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W NS e, anememie Sty SaP - e T el e
-
3 4456 03L0L0OS8 3
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ORNL-4782
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Pl \fl
MOLTEN-SALT REACTOR
PROGRAM
Semiannual Progress Repott
Period gnding Cerbhuahg 29, 1972
OAK RIDGE NATIONAL LABORATORY .
CENTRAL RESEARCH LIBRARY
DOCUMENT COLLECTION
LIBRARY LOAN COPY
DO NOT TRANSFER TO ANOTHER PERSON
If you wish someone else to see this
" document, send in name with document
and the library will arrange a loan.
LUCN-7969
i3 3-67,
L
OAK RIDGE NATIONAL LABORATORY
OPERATED BY UNION CARBIDE CORPORATION e FOR THE U.S. ATOMIC ENERGY COMMISSION
Printed in the United States of America. Available from
National Technical Information Service
U.S. Department of Commerce
5285 Port Royal Road, Springfield, Virginia 22151
Price: Printed Copy $3.00; Microfiche $0.95
This report was prepared as an account of work sponsored by the United
States Government. Neither the United States nor the United States Atomic
Energy Commission, nor any of their employees, nor any of their contractors,
subcontractors, or their employees, makes any warranty, express or implied, or
assumes any legal 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,
ORNL-4782
UC-80 — Reactor Technology
Contract No. W-7405-eng-26
MOLTEN-SALT REACTOR PROGRAM
SEMIANNUAL PROGRESS REPORT
For Period Ending February 29, 1972
M. W. Rosenthal, Program Director
R. B. Briggs, Associate Director
P. N. Haubenreich, Associate Director
OCTOBER 1972
s T
for the 3 quE D
US. ATOMIC ENERGY COMMISSION 360108 3
ORNL-2474
ORNL-2626
ORNL-2684
ORNL-2723
ORNL-2799
ORNL-2890
ORNL-2973
ORNL-3014
ORNL-3122
ORNL-3215
ORNL-3282
ORNL-3369
ORNL-3419
ORNL-3529
ORNL-3626
ORNL-3708
ORNL-3812
ORNL-3872
ORNL-3936
ORNL4037
ORNL4119
ORNL-4191
ORNL4254
ORNL4344
ORNL-4396
ORNL4449
ORNL-4548
ORNL4622
ORNL4676
ORNL-4728
This report is one of a series of periodic reports in which we describe the progress of the program. Other reports
issued in this series are listed below.
Period Ending January 31,1958
Period Ending October 31, 1958
Period Ending January 31, 1959
Period Ending April 30, 1959
Period Ending July 31, 1959
Period Ending October 31, 1959
Periods Ending January 31 and April 30, 1960
Period Ending July 31, 1960
Period Ending February 28, 1961
Period Ending August 31, 1961
Period Ending February 28, 1962
Period Ending August 31, 1962
Period Ending January 31, 1963
Period Ending July 31,1963
Period Ending January 31, 1964
Period Ending July 31, 1964
Period Ending February 28, 1965
Period Ending August 31, 1965
Period Ending February 28, 1966
Period Ending August 31, 1966
Period Ending February 28, 1967
Period Ending August 31, 1967
Period Ending February 29, 1968
Period Ending August 31, 1968
Period Ending February 28, 1969
Period Ending August 31, 1969
Period Ending February 28, 1970
Period Ending August 31, 1970
Period Ending February 28, 1971
Period Ending August 31, 1971
Contents
PART 1. MSBR DESIGN AND DEVELOPMENT
L. DESIGN . ottt e e e e e e e 2
1.1 Molten-Salt Demonstration Reactor Design Study ........ ... ... i 2
0 I 7= 1 =3 O PR 2
1.1.2 ReaCtOT COTC . ittt ittt e e e e e e e e e e e e e e e s 2
1.1.3 Graphite Temperatures . .. ... ..ottt e 6
