and "oxidized" specimens.
Therefore, it appears that brazing graphite in air results in some increase in
its porosity.

In the preheating of a molten-salt reactor preparatory to filling it with
salt, the temperatures will be lower (932 to 12920F) but the time will be long.
The above results indicate that in the preheating care should be taken to
exclude or minimize oxygen or oxidation catalysts,ll particularly at tempera-
tures in excess of 7520F. Otherwise, the graphite pores may be enlarged and
the potential for salt permeation increased.

L.6 PRECIPITATION OF UO, FROM FLUORIDE SALTS

The studies by Cookl2 which revealed the problem of UO, precipitation from
fused salts were conducted initially on fuels which contained neither zirconium
nor thorium fluoride. It was suggested that zirconium and thorium would act as
inhibitors to the UO, formation, and tests were conducted on several salts con-
taining these metals. It was readily shown that 5 mole % ZrFu eliminated any
precipitation of uranium. In the past six months tests with LiF-BeFe-ThFh-UFh
(65-30.5-4-0.5 mole %) were conducted both in contact with graphite and in a
container with no graphite present. Radiographic examination revealed precipi-
tation within 20 hr, and the tests were concluded after 100 hr. As expected,
the greatest amount of precipitation occurred when graphite was present. The
precipitates were analyzed by x-ray diffraction and identified as UO,. Thus it
may be concluded that 4 mole % ThF), will not inhibit uranium precipitation.
However, several tests were conducted for more than 3000 hr with salts contain-
ing 14 mole % ThFh, and no precipitation of any type was detected. Thus thorium
is an effective inhibitor of U02 formation when present in the salt at some
concentrations greater than 4 mole %.

4.7 PRECIPITATION OF ZrO, FROM MSRE FUEL

2
Standard tests of the compatibility of graphite and fuel were made with
the MSRE fuel mixture LiF-BeF,-ZrF) -UF) (70-23-5-1-1 mole %). The salt mixture
was exposed for 500 hr to AGO% graphite at 1300°F. A similar charge of salt
was used as a control in a capsule containing no graphite. The tests were
monitored by radiography at 100 and 500 hr. Precipitates were noted in both
capsules at the end of 100 hr, with approximately five times as much being
present in the test capsule containing graphite. No additional precipitation
was noted after 500 hr. X-ray analyses indicate that the precipitates are
entirely monoclinic ZrO,. No evidence of BeO, Th02, or U0, was found. It
appears that the usual oxygen contamination was present in the graphite and
that ZrO, was formed in preference to other possible oxides. The slight amount
97

of precipitate in the control capsule probably resulted from chemisorbed oxygen
on the inner surface of the metal container. It thus appears that zirconium is
behaving as expected in preventing the formation of UO,.

4.8 REMOVAL OF OXYGEN CONTAMINATION FROM GRAPHITE
4.8.1 Purging with Molten Fluorides

Work hag continued on the evaluation of various molten fluoride salts as
purging agents to remove the normal oxygen contamination from graphite. As
reported previouslyl3 and discussed above, uranium or zirconium, depending on
the composition of the salts, will precipitate as oxides from the molten salts
1n Lhe presence of oxygen. A lerge snurce of oxygen in the MSRE is the graphite
moderator.

One technique suggested to remove most of the oxygen from the graphite is
to purge the system with a fused salt. The usual method is to purge a standard
erucible of AGOT graphite (the bulk volume of the graphite is approximately 190
cc and the volume ratio of graphite to molten salt is 27:1 at 1300°F) with the
purging salt of interest at 1300CF in a vacuum for various times. To evaluate
the effectiveness of this treatment, the purging salt is replaced with LiF-BeFe-
UF), (62-37-1 mole %), and the test conditions are repeated with exposure time
as a variable. If oxygen is present, a U02 precipitate readily forms which can
be detected by radiography.

Neither LiF-BeF,-UF), (62-37-1 mole %) nor LiF-BeF,-ThF) -UF), (67-18.5-14-0.5
mole %) have proved effective as purging salts for 20- and 100-hr purging
periods.

To date, NaF-ZrF)-UF), (50-46-4 mole %) has been the most effective purging
salt Lrled. The teot conditions and results are summarized in Table 4.6.
These precursory tests indicate that the effectiveness of purging with the
NaF-ZrF -UF), may be time dependent and may be associated with the rate of
oxygen fiiffusion through the graphite. Purging tests of graphite with the
NaF-ZrF),-UF) are in progress and will continue for 500 hr to further evaluate
this possibility.

Table 4.6. Summary of Oxygen Purging of Standard Crucibles of AGOT
Graphite with NaF-ZrF)-UF), (50-46-4 mole %) at 1300°F in a Vacuum

UO» Precipitation Detected by
Radiography in Purged Graphite Crucibles

Purging Time at Intervals of Their Exposure to LiF-
with NaF=7rF) -UF), BeF,~UF), (62-37-1 mole %) at 1300°F (704°C)
Test 50-46-4 mole %, '
No. (hr) 100 hr 500 hr 1000 hr
Al1813 20 No Yes Yes
-A1865 20 No Yes Yes
A1809 100 No No . Yes

A1859 - 100 - No Yes -

98

: As a result of the tests that have been conducted, it appears that some
benefit can be derived from purging of graphite with molten salts. Moreover,
it should be emphasized that the metal surfaces as well as the graphite must be
purged after construction of the reactor is completed, and a barren-salt purge
would serve to check the operation of the system as well as to clean the metal
surfaces and purge the graphite.

4.8.2 Purging with Ammonium Bifluoride

It has been reported13 that in less than 20 hr the thermal-decomposition
products of NH,F.HF crystals in the presence of graphite apparently removed the
normal oxygen contamination in the graphite to such an extent that no U0,
precipitation occurs when LiF-BeFo-UF), (62-37-1 mole %) contacts the graphite
at 1300°F. : -

~ An AGOT graphite crucible that received the NHhF-HF treatment was
monitored radiographically at periodic intervals during exposure to LiF-BeFe-
UF), (62-37-1 mole %) at 13000F. After 2000 hr of exposure, there was no
precipitate detectable in the LiF-BeFe-UFu. The test was prematurely terminated
- after 3000 hr because a small leak developed in the Inconel protective container.
The leak permitted oxygen .to enter the crucible and a precipitate was formed.

In the pretreatment of graphite with the thermal-decomposition products
of NH,F.HF erystals cited above, the crystals wére placed inside the graphite
crucible and decomposed. To determine whether the intimate contact aided the
pretreatment, a graphite crucible was given the same NH)F:HF. pretreatment-
except that the‘NHhF-HF crystals were not in contact with the graphite. The
thermal-decomposition products had to travel a short distance to reach. the
graphite. The treated graphite crucible contained LiF-BeF,UF), (62-37-1 mole %)
for 500 hr at 1300°F, and radiographic examinations did not detect any precipi-
tation. A leak developed in .the protective container during the 500- to 1000=hr
test interval, and the test was terminated. The test indicates:that.it might
be possible. to pump the-decomposition products of.NH,F-HF to heated graphite ..
and. remove: oxygen: - More- elaborate- tests will be necessary to verify this .
possibility under conditions more like those that would be encountered in a
molten-salt reactor. .

%.8.3 Reaction of INOR-8 with the Thermal-Decompoéition Products of NH),F - HF

The pretreatment of graphite as described in the methods above would also
require that the INOR-8 structural material of the MSRE be similarly exposed to
the thermal-decomposition products of the NHL,F-HF crystals. Precursory tests
have indicated that there is a slight reaction. Three INOR-8 tab, nominally
0.040 by 1/4 by 3/4 in., were suspended in & nickel container. The test con-
ditions were the same as those used for the pretreatment of the graphite
crucibles except that there was no graphite present, a nickel container was
used rather than Inconel, the weight of the NH}F.HF crystals was doubled, and
the NH),F.-HF thermal-decomposition products were retained in the test for 100 hr.
These four exceptions and the use of small tabs of INOR-8 were all in an effort
to. concentrate the reaction on the INOR-8. Two tests of this kind were made.
Metallographic examination of specimens from the tests indicated an approxi-
mately 0.0005-in.-thick nonporous reaction layer on the INOR-8. A similar
layer was found on the nickel support wire. X-ray analyses of samples from
the layer on ‘the INOR-8 indicated two or more complex structures which could
not be identified by the current ASTM data cards.
99

A series of tests is in progress in which both graphite and INOR-8
specimens are exposed for 20 hr to the thermal-decomposition products of
NH) F.HF at 13009F and lower temperatures. The tests are to determine
(1) whether the removal of oxygen from graphite can be done effectively at
lover temperatures and, if so, the change in the extent of the reaction with
INOR-8 and (2) the effects, if any, of the reaction layers on the fluoride
corrosion and tensile strengths of INOR-8 at room temperature and 1250°F.

REFERENCES

1. MOR Quor. Prng. Rep. Oct. 31, 1957, ORNL-2L431, p 23.

2. MSR Quar. Prog. Rep. Jan. 31, 1959, ORNL-2684, p 59.

3. MSR Quar. Prog. Rep. Apr. 30, 1960, ORNL-2973, p 17.

4. B. W. Kinyon, Effects of Graphite Shrinkage in MSRE Core, ORNL CF-60-9-10
(sept. 2, 1960).

5. MSR Quar. Prog. Rep. July 31, 1959, ORNL-2799, p Tl-T2.

6. MSR Quar. Prog. Rep. Oct. 31, 1957, ORNL-2431 p 18-28.

7. E. F. Nippes et al., "An Investigation of the Hot Ductility of High-
Temperature Alloys," Welding J. 34 (4), 183-85 (1955).

8. MSR Quar. Prog. Rep. July 31, 1960, ORNL-3014, p 58-61.

9. MER Quar. Prng. Rep. July 31, 1960, ORNL-301k, p 66.

10. MSR Qu;r. Prog. Rep. Apr. 30, 1960, ORNL-2973, p 5k4.

11. National Carbon Company, The Industrial Graphite Engineering Handbook,
p 5D.0e.01-5D.03.03, New York, 1959.

12. W. H. Cook and D. H. Janse, A Preliminary Summary of Studies of INOR-8,
Inconel, Graphite and Fluoride Systems for the MSRP for the Period
May 1, 1958 to Dec. 31, 1958, ORNL CF-59-1-4, p 1-20.

13. MSR Quar. Prog. Rep. Apr. 30, 1960, ORNL-2973, p 59.

5. IN-PILE TESTS

5.1 GRAPHITE-FUEL CAPSULE EXPERIMENTS

The first two MSRE graphite-fuel capsule experiments, each containing four
capsules (ORNL-MIR-47-1 and -2), were examined in hot cells at the Battelle
Memorial Institute. 1In these capsules, delicate bellows were used for encapsu-
lating graphite immersed in fused-salt fuel. It was intended that the permeation
of graphite by fuel and the effects of fissioning at the surface and in the pores
be determined. The bellows provided a flexible wall through which a hydraulic
pressure of 100 psia was transmitted to the fuel from a sodium bath in which the
capsules were submerged. The flexible bellows wall was also intended to provide

' for the expansion of the molten-salt fuel on melting. Because of the high heat

rate in the fuel, 200 w/cc, and the associated requirement for heat dissipation
with the reactor at power and because of the space limitations of the experiment,
it was impossible to maintain the salt molten during a reactor scram. Of the
eight capsules examined, only one was found intact; the bellows walls had rup-
tured in the other seven. Areas in which the ruptures had occurred appeared to
be extremely brittle; in some cases portions of the bellows wall had broken com-
pletely away, exposing the solidified salt underneath. The bellows were of
0.005-in.-thick Inconel and INOR-8 in experiments 1 and 2, respectively. Based
‘on the condition of the ruptured capsules, which were grossly distorted, and the
location of cracks found in the inner surface of the bellows walls of the one
good capsule, it is believed that the expansion of the molten salt fuel on
melting caused the ruptures. '

The one capsule having good external appearance was examined extensively.
Dimensional changes were of the order expected, a 0.002-in. increase in the 2.5-
in. length and a 0.0002-in. increase in the 1-in. diameter. The weight gain of
the graphite was approximately 1%, which corresponds favorably with out-of-pile
- tests for the same type of graphite and indicates that radiation did not increase
fuel penetration. Metallographic examination of the bellows revealed small
cracks radiating from the inner surface of the bellows at the base of the end
convolutions. These cracks are attributed to the strain imposed on the bellows
during melting of the salt. Although in one instance a crack penetrated the
bellows wall, there was no evidence of leakage based on salt analyses. Any
intermixing of sodium in the fuel which may have occurred is believed to have
been so slight that it did not affect the graphite. The radiochemical analysis
of the graphite is covered in Sec. 5.2.

The bellows capsule had been accepted as a reliable means for pressure
transmission, but with considerable misgiving concerning its structural integrity.
However, out-of-pile tests appeared to justify the design, although it was not
possible to duplicate the melting process of the in-pile experiment. Twelve out-
of-pile tests were made in which prototype capsules were heated to operating '

" temperature. The in-pile test capsules also were subjected to this same test
prior to assembly. Although a few prototype failures occurred, all were attribut-
able to causes other than expansion of the melting salt. The out-of-pile cap-
sules were heated in a furnace, causing melting of the salt from the outside
inward at a slow rate. In the in-pile -case a rapid melting occurred during a
fast return to reactor power. Since the heat source was in the 'salt fuel and

100
101

since the bellows wall was submerged in relatively cold sodium, the salt melted
from the inside while the bellows convolutions were held rigid by the still-
frozen salt within. By this mechanism it is postulated that expansion of the
melting salt on the inside strained the bellows until rupture occurred. The
ruptured bellows showed considerably more damage than could be attributed to
attack by the salt or sodium alone.

Although the original cause of fallure seems clear, the sequence of events
which followed the failure and led to the heavy damage to the bellows is less
easy to postulate, Metallographic examinations unfortunately shed no light on
the problem. Results of chemical analyses now in progress may provide some
clues,

A new capsule design has been made for the third experiment, based on some-
vhat different criteria. One objective of this experimenl, which alco containsg
four capsules, is to study samples of graphite that have been irradiated under
conditions more extreme than those expected in the MSRE reactor. For this pur-
pose two of the four graphite samplee will be preimpregnated with a modified
MSRE fuel. The graphite will be impregnated by submersion in fuel under vacuum
followed by pressurization to drive the fuel into the graphite. The temperature
within the impregnsted graphite will be 1900°F, and the fuel is expected to reach
2000°F. Respective temperatures in the MSRE reactor will be limited to 1275°F.
The graphite to be ueed in the impregnated case will be type AGOT (National
Carbon Company graphite, reactor grade), which is more porous than the MSRE-
grade graphite and consequently can be impregnated with more fuel.