1.1.4 Cell COOlNG . .ottt it et it e et et et e 8
1.2 Side-Stream Processing of MSBR Primary Flow for lodine Removal ......................... 8
1.3 MSBR Industrial Design Study .. ... ...t 9
1.4 MSBE DESIgN ..ottt ittt e et e e 12
1.5 Bubble Behavior in the MSBR Primary Salt System ... ... ... . . .. . i i 13
2. REACTOR PHY SICS . .ttt ettt e e e e e e e ettt e e 18
2.1 Experimental PhySics ... . ... ..ottt 18
2.1.1 HTLTR Lattice Experiments ......... ... .. oot 18
2.2 Physics Analysis Of MSBR . ... ... . 21
2.2.1 Radiation Heatingin MSR Pumps . ... ... ... i 21
2.2.2. MSBE Control-Rod Worths .. ... e e 22
2.2.3 Molten-Salt Converter Reactors Using Plutonium . .. .......... ... i, 23
3. SYSTEMS AND COMPONENTS DEVELOPMENT ... ... . i e 28
3.1 Gaseous Fission Product Removal . .. ... ... . . e 28
3.1.1 Bubble Separator and Bubble Generator .............. ... ... 28
3.1.2 Bubble Formation and Coalescence Test .. .. ... ... ..o .. 29
3.1.3 Bubble Separator ANalyses .. .. ... ...ttt 29
3.2 Gas System Technology Facility ......... ... .. i 32
3.3 Molten-Salt Steam Generator Industrial Program . ........ ... . . . . . ... i il 33
3.4 Coolant-Salt Technology Facility . ... ... ... i 33
3.5 SAlt PUMPS . .ottt ettt e ... 34
3.5.1 Salt Pumps for MSRP Technology Facilities . . ............. ... oo, 34
3.5.2 ALPHA PUMD .. oottt ettt e et e e e e e 35
3.5.3 Molten-Salt Mixer, Laboratory Scale . ........ .. ... . i 37
4. INSTRUMENTATION AND CONTROLS . .ttt i e e e e e e 38
4.1 Transient and Control Studies of the MSBR System Using a Hybrid Computer ................. 38
iii
iv
5. HEAT AND MASS TRANSFER AND PHYSICAL PROPERTIES . ....... ... ... ..., 39
5.1 Heat Transfer . ... e e e e e e e e 39
5.2 Wetting Studies . ..... .. i e e 40
5.3 Mass Transfer to Circulating Bubbles .. .. .. ... .. . . e 41
PART 2. CHEMISTRY
6. FISSION PRODUCT BEHAVIOR . . .. o e e e it eane s 45
6.1 Some Factors Affecting the Deposition Intensity of Noble-Metal Fission Products .............. 45
6.2 Effects of Selected Fission Products on Hastelloy N, Nickel,
and Type 304L Stainless Steel at 650°C . . ..o\ttt i ittt e 50
6.3 Reaction of CoF3 with Tellurium . . . ... oo e e 51
7. BEHAVIOR OF HYDROGEN AND ITS ISOTOPES . .. ... . e 53
7.1 Solubility of Hydrogenin Molten Salt ... ... ... ... i i i s, 53
7.2 Initial Tritium Chemistry in the Core of a Molten-Salt Reactor .......... ... ... .. ... .. .... 53
7.3 Permeation of Hydrogen Through Metals at Low Pressures ........... ... ... ... .. .. ... 54
7.4 Influence of Films or Coatings on Hydrogen Permeation Rates ,........... ... ... ... ..... 56
7.5 Experiments on Hydrogen Evolution from Fluoroborate Coolant Salt . ............. ... ... .. 57
7.6 Apparatus for Infrared Spectral Studies of Molten Salts . .......... .. .. .. ... ... ... 59
7.7 Infrared Spectral Studies of the Chemical Behavior of BF;OH ™ and
BF;0D " Ionsin Molten NaF-NaBF, ... ... ... . i 59
7.8 Thermal Stability of NaNO3, KNO;, NaNO, ,and RITEC ... ......... ... ... .. ... ... .. .. 62
8. FLUOROBORATE CHEMISTRY . ..t it et e e e e et et e e 63
8.1 Solubility of BF3 in Fluoride Melts ... ... .. e 63
8.1.1 Reactor Applications .. .... ... ...t e e i 63
8.2 Free Energies of Formation of NaFeF; and NaNiF;; Their Relationship
to the Corrosion of Hastelloy N by Fluoroborates .. ........ ... ... ... 65
8.3 Preparation of Fused Sodium Fluoroborate for the Coolant Salt Technology Facility ............ 68
0. PROTACTINIUM CHEMIS T RY ... e e e e e 70
9.1 Oxide Chemistry of Protactinium in MSBR Fuel Salt . .. . ... ... ... .. ... ... . ... o .. 70
9.2 Binary Solid Solutions of PaO, and Other Actinide Dioxides and Their
Exchange Equilibria with Molten-Salt Reactor Fluorides .. ... ... ... ... ... .. .. .. .. ... ... 72
9.3 Termary Solid Solutions of ThO,,Pa0,,and UQ, ... .. ... . i 74
10. DEVELOPMENT AND EVALUATION OF ANALYTICAL METHODS
FOR MOLTEN-SALT REACTORS . .. ..o e e e e e e e e 77
10.1 In-Line Chemical Analysis of Molten Fluoride Salt Streams .. ...... ... ... ... .. ... ... ... 77
10.2 Theoretical Considerations of the Voltammetric In-Line Determination
of Uranium(III) . .. ... e 80
10.3 Electroanalytical Studies of Titanium(IV) in Molten LiF-BeF, -Z1F,
(65.4-29.6-5.0 MOIE 20) -« oottt e 81
11.
10.4 Electrochemical Studies of Bismuth(III) in Molten LiF-BeF,-ZrF, at 500°C .................. 82
10.5 Voltammetry of Chromium(III) in Molten NaBF,-NaF (92-8 Mole %) .......... ... ... ...... 82
10.6 Voltammetric and Hydrolysis Studies of Protonated Species in Molten NaBF, . ................ 83
10.7 Determination of Hydrogen in NaF-NaBF, Salts ... ...... .. .. ... . . . . L. 83
10.8 Spectral Studies of Molten Salts . .. .. ... .. 84
OTHER MOLTEN-SALT RESEARCHES . ... .. e e e 87
11.1 The Oxide Chemistry of Niobium in Molten LiF-BeF, Mixtures ............... ... ... ..... 87
11.1.1 Equilibrations of Nb, Os and BeO with Molten LiF-BeF, Mixtures . ................... 87
11.1.2 Equilibrations Involving Nickel Niobates in Molten Li,BeF, .......... ... ... .. ... ... 89
11.2 The Reaction of MoF¢ with Niobium ... ... ... . e 91
11.3 Thermodynamics of LiF-BeF, Mixtures .. ... ... ... . .. . . i 95
11.4 Electrochemical Mass Transport in Molten Beryllium Fluoride—Alkali
Fluoride MiXIUTES . . . ..ottt e e e et e e et e e e e e e e e 96
11.5 Electrical Conductance in Beryllium Fluoride Rich NaF-BeF, Mixtures ...................... 98
11.6 The Disproportionation Equilibrium of UF; Solutions . ............ ... ... ... ... .. ..., 98
11.7 The Raman Spectra of Be, F,> and Higher Polymers of Beryllium
Fluorides in the Crystalline and Molten State ... ... ... .. . . . . . . i 100
11.8 Raman Spectra of Molten and Crystalline Potassium Dichromate ........................... 103
11.9 Nonideality of Mixing in the Systems Li, BeF,-Lil, Na, BeF,-Nal,
and C82 BeF4 B 106
PART 3. MATERIALS DEVELOPMENT
INTERGRANULAR CRACKING OF STRUCTURAL MATERIALS
EXPOSED TO FUEL SALT .. .. e e e e e e e e et et i 109
12.1 Examination of Hastelloy N Components fromthe MSRE .. ... . ... ... ... .. .. ... .. .. 109
12.1.1 Freeze Valve 105 . ... .. ot e e et e 109
12.1.2 Control Rod Thimble . . ... ... e e e e 111
12.1.3 Sampler Cage Rod ... ... e e 115
12.1.4 Mist Shield . . .. ... e 116
12.2 Auger Analysis of the Surface Layers on Graphite from the Core of the MSRE .. ............... 117
12.3 Auger Analysis of the Surface of a Fractured Hastelloy NSample . .......................... 122
12.4 Intergranular Corrosion of Hastelloy N . ... ... .. . . 123
12.5 Tube-Burst EXperiments . .. .. ..ot 126