A second objective is to detect changes in the wetting characteristics of
the fuel in contact with graphite under high-temperature irradiation. This will
be done by comparison of the postirradiation frozen-salt miniscus and permeation
of salt into the graphite for irradiated and nonirradiated graphite samples sub-
jected otherwise to as near as possible the same salt fuel and time at tempera-
ture. For this purpose two capsules will contain unimpregnated graphite type
RO025 (National Carbon Company graphite, experimental grade), similar to that of
the MSRE, and will be operated at graphite and fuel temperatures of 1800 and
1900°F, respectively.

A third objective involves the inclusion of specimens of molybdenum, INOR-8,
and pyrolytic graphite in contact with both the graphite and the fuel to demon-
strate the compatibility of the principal components of the MSRE system.

The four capsules consist of heavy-walled INOR-8 cylinders, each containing
a 1-1/4-1in.-diam 3-in.-long graphite cylinder, The upper half of each graphite
cylinder is machined in the form of a boat-like cavity to contain 5 cc of molten
salt fuel, in which the pyrolytic graphite and the metal specimens will be sus-
pended.

The remainder of the experiment assembly is similar to that of the previous
experiments. The capsules are suspended in a tank of sodiwn vhich serves as a
heat transfer medium, The sodium tank is a tapered cylinder with a small variable

experiment water jacket. The gas gap is varied by moving the sodium tank axially
relative to the water jacket through an externally driven screw. The heat gener-
ated within the capsules by fissioning and within the sodium tank by gamma ab-
sorption is dissipated to the experiment water jacket through the gas gap, and

the experiment temperature is controlled by the adjustment of this gap. The

total fission heat for the four capsules is 6000 w, including that of the fuel

in the impregnaled graphite. The 1ntal heat from all sources, including the gamma
heat in the sodium tank will be 18,000 w. '
102

This experiment 1s scheduled to be irradiated in the MIR during the perlod
May 1 through July 24, 1961.

5.2 EXAMINATION OF GRAPHITE SPECIMEN FROM EXPERIMENT ORNL-47T-2

A graphite specimen from ORNL experiment 47-2 was examined in the Battelle
Memorial Institute hot cells. This specimen (type SUA graphite) had been irradi-
ated in the MIR at ~1300°F for 1600 hr, at reactor power, in contact with LiF-
BeF o>-ThF4-UFq - (67-18.5-14-0.5 mole %), as a part of the experiment described in
Sec. 5.1. Optical examination of the specimen up to 34X magnification revealed
that the surface generally appeared to be unaffected by the irradiation. There
was evidence; however, of small amounts of fuel salt adhering to the bottom
surface of the specimen.

The specimen was sectioned as diagrammed in Fig. 5.1, and an autoradio-
graphic study was made of each of the cut faces. The high radiation levels of
the sections, which had contact readings in the range of 50 to 100 r/hr, re-
quired the development of special autoradiographic techniques.

UNCLASSIFIED
ORNL—LR—DWG 55948A

SURFACE TO BE EXAMINED

N\

'/sin.A-‘ ,-— 5 in. A\-r—i 3, in.—————=]
| f |
o CORE—,DRILL 47-2-3
S—_— | | sAamPLE |
-~ : |
_____ I | METALLOGRAPHIC - /SEREQ?A%MT\;OED
™ | | SPECIMEN 2 /
/ | . B ———cut 2
| _ | | 47-2-3 " o
. |
- | | IDENTITY :
CHEMICAL—ANALYSIS SAMPLE/I/I NOTCH\"Q METALLOGRAPHIC 1
N .
[ |
l CuT 3 CUT 1
. CuT 4
- : 2.50 in. -
i A-36334

- Fig. 5.1. Diogrom for Sectioning of Graphite Specimen.S4A, Capsule 3, Experiment 47-2.

Autoradiographs of the diametrlcal faces are shown in Fig. 5. 2 arranged
looking down from above. All evidence, from both the autoradlographs and the
core-drilling tests to be described later, indicates that the interior regions
of high activity were not merely on the cut diametral surfaces, but extended
longitudinally through the graphite. It is postulated that the graphite speci-
men contained planes of high porosity (visible in cross sections as chords)
which had channels to the circumference and possibly to the bottom, into which
molten salt fuel intruded.  Similar results were noted in the radiotracer study
of the diffusion of Eu*®2’?5% jinto AGOT graphite.l Core-drilling apparatus,
103
consisting of a Dumore high-speed bench drill and modified No. 16 hypodermic-
needle drills, had been previously developed for the Fu'®2’15% diffusion experi-
ments.® This apparatus was used to obtain samples of graphite in specific areas

selected from observation of the autoradiographs. Because of the high level of
radioactivity it was necessary to conduct the drilling remotely in a hot cell.

UNCLASSIFIED
ORNL—LR—DWG 55949

a. Top 1/8-In. Section

RC27

b. Bottom 1/8-In. Section

é—in. saw kerf —
RC28

c. Top 1/2-In. Section
(Drill Face)

RC31

d. Bottom 1/2-In. Section

RC29

|
|
|
é -in. saw kerf —-—:
|
|
i

e. Top Metallographic
Specimens 1 and 2

RC48, RC43

f. Bottom Metallographic
Specimens 1 and 2

RC49, RC44

A- 36335

Fig. 5.2. Schematic Diagram and Autoradiographs of Diametral Faces of Metallographic Specimens 1
and 2 from Graphite Specimen S4A, Capsule 3, Experiment 47-2.
104

Samples were taken along the high-activity chord and along a number of radii
for the measurement of activity as a function of distance from the circumference.
An enlarged autoradiograph of the drill face is shown in Fig. 5.3, and a sche-
matic sketch of drill-hole locations is shown in Fig. 5.4. Upon completion of
drilling, the specimen was again autoradiographed and photographed (Fig. 5.5),
and a definite reduction in total radiocactivity was noted, especially along the
chord.

The cores were removed from the drill needles and weighed, and their rela-
tive gross gamma activities were determined. These data are shown in Table 5.1.
There was a great variation in unit activities among these cores, which, however,
correlated quite well with the location of the hole and activity of the region.
With the radial cores, activity increased rapidly as the periphery of the
graphite was approached. Very high levels of activity were attained where the
drill hole broke out the side; for example, core 5 had a count of 800,000 cpm/mg.
Core 4, with a thin wall between it and the periphery, had 62,500 cpm/mg; core 3,
with a slightly thicker wall, had 21,000 cpm/mg; and core 1, with a wall thick-
ness of ~0.T mil, had TT4O cpm/mg. Interior cores, when located in areas shown
by the autoradiograph to be relatively free of activity, were characteristically
low, 2000 to 6000 cpm/mg.

UNCL ASSIFIED
PHOTO 52893

~4X \/ RC31
Notch

Fig. 5.3. Enlarged Autoradiograph of Upper Drill Face of ]/2-in. Section from Graphite Specimen S4A,
Capsule 3, Experiment 47-2, Before Drilling.
105

To investigate the extent to which the activity extended below the drilled
surface, one core (core 18) was broken in half, and the two sections were weighed
and counted separately. Relative activities were nearly the same for both the
top and bottom sections. Core 14 was also broken in half, and the sections were
counted. Both pieces of this core contained considerable activity.

Cores 1, 8, 9, 11, 12, 28, 32, 33, 34, 35, and 36 were analyzed with a
multichannel §amma spectrometer. Peripheral cores showed principal peaks indi-
cating Cel4ls144 pgl33 gy103,106. ang 7r9S, Nb°S, similar to those exhibited
by the fused-salt fuel. Cores from the high-activity chord showed these same
peaks, suggesting that the activity resulted from the intrusion of molten-salt
fuel.

Interior cores in sound graphite areas showed only one major gamma activity,
that of cesium. Based on analyses of wlhiolc cores by gamma spectrometry, approxi-
mately two-thirds was Cs1®% and about one-third was Cs'>7., Table 2.2 contains
/

the results of gamma analyses of several cores, together with Cs137/cs134 ratios.

UNCLASSIFIED
ORNL—LR—DWG 55950A

NOTCH

Fig. 5.4. Scliematic Diagram of Core-Drilling Locations in l/1,-in. Graphite Specimen S4A, Capsule
3, Experiment 47-2.
106

UNCLASSIFIED
PHOTO 52894

ax V HC6092
Notch

Fig. 5.5. Autoradiograph and Photograph of l/2-in. Section from
Graphite Specimen S4A, Capsule 3, Experiment 47-2, After Core
Drilling.
107

(a)

Table 5.1. Relative Gross Gamma Activities of Cores

Diameters High-Activity Chord
Core Weight Activity . Core Weight Activity
(mg) (cpm/mg) (mg) (cpm/mg)
1 5.3 _ 7, T40 19 9.3 47,260
20 6.7 2,610 18 16.5 370,800(°)
21 11.5 2,630 17 6.2 87,900
4 6 4.9 92,700
3 4.5 21,100 8 9.0 55,100
2L 5.k 4,570 9 6.7 37,500 .
25 3L - 9,000 10 10.9 13,600
11 8.1 2,200
5 - 9,8 ~800,000 12 11.2 1,600
27 9.3 21,700 .13 9.9 76,700
26 10.k 6,690 1l 6.2 - 13h,800(%)
15 3.0 42,900
L 10.0 62,500 16 8.4 83,700
23 7.2 ' 4,550
22 9.7 4,570
Reats (®)
28 5.6 15,300 Cs*37 standard,|] = 5180 cpm
29 8.0 5,700 [;.18 pe ] - 4380 ecpm/uc
- 30 12.3 3,720
31 3.1 2,430
32 9.8 55,700
33 10.8 4,910
34 9.2 3,620
.35 8.5 - 3,090
36 6.9 200, 000
37T 7.2 4,100
38 6.7 4,040
39 11.4 3,210

(a) . Relative gamma activities based on measurement 10 in. from 2 in. x 2 in.
crystal (179 mg/em® aluminum abhsorher). High level cores at 20 in. cor-
rected to 10 in, distance.

(p) Radial holes drifted toward circumference going from 28 to 32 to 36.

(c) Core broken in half on removal. Top half: 9.3 mg, 191,600 cpm/mg.
Bottom half: 7.2 mg, 179,200 cpm/mg. -

(d) Core broken in half and counted. 14A: 2.8 mg, 64,340 cpm/mg.
14B: 2.6 mg, 161,820 cpm/mg.
108

Table 5.2. Concentrations of Cesium—l3h and -137 in Graphite Cores

Core No. ' Salt
29 33 37 38 34 39 35 Fuel
Weight of core, mg . 7.0 10.5 5.4 5.7 8.9 1.0.4 7.2 -
Distance from circum-
ference, mm 2.25 2.5 2.5 h,25 4,5 5.5 6.5 -
cs37, 10° dpm/mg 1.44%  1,86% 2.46% 4,32 3.76% 3.38 3.56 236
cs13%4, 105 dpm/mg 10.3 9.55 9.35 8.120 5.72 7.08 6.68 7.26

cs*37/cs?3% ratio 0.139 0.195 0.263 0.533 0.658 0.478 0.533 32.5

* .
Presumably low due to interference from the Zr®S, Nb®5 protopeak.

The cesium ratio determined from analysis of the molten-salt fuel is also listed
for comparison. The marked difference between Cs13% and Cs'®7 distribution be-
tween the fuel and graphite indicates that the cesium diffusion is not primarily
as cesium. Also, the diffusion rates do not appear to be in a ratio consistent
with the half-lives of Xel33 and Xel37, Since the Csl34 chain is fairly compli-
cated, several combinations of diffusion pairs are possible. Further work on
this problem is needed. o

All the data point to substantial diffusion only as vapors, or perhaps as
noble gases. Thus serious buildup of the high-cross-section rare-earth poisons
appears quite unlikely, since they have no gaseous precursors and since their
fluorides have exceptionally high melting and boiling points. Serious buildup
from molten-salt intrusion into the faults in the graphite likewise appears
unlikely; these faults would fill almost at once and therefore have essentially
no -exchange with the main body of the fuel salt.

The core-drilled samples of graphite were returned to ORNL for additional
study. These cores are now being analyzed more completely for fission-product
content, especially Sr8°9,90, The results will be used in an attempt to charac-
terize better the mechanism of diffusion of fission products into the sound
areas of the graphite specimen.

REFERENCES

1. MSR Quar. Prog. Rep. Apr. 30, 1960, ORNL-2973, p 8k.

2. MSR Quar. Prog. Rep. Oct. 31, 1959, ORNL-2890, p 67.
6. CHEMISTRY

6.1 PHASE EQUILIBRIUM STUDIES

6.1.1 The System LiF-BeFo-ZrF4-ThF4-UF,

The inclusion of ZrF4 in the MSRE fuel greatly complicates the problem of
providlug a complete and systematic representation of the phase behavior of the
system to which the fuel belongs. The problem is alleviated to some extent by
similarities in the effects of ZrF,, UF4, and ThF4, and can thus be posed as a
search for differences. The recently completed investigation of the quaternary
system LiF-BeFo-ThF4-UF, (refi 1) and the current studies of the ternary system
LiF-BeF5-Z2rF4 are particularly helpful. .

In the quinary system proper, two composition sections, as shown in Figs.

. 6.1 and 6.2, have been examined. In the vicinity of the nominal MSRE fuel compo-
sition, LiF-BeF5-ZrF4-ThF4-UFs (T70-23-5-1-1 mole %, melting at 442 * 3°C), the
first two phases to precipitate do not contain UF4 or ThF4, but one of them does
contain ZrF4. As a consequence, the ratio of ZrF4 to UF4 in the remaining

liquid is decreased, giving rise to conditions such that the presence of oxide
results in the precipitation of UO,. rather than ZrOo. A good example of this
behavior was recently encountered in a freeze-valve test involving an intention-
ally slow freeze to induce as much segregation as possible. All the oxide in
the frozen portion, or "valve,'" appeared to have precipitated as UOp. The amount
of soluble oxide initially present, representing unintended contamination, prob-
ably corresponded to saturation with ZrOz. Actually the analysis of a filtered
sample gave TOO ppm oxide.

UNCLASSIFIED
ORNL-LR-—-DWG 56483

s20 l l | I |
L =LIQUID Zrf=7mole %
A=6LiF-BeF2~ZrI-; UR,= 1 mole %
500 [— . ‘
B8 =2LiF-BeF, Thi-:‘= 1 mole %
€ = 7LiF-6(U,ThIF, ss
CONTAINING 15 mole % UF,
‘;6 480 — ‘
w
o __-~—-._.
2 e ——— L |
< 460 ————
= -
o L+4
UEJ.
=3 | G
- L+A+EB+C
420 :
400 -
15 16 A7 18 19 20 24 22 23 24 25
BeF, {mole %)

Plg. 6.1. The Systém Lik-Bek ,-ZrF ;-ThF -UF , (7 mole % ZrF Section).

109
110

UNCLASSIFIED
ORNL-LR-DWG 56484

520 | | . |
L = LIQUID ] Zrl74= 5 mole %
A= 6LiF-BeF, ZrF, ' . UF=1mole %
500 |— .
8= 2L|F-'-Bef-'2 Thf—;= i mole %
¢ = TLiF - 6(U,Th)F4 3
TAINI o s
5 480 — CONTAINING 15 mole % Uf,
[ MSRE FUEL
|
@
>
2 460
o
&J - [ L :
= —
-\
- 440 L+ A4 ——
\
- L+A+8B+C
420
400
16 17 i8 19 20 24 22 23 24 25 26
BeF, {mole %)

Fig. 6.2. The System LiF-BeF,-ZrF ,-ThF ,-UF , (5 mole % ZrF, Section).

When the fuel freezes completely under equilibrium conditions, much of the
UF4 and ThF4 appears as a solid solution in a familiar phase whose. stoichiometry
is TLiF.6(U,Th)F4. The ratio of UF4 to ThF4 in this solid solution is lower
than in the equilibrium liquid; such has commonly been the case for this T:6
phase in the systems containing both UF4 and Th¥4 with alkali-fluorides. A re-
liable estimate for the UF4 content in TLiF- 6(U Th)F4 precipitated from the MSRE
fuel is 15 mole %.

6.1.2 The System LiF-BeFo-ZrFg4

The anticipation that the 5 mole % ZrF4 along with 1 mole % each of ThFg4
and' UF4 in the nominal MSRE fuel might be approximately modeled by T mole % of
ZrF4 in the ternary system LiF-BeFo-ZrF4 has encountered a difficulty. No ter-
nary compounds are found in the LiF-BeF5-UF4 or the LiF-BeFo-ThF4 systems nor in
the quaternary system obtained by combining them; ternary compounds are quite
generally rare in fluoride systems. Nevertheless, the LiF-BeFo-ZrF4 system con-
tains a new ternary compound whose composition, probably 6LiF-BeFz-ZrF4, is near
enough that of prospective fuels to give the compound a predominant role in the
crystallization paths of fuels or fuel models. Solid solution or substitution
of ThFs or UF4 for ZrF4 in the ternary compound does not occur.

Due to the pronounced fluoride affinity of 7rt4 ions, ZrF4 is an exception-
ally good freezing-point depressant for LiF. This factor, combined with the
absence of LiF-BeF, compounds containing less than 33% BeF,, gives rise to
liquidus temperatures, near 70% LiF in the LiF-BeFo-ZrF4 system, that are 4O to
50°C lower than for comparable compositions in systems containing UFs and ThF4
rather than ZrF4. Thus the intrusion of the ternary compound was more probable
because of the extended lower temperature limit of stability of the liquid state
in the LiF-BeF>-ZrF4 system. In any case the LiF-rich corner of the LiF-BeFo-
ZrF4 system, shown in Fig. 6.3, provides high-temperature liquids which appear to
be favorably optimized with respect to the following properties, all of which
are relatively low melting point, viscosity, vepor pressure, density, and
corrosivity. ‘ : : : ’
11

UNCLASSIFIED
ORNL- LR-DWG 55482

PRIMARY PHASE AREAS

1 2LiF-2ZrF,
II 3LiF-ZrF,

I 6LIF-BeF, ZrF,
IV Lif

V 2LiF-Bek,

VI 3LiF- 4Z/F,

2LiF-BeF, A~ 1{453)

Fig. 6.3. The System LiF-BeF,-ZrF,.

6.1.3 The System LiF-BeF,

The selection of LiF-BeF, (66-34 mole %) as the coolant composition for the
MSRE led to a re-examination of the portion of the phase diagram near the com-
pound LizBeF4, which revealed barely noticeable discrepancies in a previously
published drawing. '

Figure 6.4, essentially a corrected drawing of the previously intended
phase diagram,® includes the recently re-established intormastion that LisBeF4
melts incongruently at 453 *+ 3°C and that the melting point of the coolant is
450°C.

6.1.4 Phase Equilibria in the System NaF-BeF5-ThFs

Investigation of phase equilibria in the system NaF-BeF>-ThF4, as shown in
Fig. 6.5, was completed. The invariant equilibria in the system are listed in
Table 6.1,

6.1.5 The System CeF-ZrF,

In a continuing study of phase equilibria in the alkali fluoride-MF4 systems,
where MF4 1s ZrF4, HfF4, ThF4, or UF,s, additional thermal-analysis data were
obtained for CsF-ZrFs (refs. 3 and 4). The study is hampered because the high-
temperature crystal modification of the compound CsF-ZrF4 is not stable at room
112

UNCLASSIFIED
ORNL—-LR-DWG 16426R2

900

800 \\
700 \

600 |—— LiF + LIQUID \
500 \ /

TEMPERATURE (°C)

: /////// BeF, + LIQUID
400 zur-;;;\\\ .
+ N ’
LIQUID
LiF + 2LiF - BeF, o
' by 2LiF - BeF, + BeF, (HIGH QUARTZ TYPE)
300 © . | l v
5 2LiF-BeFp & ! !
~ 2 3 LiF - BeFp + BeFp (HIGH QUARTZ TYPE)
P Ber, il ! |
200 LiF-BeFp 5 LiF - BeF2+ BeF2 (LOW QUARTZ TYPE)
LiF o 20 30 40 .50 60 70 80 90 BeF,
BefF, (mole %)

Fig. 6.4. The System LiF-BeF,.
temperature and because the solid phases containing less than 34 mole % ZrFa
are unusually hygroscopic and susceptible to hydrolysis.

From the latest results, as reported in detail elsewhere,5 the system CsF- .
ZrF4 can be summarized in terms of the invariant equilibria and singular points
listed in Table 6.2.

6.1.6 Crystal Structure of LiF:SbFs

The compound LiF-SbFs has been isolated during investigations of the use
of HF-SbFs solutions in the dissolution of rare-earth fluorides from MSR fuels.®
The unit cell and space group of LiF:SbFs have been determined from single
crystals, LiF+SbFs is monoclinic, with the following cell dimensions: '

5.42 + 0.02 R, ¢ = 7.50 + 0.02 &,
b =5.18 + 0.02 4, B = 93.6 * 1°.

a

n

Systematic extinctions indicate that the space group is Cs®-Im. The theoretical
crystal density is 3.83_g/cm3 if two formla weights per unit cell are assumed
to be present. Optically the crystals are anisotropic with a birefringence of
<0.002 and a mean refractive index of 1.393.
UNCLASSIFIED
ORNL-LR-DWG 30408

TEMPERATURE IN °C |
COMPOSITION IN male 7,
£ = EUTECTIC

P = PERITECTIC

PRIMARY PHASE FIELDS
. A=NaF
B = 2NaF ‘ThF,
C = 2Naf- Bef,
0=a-NaF-Th;
E = 3NaF-2ThE,
F = NaF-2Thf,
G = NaF - BeF,
H = NaF- BeF2'3ThE‘
I =Thi
J = BeF,
K = a-4NaF -ThE,

£834 L = B-4NaF Thi
M= B-NgF -Th§,
a

NaF ThE, )
PT3T
£ 705,
3NoF-2Thk, A
£ 690 P683 .
ZNGF'ThF-;
V4
"D
£ 618

P 645
ANaF-Thi, AR
N\
"

Lo
\ \ d‘%‘j(\)@

N, \ * . 00 =
XO%\&E 509 555 M \—._ ‘ A
. \
Z 375
' c - S 37 500 } £527
/\/ . N _AE320\ 6. o7 AV NL—\/ Pser Be,

NaF-2Thi,

o)

3 2

% N \A
p O
?
NaF \\/

995 £5707 2NoF-BeF, £ 34 NoF-BeF, £ 375 £ 365 548

Fig. 6.5. The System NoF-Ber-ThFA.

6.1.7 Index to ORNL Work on Fused-Salt Phase Studies

Data on many partially studied systems are cited in a decennial index®
(1950 to 1960) which attempts to include all ORNL reports of fused-salt phase
studies, thereby supplementing a previous compilation® of diagrams that are
complete, or nearly so.

6.1.8 Solubility of HF in Molten Fluorides

As part of a systematic study of gas solubilities in molten fluorides, the
solubility of HF in the molten mixture LiF-BeFs-ZrF4-ThF4-UFs (65-28-5-1-1 mole %)
was determined; this melt corresponds closely to the fluoride mixture proposed
as the fuel for the MSRE and is the same as that used for the recent determina-
tion® of the . solubility of BFa. A comparison of the solubility behaviors of the
two acid gases in molten-fluoride solvents is expected to provide a basis for a
better description of the liquid state of molten-sali systems.
114

Table 6.1. Invariant Equilibria in the System NaF-BeF>-ThF4

Composition (mole %) Invariant Temp.
Solid Phases Present NaF BeF o> ThF4 Behavior (°c)
ThF4 ss, BeFp, 2 96 2 Peritectic 527
NaF .BeF 2. 3ThF4 .
NaF.BeF5, BeFy, 43 55 2 Eutectic 365
NeF .BeF 5+ 3ThF4 '
NaF.BeFo, NaF:2ThF,, W 51 2 Peritectic 375
NaF .BeF 5+ 3ThF4 ‘ _

NaF -BeF,, NaF.2ThFg4, 48 50 2 Peritectic 375
NaF‘nThF4

NaF-BeFp, 2NaF-BeFy, 55 43 2 Eutectic 320
NaF . ThF 4

oNaF..ThF4, 2NaF-:BeFs, 57 41 2 Peritectic 415
NaF .ThF4 . .

2NaF -ThF4, 2NaF-BeFy, T2 22 6 Eutectic 509
NaF '

ONaF-ThF4, 4NaF-ThF,, 76 10 14 Peritectic 540
NaF - ' : | - .

2NaF-ThF4, 3NaF-.2ThF,, 62 2 36 Peritectic 683
NaF - ThF4 :

NaF -2ThF4, ThF4 ss, 42 31 27 Peritectic 740
N&F'B&F2‘3ThF4 .

~

The experimental method for determining the solubility of HF in molten fluo-
rides has been described!® in connection with measurements in the NaF-ZrF4 system.
Additional measurements of the solubility of HF in molten fluorides have been -
reported!? for the systems NaF-BeFo and LiF-BeFo. The method involves saturation
with HF at constant temperature and pressure; then a portion of the saturated
liquid phase is transported at constant pressure to an isolated section of the
apparatus and stripped of dissolved HF by helium sparging. The evolved HF is
measured by absorption in a standard solution of aqueous KOH.

Within experimental precision, the solubility of HF follows Henry's law over
the pressure range studied. Results, expressed as Henry's law constants, are
summarized in Table 6.3. The temperature dependence of the Henry's law constants
indicates a constant heat of. solution of approximately -5.65 kcal/mole over the
temperature range studied. The entropy of solution, ASp, calculated for a pro-
cess involving a concentration-in moles per liter in the liquid phase which is
numerically equal to the pressure in atmospheres in the gas phase is -15.3 eu at
1000°K. The entropy of solution, AS., calculated for equal concentrations of
HF in the gas and dissolved phases at 1000°K was found equal to -6.65 eu. The
precision of the experimental determinations shows a 1. 01% standard deviation
of experimental values of K from values calculated by the equation
1n K = ASp/R - 2H/RT.

6.1.9 Solubility of BFs in MSRE-Type Fuels®

The use of a nuclear poison dissolved in the molten fluoride fuel might be
a convenient method of controlling nuclear reactivity; if the nuclear poison were
Table 6.2.

Invariant Equilibria in the System CsF-ZrFg4

Invarient .
ZrF4 Conc Temperature Phase Rsaction at
(mglé %)' ‘1(:'00) Type of Zquilibrium Invariant Temperature
7 616 Eutectiz Liquid S=CsF + 3CsF-ZrFa4
25 782 Congruent melting point Liquid == 3CsF.ZrF4
3CsF.-ZrF4 4
33-1‘/3 527 Incongruent melting point Liquid + 3CeF-ZrF, e= 2CsF-ZrF4
2CsF.2rF4 .
34 527 Peritectic Liquid + 3CsF:ZrF4 == 2CsF-ZrF4
38-1/2 416 Eutectic Liouid * 3CsF-ZrF4 + 2CsF.ZrF4
50 511 Congruent melting point Liquid &= a—CsF-ZrF4
Q-CsF+ZrFq4
50 228 Inversion CsZ.ZrF,4 C-CSF+21F4 e B-C3F.ZrF4
58 663 Eutectic Liquid == CsF.ZrFy + PB-CsF-27ZrF,
56 579 Peritectic Liquid + 0-CsF:22rFy == ZrF4
66-2/3 579 Incongruent melting point Liquid + ZrF, 2= Q-CsF-2ZrF4
' CsF.2ZxF4
56-2/3 LeT Inversion CsF.2ZrF4 Q-CsF.2ZrF, e= B-C(sF.27rF.

SLL
116

Table 6.3. Henry's Law Constants for the Solubility of
HF in LiF-BeF5-ZrF4-ThF4-UFs (65-28-5-1-1 mole %)

Temp. Saturating moles HF
(°C) Pressure (atm) cc melt-atm

x 10°

500 1.40 ) : 17.5
' 1.66 ) 17.0

2.20 4 16.6

Avg. K 17.0

Calc. K 16.8

550 1.47 13.2
1.68 13.3

2.25 13.5

Avg. K 13.3

Calc. K 13.h4

600 1.46 11.0
1.82 10.6

2.36 10.9

Avg. K 10.8

Calc. K 11.0

TOO0 1.40 7.9
1.80 8.3

2,22 7.8

Avg. K 8.0

Calec. K 7.9

a chemically compatible gas and easily removed by volatilization, much greater
versatility could be achieved. Because of the relatively high neutron-absorption
cross section of boron, the volatile compound, boron trifluoride, is being evalu-
ated for this application. Accordingly, the solubility of BF3 in the molten
mixture LiF-BeFy-ZrF,-ThF,4-UF,4 (65-28-5-1-1 mole %) was determined. .