12.6 Cracking of Samples Electroplated with Tellurium . ....... ... .. ... ... ... ... oLt 128
12.7 Cracking of Hastelloy N Being Creep Tested in Tellurium Vapor . ......... .. ... ... ... ..., 128
12.8 Intergranular Cracking of Materials Exposed to Sulfur and Several
Fission Product Elements .. .. ... o e et e e e 134
12.9 Mechanical Properties of Hastelloy N Modified with Several Elements .. ...................... 136
12.10 Status of Intergranular Cracking Studies . ......... ... . . . . . i 141
vi
13. GRAPHITESTUDIES . .......... P 144
INtrOdUCHION . . .. e 144
13.1 Graphite Development . ... . e e 144
13.2 Procurement of Various Grades of Carbon and Graphite .................... ... .. ........ 149
13.3 Texture Determinations ... ... ...ttt ittt et it e 149
13.4 Thermal Property Testing . .. ... ...ttt e it 150
13.5 Nominal Helium Permeability Parameters for Various Grades of Graphite ..................... 151
13.6 Reduction of Helium Permeability of Graphite by Pyrolitic Carbon Sealing ................... 151
13.7 Characterization of Pyrocarbon Sealants for Graphite Using Reflected Light
and Scanning Electron MiCrosCopes .. . ... ...ttt ettt 155
14, HASTELLOY N ..o e e e e et e e e e e e et et e et 161
14.1 Development of a Titanium-Modified Hastelloy N . . ... ... ... . o o i i 161
14.2 Alloys with Exceptional Strength . .. ... .. .. . e 163
14.3 Weldability of Commercial Alloys of Modified Hastelloy N . ......... ... .. ... .. ... .. ... .... 166
14.4 Electron Microscope StUudies ... ... ... ...ttt i e e 171
14.4.1 Intermetallic Precipitation in Hastelloy N .. ... ... ... .. ... .. o L. 172
14.4.2 Precipitation in New Commercial Alloys . .... ... ... .. i 173
14.4.3 Modification of X-Ray Diffraction Techniques ............. ... ... ... ............ 173
14.5 Salt Corrosion StudIes . .. ... ..ottt i e e e e e 174
14.5.1 Fuel Salt ... . . i i ittt e e e i e et e 174
14.5.2 Fertile-Fissile Salt . . ... ... ... e e 177
14.5.3 Blanket Salt .. ... ... . e 177
14.5.4 Coolant Salt ... .. ...t e e e 177
14.6 Forced-Convection Loop Corrosion Studies ... ... ...ttt i e e 179
14.6.1 Operation of Loop MSR-FCL-1A . .. ... ... it 179
14.6.2 Results from Loop MSR-FCL-1A . ... .. e et icee e 179
14.6.3 Operation of Loop MSR-FCL-2 ... ... .. . e 180
14.6.4 Results from Loop MSR-FCL-2 . ... ... e 180
14.7 Corrosion of Hastelloy Nin Steam . . ... ... . i ettt e e eee e 182
14.8 Evaluation of Duplex Tubing for Use in Steam Generators ............... ... ... ..., 189
15. SUPPORT FOR CHEMICAL PROCESSING .. ... . it e e et et e e 192
15.1 Construction of a Molybdenum Reductive-Extraction Test Stand . .......................... 192
15.2 Fabrication Development of Molybdenum Components .......... ... ... ... . ... ..., 194
15.3 Welding of Molybdenum ... ... . e e e 195
15.4 Development of Brazing Techniques for Fabricating the Molybdenum Test Loop ............... 195
15.5 Compatibility of Materials with Bismuth . ....... ... ... . ... ... ... ... . . ... 197
15.5.1 Tantalum and T-1 10 ... o i i ittt e et ettt e 197
15.5.2 Graphite . ... ... o e 198
15.5.3 Tungsten-Coated Hastelloy N . . ... ... . e e 199
15.5.4 Molybdenum ... ... ... e e 199
15.6 Molybdenum Braze Alloy Compatibility ...... ... . ... .. . .. 202
16.