‘The method used for measuring the solubility of BFa was the same as that
previously used*® for determining the solubility of HF in molten fluoride mix-
tures. Chemical analyses of BFsz were made by a method described by Booth and

Martin.?

The solubilities of BF3 were measured between 500 and TO0O°C at pressures
ranging from 1.2 to 1.9 atm; as shown in Table 6.4. The solubility was found to
obey Henry'!s law and to decrease with increasing temperature. The temperature
dependence of the Henry's law constants, expressed as moles of BF3 per liter of
melt per atmosphere of BFs, is shown in Fig. 6.6. The enthalpy of solution was
constant, at least within experimental error, and equal ‘to approximately -15.1
kcal/mble over the temperature range studied. The entropy of solution, ASp,
calculated for a process involving a liquid-phase concentration (in moles per
liteér) numerically equal to the pressure (in atmospheres) of BF3 was equal to
-22,2 eu at 1000°K. The entropy of solution, AS., calculated for equal concen-
trations of BF3z in the gas and liquid phases at 1000°K was- found equal to -13.5
eu.

The p01soning effect of dissolved boron is proportional to the atomic concen-
tration ratio of boron to U2%%, Assuming a 20% enrichment factor for U235 in
117

Table 6.4. Solubility of BFs in the Molten Mixture
LiF-BeF5-ZrF4-ThF4-UFy4 (65-28-5-1-1 mole %)
Saturating Saturating _ K
Temperature Pressure BFa Solubility moles BFg3 )
(°c) (atm) (moles/cc melt) cc melt-atm
x 10° X 10°
500 1.21 31.L4 25.9
1.45 40.1 27.7
1.60 42.3 26.5 .
Avg. K 26.7 * 0.5
Calc. K 26.5
550 1.32 17.3 13.1
1.57 20,8 x3.3
Avg. K 13.2 = 0.1
Calc. K 1k.5
600 1.30 11.17 8,59
1.57 13.08 8.3k
1.78 14.03 7.91
Avg. K 8.26 + 0.25
Calc. K 8.46
700 1.37 5.1k 3.75
1.63 547 3.35
1.90 6.30 3.3k4
Avg. K 3.46 * 0.17
Calc. K 3.51
UNCLASSIFIED
ORNL—-LR-DWG 57692
TEMPERATURE (°C)
40 70[0 G(I)O 5?0 5(|)0
J
20 //
/_E /
ol T /®
H{ //
R .
El= /
NQ 6 /
<
4
®
/
)4
2
9 10 i2 13
10.000/7 ok

Fig. 6.6. Température Dependence of Henry’'s Law Constants for the Solubility’of BF 5 in LiF-vBer-ZrFA-

ThF ,~UF, (65-28-5-1-1 mole %).
. 118

the molten mixture LiF-BeFo-ZrF4-ThF4-UF4 (65-28-5-1-1 mole %), the atomic con-

centration of U235 in the molten mixture would be 2.70 x 10*° gram atoms per

gram of melt. Thils value can be converted to volume concentrations by the density
. 13 .

equation,

o (g/cé) = 2.8k - 0.00056t (°C),

which has been experimentally measured for the MSRE fuel mixture (see Sec. 6.3.1).
By similar calculations, the solubility of BF3 in the molten fluoride mixture can
be expressed as gram atoms of boron per cubic centimeter of melt at various
pressures of BF3. The atomic concentration ratio of boron to U235 as a function

. of the BFs pressure is shown in Fig. 6.7 in a convenient form for indicating the
regulation of the partial pressure of BF3 required to control the nuclear reac-
tivity of the reactor. The temperature sensitivity of BF3 poisoning is illus-
trated by the temperature dependence of the atomic concentration ratio of boron
to UZ35, Fig. 6.8, for BFs pressures of 1 atm.

Although low-solubility regions of practical interest have not been explored,
extrapolations of the experimental data to lower BF3 pressures are not expected
to be in serious error since the solubility appears to follow Henry's law. Also,
minor alterations in the composition of the molten fluoride mixture are expected
to give small changes. The physical properties and solubility of BFs in the
molten fluoride fuel of the MSRE appear to be suitable for purposes of nuclear
control; corrosivity aspects are not obviously bad and are receiving further
study. Corrosion encountered during the solubility:measurements has been attri-
buted to sulfur. impurities which have been identified in the BFa.

UNCLASSIFIED
ORNL-LR-DWG 57694 -

UNCLASSIFIED TEMPERATURE (°C)
ORNL-LR—DWG 57693 700 600 550 500
S .4.0
| | |
/.
—_ 500°C ——
S| Zlw
x(g 4 & |10 .
a|% / > 20 /
124 (2] ,. _m %]
£| € L El §
2| g : 3|5
| o ! )
o & ~ . z// 2 o
~—— 3 / (o} ®
) £ /
iz ® < 1.0
a / 1< /
v 4
z & 0.8 /
S / 550°c_~ 2 ~
2 2 & //
& M = 0.6
z 8 /
600°C
% /A/ § /
o L A o 0.4
= [e]
3 ’,/”/" — 700°C = o
o] 0.2
0 0.5 1.0 1.5 2.0 9 10 1 12 13 19
SATURATING BF; PRESSURE (atm) : 10,000/ 7 (o)
Fig. 6.7. Pressure Dependence of the Atomic Con- Fig. 6.8. Temperature Dependence of the Atomic
centration Ratio of Dissolved Boron to U235 in the Concentration Ratio of Dissolved Boron to U233 in the

Molten Mixture LiF-BeF,-ZrF,-ThF,-UF, (65-28-5- Molten Mixture of LiF-BeF,-ZrF,-ThF,-UF, (65-28-5-
1-1 mole %) ot Various Pressures of BF,. 1-1 mole %) at a BF, Pressure of | atm.
119
6.1.10 Freezing-Point Depressions in Sodium Fluoride

Measurements of the freezing point of NaF were made with solute fluorides
selected from the viewpoint of associating structural characteristics with thermo-
dynamic behavior. Three fluorides whose interionic distances approximate that
of NaF are CaFp, YF3, and ThFs. The significant structural difference among

" these solutes lies in cation charge. :

The results in Fig. 6.9 conform to the well-recognized principle that the
larger the cation charge of the solute the greater the freezing-point depression.
Stated differently, the stronger the electric field of the solute cation, the
greater will be the attraction for fluoride ions, which, in turn, will make the
establishment of long-range order (crystallization) more difficult for sodium
fluoride. For the solutes studied, the deviations from ideality manifested by
the trivalent fluoride are somewhat greater than the arithmetic average of the
deviations for the divalent and the tetravalent flunride.

. ZnFp, and MgFs>, as solutes, have about the same cationic size and charge.
Presumably, any significant structux;g.l difference is the result of the greater
polarization associated with the Zn= ion. '"The results in Fig. 6.10 demonstrate,
as expected, that the greater the polarization, the greater the freezing-point
depression. ‘

6.2 OXIDE BEHAVIOR IN FUELS ‘ . e

6.2,1 Tdentification of Oxides Precipitated from Fuels

A convenient means of providing regulated oxide contents with inconsequential
alteration of the solvent composition is the addition of known amounts of BeO to
fuel mixtures based on LiF-BeF,. A sequence of eight mixtures was prepared in
which the ZrF4/UF4 mole ratio varied from O to 15, and to which BeO was added in "
an amgunt generally equal to the number of moles of UF4 present (1 mole % in six =
cases). ‘ .

UNCLASSIFIED

-LR- UNCLASSIFIED
ORNL-LR-DWG 54586R ORNL - LR~ DWG 54583R

\‘ [ — e fi\

-20 % ] \
NN N

N

£ : .
T <
W \ 2 \
& —60 [~ M-F DISTANCE (A) N \ g \;:\\
' * CoF,:236 DR 3
KO _go |- * YF3: (8)12.25-2.32,(1)260 \ o M- F DISTANCE (A) \
* Thfy: 2.3-2.4 ESTIMATED 50 s MgF,: (4)1.997, (2)1.982 \
- * ZnF,: (4) 2.04,(2)2.015 A
NaF : 2.34 nfa: (4)2.04,(2) \
—100 — \ (0) INDICATES NUMBER OF BONDS t
{O) INDICATES NUMBER OF BONDS 4 -60 \
o ] |
: -70
0 S 10 5 0 5 10 15 20
SOLUTE (mole %) SOLUTE (mole %)
Fig. 6.9. Freezing-Point Depressions in NaF Fig. 6.10, Freezing-Point Depressions in NaF Caused

Caused by Ct‘J’Fz, YF,, ond ThF,. by Mgk, and ZnF,.
120

After having been heated to 750°C, the mixtures were frozen, and the frozen
ingots were examined petrographically and by x-ray to establish the identity of
the precipitated oxides. As expected, no BeO was found. At ZrF4/UF4 mole ratios
of 4 or greater, the only oxide phase was monoclinic ZrOp, while for ratios of
2 or less only UO> was detectible. These qualitative results, coupled with
related information accumulated over a period of years, led to the specification
of 5 mole % of ZrF4 in the MSRE fuel as a preventive against precipitation of
UO0o. }

The monoclinic ZrOp precipitated in the presence of UF4 appears to be pure;
at least it contains too little U0z to be readily detected. Perceptible solid
solution of UQz in monoclinic Zr0z was not mentioned in a recent investigationl?
of the effect of UO- on the transition of monoclinic to the tetragonal ZrO»
(high—temperature form) in which UO- is quite soluble. However, recent British
work?® masintains that solid solutions of as much as 10% U0, can be prepared in
monoclinic ZrOo.

6.2.2 Removal and Precipitation of Oxides from Fuels

- Recent improvements in sampling techniques and analytical procedures for
determining parts per million of oxide in fluoride salts have permitted a quanti-
tative study of the oxide removal from fluoride melts during hydrofluorination
and have also provided preliminary values for the solubility products of slightly
soluble oxides in fuel solvents. The analytical method involves fluorination
with BrFs to release oxygen from metallic oxides, and. subsequent volumetric
measurement of the evolved oxygen gas (see Sec. 6.5.1). This method gives total -
oxygen and does not distinguish between oxide ions and oxide from suifate, hy-
droxide, carbonate; etc.

Initially the experiments were directed toward checking the experimental
techniques and the analytical method. Samples of melt were obtained with a dip
filter comprised of a sintered copper disk which had been welded to a small-
diameter copper tube. In order to eliminate oxide contamination during handling,
an assembly was devised to allow the complete sampling operation to be accom-=
plished without exposure of the salt sample to air or moisture; this involved
the use of a sealed detachable port for transporting the sampler. The sample
stick, when in the port, could be treated with flowing hydrogen at 800°C, placed
in the reaction vessel for sampling, and transferred to an inert-atmosphere
glove box for removing the sample for analysis, all w1thout exposure of the
sample or the melt from which the sample came.

The removal of oxides from 1500 g of a molten mixture of LiF-BeFs (63-37 mole
%) by sparging the melt at 700°C with a mixture of hydrogen in hydrogen fluoride ,
(1:1) is illustrated in Fig. 6.11. The concurrent removal of sulfur from the
same melt  1s shown in Fig. 6.12. These results show that the HF-H> treatment
can remove dissolved oxides rather rapidly and further that the analytical re-
sults are reasonably self-consistent, with the possible exception of the tend-
ency to level off at 100 to 200 ppm. Removal of sulfur from the melt is also
rapid and is believed to proceed by reduction of sulfates to sulfides with Hy,
followed by volatilization by reaction with HF and stripping as HzS5. Oxide
analyses greater than 1000 ppm are attributed to S04 or some form of oxygen’
other than dissolved oxide ion; a similar explanation, involving a difficultly
decomposable radical, might account for the residual 100 ppm depicted in Fig. 6.11.

After removal of the initial oxide to a level of about 100 ppm, measurements
of the oxide concentration of the LiF-BeF, melt saturated with Be0O were made.
Specifically, after adding 2000 ppm oxygen as BeO to the purified mixture of LiF-
BeFs (63-37 mole %), a dissolved oxygen concentration of approximately 1000 ppm
121

UNCLASSIFIED UNCLASSIFIED

] ORNL-LR~-DWG 57695 OR -LR-DWG 5
2500 Lt 500 NL-L 7696
HF: Hp =~ 1 : J HF: Hp=~1
l\
2000 400 |-
E
Q
&
3 5
2 1500 E 300
=] -
< Q
2 a
2 z
(& ]
3 3
© 1000 U - S o0
x . w
© o od
=2
W
-J
oo}
(73]
500 100
0 5 10 15 20 o . P T 20
HF PASSED (equivalents) . HF PASSED ( equivalents)
Fig. 6.11. Removal of Oxide from LiF-BeF, (63-37 Fig. 6.12. Removal of Sulfur from LiF-BeF, (63-37

mole %) at 700°C by Treatment with HF-H, Mixtures. mole %) at 700°C by Treatment with HF-H, Mixtures.

was found, and was verified by further trials. Analysis of a filtered sample
taken from e mixture of LiF-BeF, (63-37 mole %) at 800°C, in which BeO crystals
had been grown by saturation with water vapor, also showed a dissolved oxygen
concentration of 1180 ppm. These results demonstrate the relatively high capacity
of the MSRE coolant salt for dissolved oxygen, and hence are of interest in
connection with ‘the potential use of thls mixture as a flush salt for cleaning
residual oxide from the reactor.

After addition of approximately 0.5 mole % UF4 to the mixture LiF-BeFs,
(63-37 mole %) and further purification of the melt with HF and hydrogen, small
weighed increments of BeO were added to cause a step-wise precipitation of UOj.
Measurements of the dissolved oxide concentration of the melt during the UO>
precipitation, for reasons as yet unaccountable, did not follow a step-wise
pattern, but rather indicated an approximately constant 600 ppm oxygen. Near
the end point of UOs precipitation the dissolved oxygen concentration rose to
and remained at approximately 1000 ppm upon further addition of BeO to the fluo-
ride mixture.

In an experiment to study the precipitation of oxides from & melt contain-
ing both uranium and zirconium fluorides, BeO was added in a step-wise manner to
a mixture of LiF-BeF (63-37 mole %) initially containing 0.5 mole % UF4 and
4,0 mole % ZrF4. The analytical results for zirconium and uranium in sequential
salt samples taken during the precipitation are shown in Fig. 6.13. The initial
precipitation of ZrOs followed by coprecipitation of UOp and ZrOp is more clearly
depicted by the plot of mole ratios in Fig. 6.14. These results imply that pro-
tection against precipitation of U0, in this melt is exhausted when the concen-
tration ratio of ZrF4 to UF4 falls to approximately 1.5. Again, the oxide
analyses during the ZrO- precipitation were considerably higher than expected on
the basis of a simple application of the solubility-product principle, and the
122

UNCLASSIFIED UNCLASSIFIED
ORNL-LR-DWG 57697 ORNL-LR-DWG 57698

I
INITIAL CONCENTRATIONS:
ZrF4 , 0.92 mole per kg of MELT
® UF,, O.11 mole per kg of MELT

ZIRCONIUM

() : Fe——500 ppm Op —

X | URANIUM STARTS
/TO PRECIPITATE

X e O o :
—0—T 'Q\\<URAN|UM
2

METAL FOUND IN SOLUTION (wt %)
o
/
" MOLE RATIO OF METAL FOUND IN SOLUTION (Zr/U)
H

\ L\ . e o
0 , 0
0 25 50 75 100 125 o | 2 3 a 5
BeO ADDED (g) ‘ BeO ADDED (moles)
Fig- 6.13. Reaction of BeO with ZrF, and UF, in Fig. 6.14. Behavior of ZrF, and UF, in LiF-BeF,
Li‘F-B‘eF2 (63-37 mole %) at 700°C. (63-37 mole %) in Reacting with BeO at 700°C.

oxygen céntent of the melt was constant at approximately 500 ppm over the range
of coprecipitation of ZrOz and UOp, then rose to the expected 1000 ppm when
excess BeO was flnally present.

An oxide concentration of 500 ppm converts to an anibn fraction of 10-3,
if this oxide concentration represents equilibrium in a solution containing 0.5
mole % UF4 at the onset of precipitation of UOg, then the solubility product of
U0, is

(UT4)(07)2 =5 x 107 x (1073)2 = 5 x 107°,

Similarly, the solubility product for ZrO, is 7.5 x 10~ °. Somewhat different

- values would be expected in the MSRE fuel, or at different temperatures, but a
representative or sample calculation is of interest. If the MSRE fuel contains
5 mole % ZrF4, and if the solubllity product for Zr0z is 7.5 x 10-9, then an

anion fraction of oxide
o= | = &/ 1:5 x 107°
B 5 x 10-2

- or 200 ppm oxide, might be an approximate tolerance level expected for the onset
of precipitation of ZrOs. Many other results at variance with such a low toler-

- ance level imply tge existence of complexed oxide in solution, and quadrivalent
cations ‘such as Zr appear to be the most likely complexing agents.

6.3 PHYSICAL PROPERTIES
6.3.1 Density of MSRE Fuel

The density of the quinary fused salt proposed for use in the MSRE, nominally
70 mole % LiF, 23 mole % BeF, 5 mole % ZrF4, 1 mole % UF4, 1 mole % ThF., was -
123

measured over the temperature range of 450 to 700°C, and is given by the equation:

o (g/ce) = 2.84 - 0.00056t (°C) * 0.0L1.

The measurements were carried out by a bouyancy method, using a platinum
-bob and a thermogravimetric balance, with the fused salt contained in INOR-8 and
blanketed from the atmosphere by flowing purified argon. Correction was made
for the surface-tension effect of the salt on the platinum suspension wire. In
spite of the argon blanket, a thin black oxide film slowly accumulated on the
surface of the salt, finally causing erratic surface-tension behavior. However,
the results of a second run, which was carried out as rapidly as possible with
a frech batch of salt, confirmed the first-run results within the accuracy
attached to the density equation. Analytical results on the salt recovered from
each run indicated inappreciable conversion of fluoride to oxide and also approxi-
mated the nominal composition wlllilu 1 wt % T'or each component.

6.3.2 Stochastic Correlation of Densities of Molten Fluoride Fuels
While many density datal®~'® nave been obtained for fused fluoride mixtures
of reactor interest, relatively little attention has been paid to correlations
of density with composition. In the study reported here, data on the systems
LiF-BeF> (ref 18), NaF-BeFo (refs 16, 17), LiF-BeF-UF4 (ref 18), and NaF-BeFp-
UFs (refs 16, 17) have been examined in an attempt to relate density to
composition,

Examination of Fig. 6.15 shows that, to a close approximation, molar volume
at 800°C is a linear function of the volumes of the pure components for LiF-BeFj,
and NaF-Belo. Further trial calculations led to the conclusion that the contri-
bution of BeFs and UFs to the molar volume can be estimated from the fluoride
ion volume alone; that is, Be'2 and, at low dilution, U4 jons appear to fit in
the interstices of a fluoride ion matrix.

The deneity of BeFs has been measured only at 800°C (ref 19). Jaeger®® has
measured the densities of LiF and NaF and gives equations for the temperature
dependence of density. One of the main objectives of the present effort has been
to test the approximation that partial ionic volumes and partial molar volumes
are constant with composition in the LiF-BeFs and NaF-BeF. systems.

Since the molar volume of BeFy appears to conform to that of the fluoride
ions alone, the gram ionic volume of fluoride jions is equal to one-half the molar
“volume of BeF». If such an lonic volume for fluoride ions is then subtracted
from the apparent molar volume of the alkali fluorides, an apparent, lonic volume
for the alkali metal lons is obtained. Taking into account temperature depend-
ence, a self-consistent set of ionic volumes for the ions is given by the
following equations:

Voo = (10.32 + 0.00225t) em®, (1)
ngt = (548 + 0.00355t) cnm®, (2)
Vi ,+ = (0.046 + 0.001525t) em?, (3)

vhere t is temperature in °C.

The application of these ionic volumes to solutions involving UF4 as a third
component re%uires an assignment for the volume contribution of UFs. As noted by
Zachariacen® for solids, the volume of the uranium fluorides can, to a good
approximation, be attributed to the fluoride alone, On this basis, 1 mole of
124

UNCLASSIFIED
ORNL—-LR-DWG 57699

26

N
E
L
w
>
o
-t
o
>
o
<I
J.
O
>

18 o4

Q ——= |[DEAL SOLUTION NaF-BeF,
16 | / '® OBSERVED NaF - BeF, —
IDEAL SOLUTION LiF—-BeF,
_ O OBSERVED LiF—BeF;
: ' : |
14 4 - .
0 20 40 60 80 100
Bef, (mole %)

Fig. 6.15. Molar Volumes at 800°C for the Liquid Systems NaF-BeF, and LiF-

Ber. "

the ternary melt contains x moles of fluorides available from the alkali fluoride,
4y moles of fluoride from UF4, and 2(1-x-y) moles from BeFs, or a total of 2-x-2y
moles of fluoride, and the molar volume for the ternary becomes

Ve = Vs (x) + Vp- (2 + 2y - x), (4)

vhere Vi is the molar volume of the melt at temperature t, and V,, 1s the volume
of the alkali ion and is given in Eq. (3). The molecular weight of the melt is
divided by Vi to give the estimated densities, which are compared with the ob-
served densities in Tables 6.5 and 6.6. S

The agreement is sufficiently good to make Eq. (4) appear useful. The third
component can be considered as any quadrivalent fluoride, and a similar equation
can be used for estimating the density of five-component mixtures containing
alkali fluoride, BeF,, UF4, ZrF4, and ThF4. This involves only the assumption
that for ThF4 and ZrF4, the volume contributions are -due to fluoride alone, just
as for UFg. '
125 g

Table 6.5. Comparison of Estimated and Experimental Densities
in the System NaF-BeFo-UFg

Density (g/cm3)

Composition (mole fraction) Temp. Difference
NaF BeF, UF4 (°c) Experimental Estimated (%)
0.68 0.30 0.02 600 2.2&9 2.323 +3.3

: 800 2.144 2.208 +3.0

0.70 0.20 0.10 600 3.024 3.061 +1.2
700 2.949 - 2.985 +0.9

~ 800 - 2,874 2.912 +1.0

Table 6.6.

Comparison of Estimated and Experimental Densities

in the System LiF-BeF>-UFg4

' ‘ Density (g/cm3)
Composition (mole fraction) Temp . '

Difference
LiF BeF, UF4 (°c) Experimental Estimated (%)
0.6 0.15 0.35 T00 " L.543 4.948 +8.9
0.4 0.13 0.47 800 5.163 5.138 -0.4
0.5 0.30 0.20. 600 3.683 3.865 +4.9
0.65 0.31 0.0k 600 2.326 2.452 +5.5
0.7 0.28 0.02 600 2.126 2.218 +4.3

e TmImagm ey Sin el s

For MSRE fuel (LiF-BeFo-ZrF4-ThF4-UF4, 70-23-5-1-1 mole %), Eq. (4) becomes

V, =15.523 + (4.31 x 1073t). | (5)

Converting Eq. (5) into densities we obtain
p =2,77 - (6 x 1074t),

which at reactor te%?eratures agrees within about 5% with the experimental
equation: (see Sec. 6.1.9):

p =284 - (5.6 x 107%4¢).

6.3.3 Embirical Equation for Fluidity in LiF-BeF,-UF4 Systems

Viscosity or its reciprocal, fluidity, is not a simple additive property.
However, it is frequently possible to relate fluidity to composition empirica’ly.
For low concentrations of UF4 in LiF-BeF5-UF4 the following approximation might
be useful: ‘

® = 0.0240 (2— - o.5929> e0-00582t

\Np-
126

vhere ® is the fluidity in reciprocal centipoises, t is temperature in °C, and
N__ is the number of moles of fluoride ions per mole of melt. Values calculated
from this equation are compared to those observed at the Mound Leboratories'®

in Table 6.7.

Table 6.7. Fluidity in the LiF-BeF>-UF4 System
(Reciprocal Centipoises)

Mole Fractions ® at 600°C d at TO0°C ® at 800°C
LiF BeF» - UF,4 Exptl. Estd.  Exptl. Estd. Exptl. Estd.
0.75 0.23 0.02 - - - - - -
0.70 0.28 0.02 0.1335 0.121k 0.2033 0.2179 - -
0.65 0.33 0.02 - - 0.1152 0.1792 - -

0.60 0.38 0.02 0.0901 0.0804 0.1570 0.1439 o.2h88 0.2576
. 0.55 0.43 0.02 0.0654 0.0623 0.1379 0.1115 0.1965 0.1995
0.50 0.48 = 0.02 0.0438 0.04k49 0.0862 0.0804  0.1497 0.1439

6.4 GRAPHITE COMPATIBILITY

6.k.1 Xenon Diffusion in Graphite: Effects of Xenon Absorption
in Molten-Salt Reactors Containing Graphite®2

Estimates were made of the xenon poison fraction in a hypothetical MSRE-
type molten-salt reactor, operating with an unclad-graphite moderator, under the
assumption that xenon transport to the graphite is not influenced by the liquld
film. The poison fraction at steady-state conditions depends on the. Xe'3S con-
centration in the molten-salt- fuel, the free gas space over the fuel, the pores
of the unclad graphite contacting the fuel, and the corresponding volumes .22

Xenon transport rates were considered for various combinations of generation,
burnout, decay, removal by helium stripping, and diffusion into graphite. Par-
ticular attention was given to the effect of the graphite porosity, permeability,
and xenon diffusion coefficient. These parameters govern the rate of xenon
diffusion into the graphite.

The computations established that permeability and/or porosity reduction
coupled with increased stripping rates effectively decrease the xenon poison
fraction and increase neutron economy. For the range of graphites currently
under consideration for the MSRE, increased stripping rates appear to be the
most feasible means of reducing the xenon poison fraction. At the circulation
rates considered, it is unlikely that neutron economy (with respect to xenon)
can be effectively enhanced through permeability reduction alone.

The use of unclad graphite in direct contact with the MSRE fuel leads to
several contingencies which require continual and thorough re-examination as con-
ceptual designs approach specification form. Among these contingencies are:

1. deposition of solid oxides arising from oxygen introduced
via the graphite,

2. 1invasion of the graphite by the molten fuel,
127

3. variable reactivity resulting from a variable Xe'3%

concentration in the graphite,
L. high xenon poison fraction.

The first three items could result in erratic reactor operation; the last
item is of importance from the standpoint of neutron economy. Research efforts
to yield information for resolving these problems include studies of: volatile
impurity contents of graph1tes,23' molten-salt permeation of graphltes24 (also
UF4 to UOo conversion due to impurities), effects of pore-size distribution of
graphites and wetting agents (see Sec. 6.4.2), and gas transport in porous
media.=>

In afirwrmn to the third and fourth difficulties mentioned, it has been
shown that Cs® for which Xe®® is a precursor, is not compatible with graphlte.
The role of Cs® in gas-cooled reactors has been discussed by Rosenthal® 27 and
Cantor.2®

A comparlson of proposed methods for xenon control has been presented by
Burch.®® More recent computations have been made by Spiewak 3© e results of
the two studies are quite different with respect to xenon control and graphite
requirements, as the earlier work was premised on high circulation rates and the
presence of a large decay dome in series with the pump. These features have
been eliminated in present designs, which lead to increased driving forces for
the xenon absorption by the graphite.

The feasibility of using low-permeability graphites under the more recently
proposed conditions, and disussions of the graphite parameters of interest have
been developed in too great detail®’ for inclusion here.

The effective porosity, as well as the diffusion coefficient for a given
graphite, influences the steady-state rate at which xenon is absorbed by graphite
under conditions of burnout. Low porosity and transport coefficients (diffusion
and permeability) lead to low Xe'>> absorption rates. Figure 6.16 reflects the
importance of the porosity, €, being low in achieving a favorable neutron

UNCLASSIFIED
ORNL-LR-DWG 56201A

5
o r=0%
s~
- 2 -
l: __.——"'—-—-—-

-l

E 3 il 7//
4] 20 _L—7"
o L
a /
g 2
g D=2x10"° cm®/sec
b4

{ .