17.
18.
19.
vii
PART 4. MOLTEN-SALT PROCESSING AND PREPARATION
FLOWSHEET ANALY SIS ..o e e e e e e e e e e e e e e
16.1 Design Study and Cost Estimates of a Processing Plant for a
1000-MW(e) MSBR
16.2 Multiregion Code for MSBR Processing Plant Flowsheet Calculations
..............................................................
........................
PROCESSING CHEMISTRY . . ..ot e e e e e e e e e e et
17.1 Distribution of Lithium and Bismuth between Liquid Lithium-Bismuth Alloys
and Molten LiCl .. . ... e
17.2 Solubility of Europium in Liquid Bismuth . .. ... .. . . .
17.3 Integral Heats of Lithium-Bismuth Solutions .. ........ ... .. . . . . . . . . ..
17.4 Protactinium Oxide Precipitation Studies . . .. ... ... ... .
17.5 Chemistry of Fuel Reconstitution .. ...... ... ... i e
ENGINEERING DEVELOPMENT OF PROCESSING OPERATIONS .. ... ... ... ... .. .. ..
18.1 Lithium Transfer during Metal Transfer Experiment MTE-2 ...... ... ... ... ... ... ... .. ..
18.2 Operation of Metal Transfer Experiment MTE-2B . ...... ... ... ... ... ... .. .. ... ... .....
18.3 Installation, Testing, and Charging of Materials to the Third
Metal Transfer Experiment . ... ... .. . i i i e
18.4 Design of the Metal Transfer Process Facility . ......... ... ... ... . .. .. . . ...
18.5 Development of Mechanically Agitated Salt-Metal Contactors ..............................
18.6 Reductive Extraction Engineering Studies . .......... .. ... . .
18.7 Design of the Reductive-Extraction Process Facility ............ ... ... ... ... ... ... ...
18.8 Frozen-Wall Fluorinator Development .. ... .. ... . . . i
18.9 Engineering Studies of Uranium Oxide Precipitation .......... ... ... .. ... ... ... ...
18.10 Design of a Processing Materials Test Stand and the Molybdenum Reductive
Extraction Equipment ... ... . e
18.11 Development of a Bismuth-Salt Interface Detector .......... .. ... . ... .. . ... . ...
CONTINUOUS SALT PURIFICATION . ... e e it e
Introduction
The objective of the Molten-Salt Reactor Program is
the development of nuclear reactors which use fluid
fuels that are solutions of fissile and fertile materials in
suitable carrier salts. The program is an outgrowth of
the effort begun over 20 years ago in the Aircraft
Nuclear Propulsion program to make a molten-salt
reactor power plant for aircraft. A molten-salt reactor —
the Aircraft Reactor Experiment — was operated at
ORNL in 1954 as part of the ANP program.
Our major goal now is.to achieve a thermal breeder
reactor that will produce power at low cost while
simultaneously conserving and extending the nation’s
fuel resources. Fuel for this type of reactor would be
233UF, dissolved in a salt that is a mixture of LiF and
BeF,, but 235U or plutonium could be used for
startup. The fertile material would be ThF, dissolved
in the same salt or in a separate blanket salt of similar
composition. The technology being developed for the
breeder is also applicable to high-performance converter
reactors.
A major program activity through 1969 was the
operation of the Molten-Salt Reactor Experiment. This
reactor was built to test the types of fuels and materials
that would be used in thermal breeder and converter
reactors and to provide experience with operation and
maintenance. The MSRE operated at 1200°F and
produced 7.3 MW of heat. The initial fuel contained 0.9
mole % UF,, 5% Z1F,4, 29% BeF,, and 65% " LiF; the
uranium was about 33% ?°°U. The fuel circulated
through a reactor vessel and an external pump and heat
exchange system. Heat produced in the reactor was
transferred to a coolant salt, and the coolant salt was
pumped through a radiator to dissipate the heat to the
atmosphere. All this equipment was constructed of
Hastelloy N, a nickel-molybdenum-iron-chromium
alloy. The reactor core contained an assembly of
graphite moderator bars that were in direct contact
with the fuel.
iX
Design of the MSRE started in 1960, fabrication of
equipment began in 1962, and the reactor was taken
critical on June 1, 1965. Operation at low power began
in January 1966, and sustained power operation was
begun in December. One run continued for six months,
until terminated on schedule in March 1968.
Completion of this six-month run brought to a close
the first phase of MSRE operation, in which the
objective was to demonstrate on a small scale the
attractive features and technical feasibility of these
systems for civilian power reactors. We concluded that
this objective had been achieved and that the MSRE
had shown that molten-fluoride reactors can be oper-
ated at 1200°F without corrosive attack on either the
metal or graphite parts of the system, the fuel is stable,
reactor equipment can operate satisfactorily at these
conditions, xenon can be removed rapidly from molten
salts, and, when necessary, the radioactive equipment
can be repaired or replaced.
The second phase of MSRE operation began in
August 1968, when a small facility in the MSRE
building was used to remove the original uranium
charge from the fuel salt by treatment with gaseous F,.
In six days of fluorination, 221 kg of uranium was
removed from the molten salt and loaded onto ab-
sorbers filled with sodium fluoride pellets. The decon-
tamination and recovery of the uranium were very
good.
After the fuel was processed, a charge of 2?3 U was
aaded to the original carrier salt, and in October 1968
the MSRE became the world’s first reactor to operate
on 233U. The nuclear characteristics of the MSRE with
the 23U were close to the predictions, and the reactor
was quite stable.
In September 1969, small amounts of PuF; were
added to the fuel to obtain some experience with
plutonium in a molten-salt reactor. The MSRE was shut
down permanently December 12, 1969, so that the
funds supporting its operation could be used elsewhere
in the research and development program.
Most of the Molten-Salt Reactor Program is now
devoted to the technology needed for future molten-
salt reactors. The program includes conceptual design
studies and work on materials, the chemistry of fuel
and coolant salts, fission product behavior, processing
methods, and the development of components and
systems.
Because of limitations on the chemical processing
methods available at the time, until three years ago
most of our work on breeder reactors was aimed at
two-fluid systems in which graphite tubes would be
used to separate uranium-bearing fuel salts from
thorium-bearing fertile salts. In late 1967, however, a
one-fluid breeder became feasible because of the devel-
opment of processes that use liquid bismuth to isolate
protactinium and remove rare earths from a salt that
also contains thorium. Our studies showed that a
one-fluid breeder based on these processes can have fuel
utilization characteristics approaching those of our
two-fluid designs. Since the graphite serves only as
moderator, the one-fluid reactor is more nearly a
scaleup of the MSRE. These advantages caused us to
change the emphasis of our program from the two-fluid
to the one-fluid breeder; most of our design and
development effort is now directed to the one-fluid
system.
Summary
PART 1. MSBR DESIGN AND DEVELOPMENT
1. Design
The design study of the 300-MW(e) demonstration
reactor plant was completed. The internal structure of
the reactor was developed in considerable detail. Special
attention was given to methods for fabrication and
assembly of the different graphite pieces. Hydraulic
conditions in the reactor were analyzed and tempera-
tures of graphite were calculated. Cooling equipment
was added to the cell atmosphere circulation systems
that had been provided for heating the reactor equip-
ment cells.
A new analysis of some earlier experiments on the
stripping of iodine from LiF-BeF, melts has shown
that, if one includes the diffusion of I~ in the melt to
the gas-liquid interface as a rate limiting step, there
results a mathematical model which is in accord with
the experimental data. This model permits making an
estimate of what it would take to continually remove
iodine from the MSBR fuel by side-stream processing.
Design studies on the MSBE core were continued,
with emphasis on the design of a slab-type graphite
element. Layout studies of the core include provisions
for four cruciform-shaped control rods.
A computer program (BUBBLE) has been written to
describe in detail the behavior of gas bubbles circulating
with the salt through the MSBR primary system. This
program is being incorporated into an overall program
to describe the detailed behavior of noble gases in an
MSBR.