2 4 6 8 10 20

€, POROSITY (%)

Fig. 6.16. Effect of Porosity on Xenon Poison Frac-
tion for D = 2 x 10~% cm2/sec. The symbol r refers to
the recycle rate or fraction processed by removal of
xenon in a stripper.
128

economy. The relevant porosity is the fraction of the graphite volume accessible
to helium. If the anticipated graphite for the MSRE has a volume of 6.2% acces-
sible to kerosene, this probably corresponds to an € of about 10%.

The effect of the permeability of the graphite to gases is shown in Fig.
6.17. The abcissa, DA, is proportional to the square root of the effective dif-
fusion coefficient, D (cm®/sec), or specifically DA = (D x 1075)°*5; thus a
permeability corresponding to D = 10~° cm®/sec, near the limit of prevailing
graphite technology, corresponds to DA\ = 10-®, and requires a,15% processing rate
to reduce the xenon poison fraction to half that for no xenon removal.

A faster circulation, at the same fraction processed, also decreases the

~ Xenon poisoning, as illustrated in Fig. 6.18.

UNCLASSIFIED UNCLASSIFIED
ORNL-LR-DWG 562024 ORNL-LR-DWG 562034
T T -6 j - 5
’ Kue=1x10"®  CEY AGOT l
%
r=0% €=10.52%
_ - - "__—-—::‘ p— 4 r=10 %
3 ~ L ’.—'/ ’\: ’
- LA el o )
y’//’//‘ LA - O
8 v d / AL LT & \ D=2x10"5 cm%sec
& A / yd ///’ £ § T~
4 L ~
: / Al / 4 A T z \ T
(Z) : / /i / // / g 2X10_6 \\
a / / / / // o 2
s /oo / /| -
© i 4 S R
g y y / /V 5 wo_7 \
: /' soy’ VMV < 4 -
/ / / € =10.52% \
///’/noo’ oa= (0 10—5)1’2 /sec ——— |
|t - = X cm
T 1!l Lot |
10=7 1076 1075 10-4 o
DX (cm/sec) 0O 5 10 15 X 20 25 30
PUMP RATE (ft”/sec)
Fig. 6.17. Xenon Poison Fraction at Various Values Fig. 6.18. Effect of Pumping Rate on Xenon Poison
of Recycle Rates and Diffusion Coefficients. Fraction.

Other important conclusions are as follows:

1. Experimental data have verified that the diffusion coefficient and the
permeability coefficient of a given gas-graphite system approach equality when
the coefficients of the systems under consideration are very low, that is approxi-
matély 10~/ cm?®/sec.

2. Low-porosity graphites generally possess low transport coefficients;
however, low-porosity values are not necessary indicia for low values of the
transport coefficients.

3. For an approximate geometry for diffusion equations involving the graphite
moderator, one may utilize the semi-infinite case when Dy, < 10”° cm®/sec. At
higher D values, the cylindrical case appears to be more sultable.

L. Of all the Xel35 control varisbles studied, the rate at which xenon is
removed from the reactor appears to be of primary importance.

5. Adequate stripping coupled with graphite permeability (and porosity) re-
duction affords an effective combination for decreasing the xenon poison fraction.
The latter cannot be controlled through permeability reduction alone.
129
6.4.2 Wetting of Graphite by MSRE Fused Salt

To furnish a semiquantitetive measure of the tendency of fused fluoride salts
to wet graphite, capillary-depression tests were carried out with TSF graphite in
contact with the quinary MSRE salt and with LiF-BeF, (62-38 mole %). Holes
varying in diameter from 1/32 to 1/4 in. were bored in graphite cylinders 1 in.
high and 1.5 in. in dismeter. The cylinders were floated in fused salt at 600°C
in nickel containers under an atmosphere of helium. The salt was then frozen,
and the levels of salt in the holes were measured.

No salt penetrated into the 1/32-in. holes and less than 1 mm into the 1/16-
in. holes. For the fuel salt, the capillary depressions were 1.1 and 0.7 cm,
respectively, for the 1/8- and 1/4-in. holes. These values are about twice those
calculated from the capillary-depression equation, even assuming a 180° contact
angle., For the LiF-BeF, salt, similar behavior was observed. The discrepancy
1s probably due to changes in the salt levels during freezing. In the larger
holes and on the outer perimeter of the graphite, a definitely convex meniscus
(contact angle >120°) was observed. Contact angles less than 90° are associated
with wetting or capillary rise, while angles greater than 90° are associated with
nonwetting or capillary depression,

Further measurements of capillary depression were made on the same assemblies
after treatment with an atmosphere of HF for 4 hr at 600°C. The 1/32- and 1/16-
in. holes were still empty, but depressions in the 1/8- and 1/4-in. holes were
generally lowered about 30%, approaching the theoretical values. The measurements
were repeated after adding pieces of metallic zirconium to the salts and reheating
to 600°C. The observed depressions were close to those observed with HF treat-
ment, although earlier observations had indicated that a strong reducing agent
should increase wetting.

The same experimental assemblies were then heated in open air under visual
observation. Just after melting, the salt-graphite meniscus was convex (i.e.,
depressed or nonwetting). After short exposure to air, the external contact
angle varied from convex to approximately 90° at dit'ferent positions along the
perimeter, causing the graphite pieces to cock. A surface oxide film formed, and
when the graphite pieces were moved up and down, some of the scum adhered to the
graphite. Scum accumilation produced contact angles less than 90°. At this
stage, some of the 1/8- and 1/16-in, holes became filled with salt. Salt could
be made to f£ill even the 1/32-in. holes by moving a fine wire up and down in the
hole. With continued accumulation of oxide film on the graphite, the pieces sank
into the salt, due to the pull of the concave meniscus, until the top surface of
the graphite was only slightly above the salt and the vertical sides were covered
with salt. Superficially, at least, the graphite had become thoroughly wetted.

The principal conclusions from these experiments are:

1. Pure fused LiF-BeFy based fluoride salts do not spontaneously wet
graphite (contact angle >90°).

2. The capillary behavior is at least qualitatively described by the usual
capillary-depression equation.

3. Treatment of the salt with HF or metallic zirconium apparently decreases
capillary depression (or increases wetting) by about 30%.

L. Exposure of the salt-graphite system to air at 600°C results in an oxide
scum which adheres to graphite and causes salt to give pronounced manifestations
of wetting.
130
6.4.3 Retention of BF3 on Graphite in a Molten-Salt Reactor

In molten-salt breeder reactors a gas with a'high cross section for neutron
absorption might be utilized for emergency shutdown or for additional control of
nuclear reactivity, '

In order to evaluate boron trifluoride for this a.pplication,31 experiments
were designed to determine the amount of BF3 retained on graphite after contact
with BF3 gas at 600 to 800°C and 10 to 50 psia for 2 hr, followed by stripping
of the adsorbed gas with helium.

The experimental assembly consisted of a reactor vessel, in an electric
furnace, connected with gas metering and recovery apparatus. The test specimens
were 1/4- by 2- 1/16-in Speer No. 2 reactor-grade graphite cylinders.

The test specimens were degassed by heating to 2000°C over a 1l6-hr period
in vacuum, stored under helium, then again degassed in dry helium in situ by
heating for 2 hr at 800°C. After cooling overnight the system was evacuated,
BF3 gas was admitted, and the samples were exposed at the experimental pressure
and temperature for 2 hr. The BF3 gas in the system was removed by stripping
the samples in a stream of dry helium for 2 hr. The treated graphite samples
were wrapped in aluminum foil and stored in a vacuum desiccator.

Boron contents of the graphite samples were determined by spectroscopic
analysis and by measurements of the neutron absorption cross sections in the
pile oseillator.?2 In order to increase the sen51t1vity of the determinations
of neutron absorption cross section, 95% enriched B°Fs was used in several
experiments. :

Experimental data for BF3 retention on reactor-grade graphite were obtained
at 600, 700, and 800°C for BFsz pressures ranging from 9 to 45 psia. The BF3
retention was determined at three pressure levels for each temperature, repli-

cating several of the runs. A representative set of the data is given in Table
6.8.

The data give a mean residual boron content of 10 ppm for the graphite.
Statistical analysis of the data indicates a 125% confidence level for the ana-
lytical results, which is larger than the errors resulting from inhomogeneities
of the samples. The confidence limits for a replicated set of values at two dif-
ferent pressure and temperature levels are larger than the confidence level for
the analytical method, which indicateg/the absence of significant pressure or
temperature effects on the amount of boron retained by the graphite.

The small absolute values for the boron content, together with the lack of
detectable temperature or pressure dependence, indicate an absence of any rapid
reactions between BF3 and graphite at temperatures up to 800°C and pressures of
3 atm under the conditions of the experiments. The small quantities of boron
retained in the graphite are possibly held by lattice defects on the exposed
surfaces or by chemisorption.

6.5 ANALYTICAL CHEMISTRY

6.5.1 Proposed Procedures for Analysis of the MSRE Fuel*

A program was begun to develop methods for the determination of components
in the MSRE fuel, LiF-BeFo-ZrF4-ThF4-UF4 (70-23-5-1-1 mole %). Specifically,

*¥This review was provided by J. C. White, Analytical Chemistry Division.
Table 5.8..

Retentions of Boron Trifluoride on Reactor Graphitz

Rate of

Total He Neutron Absorption
Volume He Flcw Boron Content of Graphite (ppm) Cross Section, '
: . Pressure (Liters (1iters by Spectroscopic Analyeis Pile Oscillator
Run No. (psia) STP) STP/hr) Surface Core Core Av. (millibarns)
600°C
8 23.7 22.0 11.0 15 T T
, 700°C
3 3h.7 14%.9 7.5 100 5 _
| 10 T 10.5
15 k.7 3.k h.2 50 20 20 9.5%
800°C o
L 33.2 16.2 - 4.0 100 5 . h 9.2%
3
Control <0.5 <0.5 <0.5

*
95% enriched B°Fs.

1€l
132

methods are required for the determination of uranium, thorium, zirconium, and
lithium as major constituents; iron, nickel, and chromium as corrosion products;
and oxygen-as metallic oxides. Uranium and oxygen are considered of greatest
importance to reactor operations. Although suitable methods are available for
all theseelements, considerable modification will bhe necessary to adapt them for
use in the High-Radiation-Level Analytical Facility. The sampled fuel will, of
course, be extremely radiocactive, and methods of analysis must be compatible
with remote-handling techniques.

A survey was made of procedures for dissolution of the fuel. Methods tested
included fusion of the sample with various fluxes such as potassium acid sulfate,
" boric oxide, and sodium acetate; treatment with various mineral acids; and elec-
trochemical decomposition in which the fluoride is removed by deposition on a
lead electrode. At the present time the most satisfactory method involves treat-
ment of the sample with a mixture of hydrochloric, perchloric, and boric acids,
folloved by extensive fuming with perchloric acid. By this treatment, the fluo-
ride compounds are taken into solution, the fluoride is volatilized, and the
uranium is oxidized to the hexavalent state. However, any zirconium which is
initially present as ZrOz is not completely dissolved. Studies are now con-
tinuing on this problem. '

Thorium

An amperometric titration method was developed for the determination of tho-
rium in the MSRE fuel. In this method thorium is titrated with ethylenediamine~
tetraacetic acid (EDTA) in an acetate-buffered medium at pH 4.5 using Fe(III) as
an indicator ion. In order to detect the equivalence point, a potential of +400
mv vs S.C.E. is applied to a platinum electrode. At this potential Fe(II) is
not. oxidized; however, Fe(EDTA) 2, which forms after the titration of thorium is
complete, is oxidized and the equivalence point is indicated by a linear increase
in current. Cations which form EDTA chelates, which are more stable than Fe(II),
interfere if the reaction with EDTA is rapid. Uranium and zirconium do not inter-
fere under the conditions reported herein., Of the anions, phosphate interferes
seriously, but a l-to-l mole ratio of fluoride can be tolerated. The other
common anions do not interfere when present in & 10-to-l mole ratio of thorium.
The coefficient of variation of replicate titrations is about 1%.

O§Egen

- The determination of oxygen at the ppm level in the highly radioactive metal
fluoride salts is being investigated. The literature was examined with reference
to ways of determining oxygen. It was decided that the method offering the
greatest possibility of success under the rigorous condition of extreme radio-
activity, which necessitates remote handling, is the fluorination procedure that
has been in use for some time for the analysis of nonradiocactive materials. The
equipment. and procedure have been described by Goldberg EE_EL.SS The principle
of the method is based on the reaction of the additive compound of potassium
fluoride and bromine trifluoride with metal oxides to release the oxygen in the
form of a gas. In Goldberg's apparatus, the volume of the evolved.oxygen is
measured after the bromine trifluoride is condensed at liquid-nitrogen tempera-
tures.

. Goldberg's ‘apparatus, although effective in bench-top operation, was not
considered to be an optimum design for use in a hot cell. Accordingly, an appa-
‘ratus was designed and built in the ORNL shops. The apparatus consists of a
nickel reaction tube, a clamshell tube furnace, a cold trap, and l-liter collec-
tion flask of aluminum metal, together with associated valves and tubing. It
is intended that a slight overpressure of helium be maintained throughout the
133

system at all times except in the evacuated collection flask. The reaction tube,
cold traps, and collection flask are connected to the system by quick-disconnect
joints, which are sealed by Teflon gaskets. These joints enable all parts ex-
cept tubing and valves to be easily removed from the apparatus. These parts may
be reattached to the system and may be made leak-tight without requiring more
than 5 to 10 1b of pressure. The reaction tube is capped before it is trans-
ported, and the cap 1s contained in a slide rack attached to the apparatus stand
during removal or attachment of the reaction tube.

The additive product KBrFs will be made externally and stored until needed.
A gufficient quantity can be transferred to the reaction tube in a dry box under
an inert atmosphere, It is anticipated that the sample will be added to the
reaction tube in the same manner. The tube will be capped and transferred to
the apparatus. After the apparatus has been flushed with dry helium, the valve
to the reaction tube will be closed. The reaction tube will then be heated by
means of a controlléd-temperature furnace that is part of the apparatus. After
heating for a suitable period of time, the furnace will be swung away from the
reaction tube. The reaction tube will be cooled, and the cold traps will be
brought down to liquid-nitrogen temperature. The valve to the reaction tube will
then be opened, and the valve to the collection flask will be opened. The system-
is then flushed with helium into the collection flask. The collection flask can
then be closed off and removed from the cell. The oxygen content of the gas in
the collection flask will be determined by gas chromatography.

The apparatus has now been completed and leak-tested. The proposed pro-
cedure is currently being evaluated.

Alternative methods are being investigated for determining the oxygen con-
tent of the helium stream following the release of the oxygen by the fluorination
procedure. A Hersch cell is being built and will be investigated for the potenti-
ometric measurement of oxygen in gases. Marsh>* has reported the successful use
of this cell in connection with the DIDO program.

1

Analysis of MSRE Cover Gas

Because UOp could precipitate from MSRE fuel by reaction of the fuel with
contaminants in the gas blanket, the maintenance of an uncommonly pure atmosphere
of helium cover gas is required. Tentatively permissible concentrations of oxy-
gen and water as low as 0.1 ppm have been proposed. The monitoring of the fixed
gases such as hydrogen and hydrocarbons may also be required at higher concen-
tration levels.

On the basis of literature surveys, gettering by active metals at elevated
temperatures, such as titaniwm at 1200 to 1400°F, has been proposed as the most
convenient means of removing contaminants. In cooperation with the Reactor
Division, a small facility to test the efficiency of such a purifier has been
designed and is now being fabricated. Helium deliberately contaminated at a
level of about 100 ppm of oxygen will be passed over the heated metal, and the
effluent gases will be analyzed with a Greenbrier 9.50 chromatograph that is
equipped with a short column packed with type 5A Linde molecular sieves to
permit the detection of ~l1 ppm of oxygen.

Current studies are directed toward the development of more sensitive de-
tectors so that continuous monitoring of oxygen, water, and other contaminants
at concentration levels less than 1 ppm can be carried out with process-gas
chromatographs. In the most promising system of detection, the effluent gases
from the chromatographic columns are passed at a pressure of about 500 p through
an ionization-gage tube>> in which the gases are bombarded with electrons of
134

energy between the ionization potential of the helium carrier gas and that of the
contaminants. Positive ions which result from the collision of electrons with
molecules of contaminant gases are collected on & negative plate to yield a cur-
.rent proportional to the concentration of each contaminant in the eluted gas
stream. Although this technique 1s theoretically capable of sufficient sensi-
tivity, adequate response has not been obtained because of the high background
from helium ions which are produced by collisions with electrons whose energy
exceeds the accelerating potential of the anode. These high-energy electrons
result from the energy distribution of electrons emitted by the high-temperature
filament. Special lonization tubes equipped with oxide-coated cathodes to re-
duce the energy spread of the electrons have been found to offer improved sensi-
tivity for contaminants such as hydrogen and methane, but actually lower sensi-
tivity for oxygen and nitrogen because of the gettering action of barium that is
evaporated from the cathode. Improved tubes will be designed with lanthanum
boride, LaBg, emitters®® or with baffle systems to exclude the evaporated metal
from the collision space.

6.5.2 Stripping Rate for CO, from Water in an MSRE Pump-
Demonstration System

In comnection with the problem of xenon removal from the MSRE, the rate of
COs- removal from water in an MSRE pump-demonstration system has been followed by
pH determinations. In the initial experiments, the COs removal rate was about
10% of the maximum theoretical rate based on the fractional volume bypassed
through the stripper. After suitable modifications of the system to increase
the area of the liquid-gas interface, the COp removal rate was increased to about
85% of theoretical. Since CO» is more soluble in water than xenon in MSRE fuel
by a factor of approximately 104, even higher efficiencies might be expected for
xenon removal from a fuel,

In the rate studies, the pump system, filled with distilled water, was .
initially charged with CO, at approximately 1 atm by addition of solid COp to
the circulating stream. The pH was measured frequently to determine the COp
removal rate. The pH values were converted to the logarithm of the molarity of
CO» for plotting as a function of time as illustrated in Fig. 6.19. From the
linear portions of the plots, approximate values for the half-life of the de-
~crease in COp concentrations were obtained. Reproducibility to within a few per
cent was achleved. Usually the flow bypassed through the stripping system was
50 gpm from a total pump volume of 100 gal. Assuming complete stripping of CO»
from a bypass flow of 50 gpm from a 100-gal system, a minimum theoretical half-
life (t 1/2 = (0.69) (100/50) of 1.4 min is obtained; increasing the bypass flow
to 85 gpm gives a minimum theoretical half-life of 0.8 min. Actual measurements
under the foregoing conditions gave half-lives of 1.9 and 0.9 min, respectively.
Numerous additional trials were useful in evaluating effects of changes in.
stripping arrangements.
135

UNCLASSIFIED
ORNL~LR-DWG 57700

107 : f
! }
5 LOOP FLOW : 1450 gpm
BYPASS FLOW: 50 gpm
SPRINKLER HEAD ON BYPASS FLOW —
RADIAL FLOW
> AIR FLOW : 4.3 13/ min
80WL LEVEL: 2in.
TEMPERATURE: 72°F
1073 ORIGINAL pH: 6.4
5
e 2 .
z \
<
3 0
2 :l\ 1
\ f/2 =1{.7 min
&
o
5 N
@
\
. N\,
1073 AN
> \‘g_
1076
0 12 16 20 24 28
TIME {min)

Fig. 6.19. Typical Stripping Rate for CO, from Water in an MSRE Pump-
Demonstration System.
1k

15.

16.

1T7.
18.
19.
20.

21.

136
REFERENCES

C. F. Weaver et al., Phase Eguilibria in Molten Salt Breeder Reactor Fuels,
ORNL-2896 (1960).

R. E. Thoma (ed.), Phase Diagrams of Nuclear Reactor Materials ORNL-25h8
(Nov. 2, 1959).

C. J. Barton et al., Aircraft Nuclear Propulsion Progect Quar. Prog. Rep.
March 10, 1953, ORNL- -1515, p 112 (1953).

cC. J Barton et al., Aircraft Nuclear Propulsion Project Quar. Prog. Rep.
Sept. 10, 1953, ORNL-1609, p 59-60 (1953).

G. D. Robbins, "Investigation of Phase Equilibria in the System CsF-ZrFg4,"
Intra-Laboratory -Correspondence (Sept. 9, 1960).

Status Report for Chemical Development Sections A and B: Week Ending July 29,
1960, et seq., ORNL CF-60-3-1%40. -

J. H. Burns, unpublished work.

R. E. Thoma, Fused.Salt Phase Equilibria - A Decennial Index to ORNL Progress

- Reports, ORNL CF-60-T7-52 (July 20, 1960).

J. H. Shaffer, G. M. Watson, and W. R. Grimes, ‘The Solubility of Boron Tri-
fluoride in Molten Fluorides, ORNL CF-61-2-35 (Feb. 17, 1961).

-J. H. Shaffer,-w R. Grimes, and G. M. Watson, J. Phys.. Chem. 63 1999 (1959).

J. H. Shaffer and G. M. Watson, Reactor Chem. Div Ann Prog. Rep. Apr. 29,
1960, ORNL-2584, p 31.

H. S. Booth and D R. Martin, Boron Trifluoride and Its Derivatives,
Chap. VII," Wiley, New York (1949).

S.'S. Kirslis, Reactor Chemistry Division, ORNL, privete commmnication,
Februaxy 1961. '

G. M. Wolten, "Solid Phase Transition in the U05-Zr0s Systems;'J Am. Chem.
Soc. 80, 9772-5 (1958). |

I. F. Ferguson, Solid Solutions of Uranium Oxide in Zirconia with the Mono-
clinic ZrO, (Baddeleyite) Structure, AERE-M-694 (May 1960). '

B. C. Blanke et al., MIM~-1076 (April 1956).
B. C. Blanke et al., MIM-1079 (April 1956).

. C. Blanke et al., MIM-1086 (December 1956).

B
J. D. MacKenzie, J. Chem. Phys. 32, 1150 (1956).

F. M. Jaeger, Z..Anorg. Chem. 101, 177, 178 (1917).
H

. Zachariasen, J. Am. Chem. Soc. 70, 2147 (1948).
22,

23.

24,

25.

26.

27.
28.

29.

30.
31.

32,
33.

3k,
| 35.
36.

137

This topic is condensed from ORNL CF-61-2-59, Xenon Diffusion in Graphite,
by R. B. Evans, III, and G. M. Watson (Feb. 15, 1961).

L. G. Overholser and J. P. Blakely, Reactor Chem. Div. Ann. Prog. Rep.
Jan. 31, 1960, ORNL-2931, p 139-148.

W. H. Cook and D. H, Jansen, A Preliminary Summary of Studies of INOR-8,
Inconel-Graphite and Fluoride Systems for the MSRP for the Period from May,
1958 to December 31, 1958, ORNL CF-59-l-4 (Jan. 30, 1959).

R. B. Evans, III, J. Truitt, and G. M. Watson, Interdiffusion of Helium and
Argon in a Large-Pore Graphite, ORNL CF-60-11-102 (Nov. 23, 1960).

E. E. Anderson, G. L. Wessman, and L. R. Zumwalt, Trans. Am. Nuc. Soc. 3(2)
(Dec. 1960), paper 10-k, presented at 1960 winter meeting, San Francisco,
Californis,. ‘ '

M. W. Rosenthal, Nuclear Safety 1, 28 (1960).

M. W. Rosenthal and S. Cantor, Some Remarks on the Contribution of Fission-
Product Cesium to the Pressure Buildup in UOp Fuel Elements, ORNL CF-60-3-81
(Mar. 18, 1960).

U’

W. D. Burch, G. M. Watson, and H. O. Weeren, Xenon Control in Fluid Fuel ..
Reactors, ORNL CF-60-2-2 (July 6, 1960).

I. Spiewak, private commumnication (Oct. 7, 1960).

A. R. Saunders, Retention of Boron Trifluoride in a Molten Salt Reactor,
ORNL CF-60-11-21 (November 1960).

ORSORT Reactor Physics Laboratory Menual, TID-5262 (July 1955).

G. Goldberg, A. S. Meyer, J. C. White, Anal. Chem. 32, 314 (1960).

W. R. Marsh, AERE C/M 377, UKAEA (1955).

5. A. Ryce and W. A. Bryce, Can. J. Chem. 35, 1293 (1957).

J. M. Lafferty, J. Appl. Phys. 22, 299 (1951).

7. ENGINEERING RESEARCH

7.1 PHYSICAL-PROPERTY MEASUREMENTS

T7.1.1 Enthalpy and Heat Capacity

The enthalpy of the MSRE fuel mixture (LiF-Bng—ZrF4-ThF4-UF4, 70-23-5-l-i
mole %) was determined. The data, shown in Fig. 7.1, can be represented by the
following equations:

Hy - Hyy = -0.16 + 0.145 t + (3.63 x 10~%) t2 (1)

for the solid in the temperature range from 100 to 430°C (where the enthalpy,
H, is in cal/g for t in °C), and

Hy - Hyy = -23.4 + 0.572 t - (8.99 x 107°) £2 (2)

for the liquid between 475 and 800°C. The heat of fusion at 450°C was 77 cal/g.

UNCLASSIFIED
ORNL-LR-DWG 57701

400 J)
°
LIQUIDUS
TEMPERATURE—|
o
< 300
L2
o ‘ a),’;/“‘T"
(&)
=2
v
o
3 200 o
>
: :
T %
g ,;/
e %°
Ig 100 ” o
3':\
. P
0 100 200 300 400 500 600 700 800

t ,TEMPERATURE (°C)

Fig. 7.1. Enthalpy-Temperature Relationship for the MSRE Fuel Mixture LiF-Ber-ZrF4-ThF4-UF4
(70-23-5-1-1 mole %).

138
139

The mean value for the heat capacity of the liquid (derived from Eq. 2 in
the temperature range from 550 to 800°C) is 0.451 cal/g-°C. In comparison, the
value predicted using the previously discussed empirical correlation,1

¢, = k.21 /w070, (3)

is 0.458 cal/g.°C; the agreement with c_ based on the experimental results is
to within 2%, | P

Measurements of the enthalpy of the coolant mixture LiF-BeFz (68-32 mole %)
were partially completed. A straight-line correlation of the datea obtained thus
far (Fig. 7.2) yields a value of 0.53 cal/g.°C for the heat capacity of the
liquid between 400 and 650°C. '

UNCLASSIFIED
ORfiL=LA=uws Y1 1UL

400 ’
LIQUIDUS ey
—_ ’ TEMPERATURE
g : of
3 e®
w 300 ®
W Y ~y
= -.r
w
o
o
w
e 0y
o
E ®
J 200 - o
T ®
'_
=
w ®
o
-]
-
®
T 100
i.\
0
0 100 200 300 400 500 600 700 800

¢, TEMPERATURE (°C)

Fig. 7.2. Enthalpy-Temperature Relationship for the Coolant Mixture-LiF-Ber (68-32 mole %).

T.1.2 Viscosity

Preliminary experimental results for the viscosity of the fuel mixture
LiF-BeF2-ZrF4-ThF4-UF4 (70-23-5-1-1 mole %), as determined by efflux-cup meas-
urements,  are shown in Fig. 7,3. For each cup, the curve represents the lower
envelope on the data; experimental scatter of about 40% above these minimum
values was experienced, Postoperational analysis of the melt disclosed an ab-
normal precipitation of ZrQOz; the wide variation in the experimental data may
be associated with this fact.

The kinematic viscosities (u/p) were converted to absolute values by means
of the density data obtained by Kirslis using a buoyancy technique (see Sec.
140
6.3.1); his data can be represented by the equation,

p = 2.84 - 0.00056 t, (4)

with an estimated precision of *0.01 g/cm3 (p in g/cm3 for t in °C). The re-
sults can be described by the following equations:

Cup 51: p = 0.0215 e5a3o/T, (5)
Cup 53: u = 0.0827 e3831/T, - (6)
Cup 56: u = 0.1534 e3598/T, (7)

where p 1is in centipoise and T is in °K.

In view of the uncertainties in these results, the viscosity will be re-
measured in the near future using a new sample of the salt mixture.

UNCLASSIFIED
6 . ORNL-LR-DWG 55799A

\
® CuP 53

¥ cup 51
! amt%l\\\
R

I N
T

p/p, KINEMATIC VISCOSITY (centisickes)

500 600 70C 800
TEMPERATURE (°C)

Fig. 7.3. Preliminary Results on the Viscosity of the MSRE Fuel Mixture
(LiF-Ber-Zer-UF4-ThF4; 70-23-5-1-1 mole %).

T.2 HEAT-TRANSFER STUDIES

The experimental study3 of the heat-transfer coefficients for the salt
mixture LiF-BeFz-UF4-ThFs (67-18.5-0.5-14 mole %) flowing in heated Inconel
and INOR-8 tubes was discontinued, after 7050 hr of exposure, due to an exces-
sive loss of salt in the region of the original break in the heat exchanger.4
Both the Inconel and INOR-8 tubes are being examined metallographically.

The results of this investigation are summarized in Figs. 7.4 and 7.5.
The data are grouped into three series (MA, MB, and MC); the divisions cor-
respond to periods of shutdown for pump repairs during which all thermocouples
were galso replaced. A preliminary analysis of the data shows no significant
Ny, /N, -HEAT TRANSFER PARAMETER

Fig. 7.4. Heat Transfer with the Mixture LiF-BeF,-

141

. UNCLASSIFIED UNCLASSIFIED
ORNL-LR-DWG 57703 : ORNL-LR-DWG 57704
00 ] | 100 1] |
50 [ I 50 I |
0.4 _ 0.8 3 3
Nyo/Np, " = 0.023Ng, ~ Wy /Np 04 = 0.023N,,° 8‘5/
4/// v
60 - x 60 a7
‘ /. ° @ /
rdd o /
2 v
A (e} <
v / v 2 3 . e d
[ hd ]
L ¢ s £ L Cmo & o
[ )
| aege® % ARGt
7 *— g 7 &
s ° : s
o 5 & o
® T @
20 & 20
[ L4
. N .
E) [ A
® SERIES MA = ® SERIES MA
) O SERIES MB o SERES MB
¢ 4 SERIES MC A SERIES MC
o |
5 x 10° 10° 2 4x10" 5x 10 10* 2 4x10°
Nge» REYNOLDS MODULUS : Nge. REYNOLDS MODULUS

UF,-ThF, (67-18.5-0.5-14 mole %) Flowing in a Heated UF,-ThF, (67-18.5-0.5-14 mole %) Flowing in a H
Inconel Tube. . INOR-8 Tube. :

trend as a function of the exposure time and yields, by least-squares calcula-
tion of the lumped Inconel and INOR-8 tube data, the correlating equation:

0.4 v 0.932
NNu/NPr = 0.00484 Np_ , (8)

LU
with a standard deviation of 1.1 units in NNu/Npg ; 75% of the data lie within
30 of Eq. (8). The results for the LiF-BeFz-UF4-ThF4 (67-18.5-0.5-14 mole %)
mixture average, in the range Nge = 10,000 to 20,000, 25% less than the values
calculated from the general heat-transfer correlation for normal fluids,5

0.4 0.8 4 '
NNu/NPr = 0.023 N " (9)

Any conclusions at this date must be tempered, however, by the fact that
there remain a number of uncertainties in the calculated results. First, the
cause of the abnormally poor heat balances is still unresolved; ratios of the
heat gained by the fluid (gp) to the electrical energy input (q.) as high as
1.6 have been noted. In view of this, the factors entering into the calcula-
tion of g are being re-examined with particular emphasis on, the influence of
fluid viscosity on the turbine flowmeter output signal. Second, there is some
doubt as to the values used for the thermal properties entering into the data
analysis. With respect to this, a postoperational examination of the salt cir-
culated indicates a composition which is at variance with both the preopera-
tional constitution and the nowminal values. This is shown in Table 7.1l where=
in the over-all range for three postoperational samples is listed. It is
planned to remeasure the heat capacity and viscosity using the salt removed
from the heat-transfer system. Further, the thermal conductivity will be
experimentally determined; the value used in the present data analysis is

Fig. 7.5. Heat Transfer with the Mixture LiF-BeF,-

eated
derived from an estimate based on a very limited amount of data in systems in

142

which neither LiF nor BeFs> was a constituent.

Table 7.1. Composition of LiF-BeFz-UF4-ThF4 Salt Mixture Samples

Composition (mole %)

Constituent Nominal Preoperation Postoperation
LiF 67.0 70.8 66.3 — 67.2
BeF2 18.5 16.6 15.8 = 15.9
UFq 0.5 0.5 0.6 - 0.7
ThF4 lh:O 12.1 16.5 — 17.3

ppa
Ni 30 20 — 35
Cr 115 380 — 420
Fe 305 155 — 165
Mg * b5 — 745
Mo * <5
*Not measured.
REFERENCEé

MSRP Quar. Prog. Rep. July 31, 1960, ORNL-3014, p 85.

MSRP Quar. Prog. Rep. Jan. 31, 1959, ORNL-2684, p 65-67.

MSRP Quar. Prog. Rep. July 31, 1959, ORNL-2799, p 39; MSRP Quar. Prog. Rep.

Apr. 30, 1960, ORNL-2973, p 27-29; MSRP Quar. Prog. Rep. July 31, 1960,

ORNL-3014, p 86-87.

MSRP Quar. Prog. Rep. Oct. 31, 1959, ORNL-2890, p 23.

W. H. McAdams, Heat Transmission, 3d ed., p 219, McGraw-Hill, New York,

195k.
8. FUEL PROCESSING

8.1 SbFs-HF SOLUTIONS

Work with SbFs-HF solutions was undertaken on the basis of Clifford's datal
that rare-earth fluorides are soluble in these solutions and his statement that
SbFs is the strongest acid in HF, if an acid is defined as a substance that in-
creases the concentratlon of the solvated cation of the solvent (H2F+) More
recent inveetigations® of the species present in SbFs-HF solutions indicated
that there is an assoc1ated un- ionlzed compound, HzF*'-SbFg™, as well as the
ionic species HoF' (analogous to Hz0') and SbFg~. Presumably the rare-earth
fluoride solubility results from the formation of a compound, such as La(SbFg)a,
which is soluble in HF, the rare earth acting as a base, ©Since ThFy4 is less
basic, it is less likely to form such compounds.

In initial tests, HF containing about 20% SbFs reacted with both rare-earth
fluorides and LiF, as indicated by heat evolution. The compound formed with LiF
was not soluble, as observed qualitatively, but that formed with rare-earth fluo-
rides was very soluble. There was no apparent interaction or solubility with
ThF, .

The solid formed by adding liquid SbFs to an HF solution of LiF was not
stoichiometric, the Li/Sb mole ratio typically being around 1.4. This solid
could be recrystallized from anhydrous HF to yield the 1l:1 LiSbFe, with a chemi-
cal analysis of 50.4% Sb, 2.67% Li, and 46.5% F and an L1/Sb mole ratio of 0.93
and F/(Li + 55b) ratio of 0.99. The mole ratio of 0.93 was quite consistent in
several samples that were analyzed more than once. This deviation from stoi-
chiometry may have fesulted from the 3% analytigal error inherent in the method,
or possibly some Li sites were replaced by Ha0 from water impurity. The
LiSbFg gave a good x-ray diffraction pattern, which persisted as the hygroscopic
solid absorbed water until it was nearly completely in solution. The compound
was somewhat soluble in HF, 0.23 mole or 1.6 g of lithium per liter at the
boiling point and less at lower temperatures, but was quite soluble in water.
The compound was purified by dissolving it in ~200 ml of anhydrous HF and either
cooling the saturated solution in dry ice to precipitate a fine powdery solld or
evaporating it slowly to about 10 to 20% of the original volume to produce
crystals. In some cases evaporation yielded excellent- crystals, up to &bout 2
mm on an edge with sharp edges and corners.

X-ray diffraction showed the LiSbFg crystals to be of monoclinic symmetry
with the following cell dimensions: 3

7.50 &
93.6°

a = 5.42 2 c
b =5.18 R B

Systematic extinction indicated the space group Cs®-Im. The theoretical density

is 3.83 g/cc. Structures for the analogous T1l, NHg, Rb, and Cs compounds (ref &)
and for NaSbFg and NaSb(OH)g (ref 5) have been published, but there is indication
that some of these may have contained some water.

143
144

The thermal stability of the LiSbFg compound was investigated qualitatively
by heating the material in an inert atmosphere in nickel or copper containers.
Decomposition, indicated by the appearance of a white smoke when the vapor con-
tacted moist air, was observed at temperatures from 400 to 650°C, but the vola-
tile material was SbFs instead of SbFs. The compound apparently reacted with
the container, forming SbFs and NiF, or CuF>, and then the SbFz vaporized.
Decomposition occurred at lower temperatures in copper than in nickel. In some
experiments in nickel, decomposition was incomplete at temperatures as high as
700°C, and the compound may be stable to higher temperatures if reaction with
the container can be prevented. The only pattern showed by one sample examined
by x-ray diffraction after being heated to 600°C in nickel was that of the
LiSbFg compound. In copper containers there was some evidence of formation of
a melt between LiF, CuF,, and SbFa at temperatures below 600°C.

Attempts to prepare the LiF-SbFs compound by reaction of solid LiF with
SbFs vapor were unsuccessful, although some reaction probably occurred. The
SbFs was carried by argon gas that bubbled through liquid SbFs at temperatures
between 25 and 60°C, and the LiF powder was spread in an Inconel or nickel boat
in a tube held at or sbove 100°C. Quantitative results could not be cbtained
because the metal surfaces became coated with a viscous liquid layer. However,
the boat containing LiF gained many times more weight than an empty boat. LiF
supported on a suitable porous carrier might absorb Sbls, and perhaps some
similar contaminants from gas streams, and thus might have application to the
fluoride volatility process as.well as to MSR fuel processing.

Attempts to purify the rare-earth fluoride - SbFs compound by recrystal-
lication were generally unsuccessful, since it is. very soluble;. evaporation of
the solutions indicated that -the solubility was at least in the range 200 to
300 g of rare earths per liter. The solid obtained by evaporation or cooling
of the saturated solution was a very fine particle-sized material associated.
with a quantity of the solution. Analysis of one sample -showed a rare earth/Sb-
retio of 2/9, but this solid was obtained from a solution containing excess .SbFs..

Since SbFs-HF solutions dissolve rare-earth fluorides but. not ThF4, an
experiment was carried out to- investigate the possibility that this reagent could
decontaminate ThFs from MSBR blanket salt. A salt spiked with Cs, Ce, and Pm
activities was first leached with 95% HF - 5% water to dissolve the LiF and per-
haps BeF,, rinsed with HF, and then leached twice with 5 vol % of SbFs in HF,
Analytical results have not yet been obtained.

8.2 DISSOLUTION RATES FOR MSRE FUEL CARRIER SALT

Dissolution rates of carrier salt, 63-37 mole % LiF-BeF,, in stirred,
boiling 90% HF - 10% H-0 were fast enough that it should not be necessary to
introduce the salt into the system in very fine particles, thus eliminating
atomizing-nozzle corrosion problems. In two experiments the dissolution rate
was ~50 mils/hr in the first 10 min, decreasing to <10 mils/hr after some
time, probably because of a layer of soft material that was left on the surface.
This material had the general properties of BeFy and is probably the BeF, remain-
ing after LiF, which dissolves more rapidly, is leached from the salt. When the
salt specimen was bounced around with a magnetic stirrer, it dissolved very
rapidly, ~200 mils/hr, probably due to mechanical removal of the soft surface
layer. The resulting solution was clear without any fine particles. The salt
samples dissolved were capsule shaped, 270 to 290 mils in diameter by about T0O
mils long; they had been prepared by casting in a split graphite mold.
145

8.3 STEADY-STATE FLUORINATOR

In four runs in a small steady-state fluorinator uranium was 97 to 98% re-
covered consistently by continuous fluorination of a fused salt. Corrosion of
the nickel container was severe in the preliminary equipment used; better equip-
ment design and better conditions should give much less corrosion, though per-
haps poorer uranium separation. The fluorinator, which normally contained about
T0 ml of salt, was a l-in.-diam nickel tube from which salt overflowed through
a l/h-in.-diam tube into a jackleg that held up enough salt to permit sampling.

The salt flow rate was controlled by pressure transfer of molten salt from
a reservoir through a coiled 6-ft by 4O-mil capillary tube. Flows were 5 g/min
with 20 psi transfer pressure, corresponding to an average salt residence time
in the fluorinator of about 40 min. This method of transferring molten salts at
low flow rates was satisfactory, but practical application would require opera-
tion with shorter residence times, and therefore higher salt transfer rates,
than .could be obtained with this apparatus.

The main operational problem was that the salt failed to flow freely from
the Jjackleg and thus backed up in the fluorinator. This difficulty was probably
caused by the salt drain line being too small (1/8-in. tube) and extending into
a, colder zone of the furnace. It was solved by operating at temperatures as
high as 700°C. The high .NiF, content of the salt, resulting from excessive -cor-
rosion, may have exceeded the NiF, solubility, thus plugging the drain line,
with more NiF, being formed at the higher operating temperature requlred to
maintain the drain line open.

The two most meaningful runs were made with LiF-NaF-ZrF, (26-37-37 mole %)
salt containing 1 wt % UF4 in one case and 0.2 wt % in the other. For the first
case, fluorination near TOO°C with excess fluorine and with an average residence
time for salt in the fluorinator of about 40 min resulted in a uranium content
in the effluent salt of 100 to 200 ppm, 1 to 2% of the original concentration.
In the other experiment, with the temperature in the range 600 to 650°C, the
uranium content of the overflow salt was 25 to 50 ppm, again 1 to 2% of the feed
concentration. Preferred conditions would include a lower temperature, about
550°C or less, and a shorter salt residence time. In general, the nickel con-
tent in the salt increased throughout a run,. never reaching a steady state.
Corrosion rates based on nickel concentrations were of the order of 1 mil/hr at
T700°C and 0.1 to 0.2 mil/hr at 610°C.

REFERENCES

1. A. F. Clifford and Kongpricha, J. Inorg. Nucl. Chem. 5, 76 (1957).

. H. Hyman et al., J. Phys. Chem. 65, 123 (1961).

. H. Burns, Reactor Chemistry Division, ORNL, private commmnication (1961)

+ W ™
2 49 o

. Schrewelius, Arkivrfoer Kemi, Min. och Geol. 16B, No. 7 (1943).

5. N. Schrewelius, Z. anorg. allgem. Chem. 238, 241 (1938).

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