The MSBR Industrial Design Study Team under
Ebasco Services completed the design report on their
selected 1000-MW(e) reference concept during this
period.
2. Reactor Physics
MSBR lattice physics experiments, performed for us
at Battelle Northwest Laboratories, have been com-
X1
pleted and reported. In the course of preparing for a
careful analysis of these experiments, we have reviewed
and revised our cross-section library. The revised data
for most nuclides are based on ENDF/B version II.
Exceptions are carbon, lithium, and fluorine, for which
other data are justified. In connection with the lattice
experiments, the problem of calculating neutron reac-
tion rates in a doubly heterogeneous system (coated-
particle fuels in a lumped fuel rod) was studied in
detail. For resonance neutron absorption, conventional
methods of analysis contain conceptual errors, at least
for a laminar system. For random spherical grains, the
conclusion with respect to validity of conventional
methods is not clear, but we found that the grain
effects could be neglected for the small grains used in
our experiment. Qur cross-section preparation code,
XSDRN, was modified to treat the grain effects
explicitly in the thermal neutron range.
In connection with the preparation of specifications
for MSR pumps, we calculated the amount of energy
deposited in various parts of a typical MSR pump by
beta and gamma rays from fission products in the salt
and in the cover gas, as well as by neutrons and gammas
from other parts of the primary salt loop.
Reactivity worths of proposed control rods for the
MSBE were calculated. Graphite salt-displacement rods
and poison rods of Hastelloy N, boron in graphite, and
europium in graphite were studied.
In further studies of MSR operation with limited fuel
processing, we calculated initial critical fuel loadings
and fuel compositions as functions of time for various
thorium concentrations with plutonium from light-
water reactors as initial fuel loading and feed. Because
of the high effective absorption cross section in 2*°Pu,
it seems desirable to use a low thorium concentration,
for example, 6 mole %, during the initial batch cycle,
when the plutonium concentration is highest, Subse-
quent cycles would use higher thorium concentrations,
for example, 10 mole %, and would burn, primarily,
bred 233U.
3. Systems and Components Development
The performance and pressure-drop testing of the
GSTF (gas system technology facility) bubble separator
design have been completed on the water test loop. The
final design has a 44-in. separation length, a tapered
casing, and gas removal from both the swirl and
recovery hubs. This design has a suitably high removal
efficiency of the small-diameter bubbles associated with
the 31% CaCl, test fluid and has a stable vortex under
all normal operating conditions.
The pressure-drop measurements of the bubble gener-
ator indicated a larger than expected increase in the
required gas supply pressure as the gas flow was
increased to the design value. Efforts to reduce this
pressure are in progress.
The bubble formation and coalescence test rig was
assembled and operated. The test results show that
bubbles present in 66-34% LiF-BeF, immediately after
agitation are larger than in the 31% CaCl, solution but
are smaller than in demineralized water. A test capsule
of 72-16-12% LiF-BeF,-ThF,; MSR fuel salt contained
suspended material that made observation of bubbles
impossible.
Analyses were started with the objective of idealizing
the swirl-flow bubble separator to develop expressions
which would be helpful in understanding the perfor-
mance of this separator. The effect on removal effi-
ciency of the bubble size distribution and the turbulent
diffusion of bubbles away from the vortex cavity are
included in the studies.
The design of most of the components for the GSTF
is nearing completion, and the detailed design of the
facility piping was started. It was determined that the
loop is too closely coupled to permit accurate predic-
tions of the pressure distributions and that variable flow
restrictors would be required to properly balance the
pressures. The loop will be operated initially with water
to permit calibration of these flow restrictors and to
obtain confirmation of the performance of the modi-
fied salt pump. The preliminary system design descrip-
tion for the GSTF is ready for publication.
The subcontract with Foster Wheeler Corporation for
a four-task conceptual design study of molten-salt
steam generators was approved near the end of the
period. After the completion of a surface arrangement
study, Foster Wheeler will proceed with the conceptual
design analysis on the unit of their selection. Task I is
scheduled for completion in October 1972.
The mechanical design of the coolant-salt technology
facility is complete, and the instrument and contro!
design is 90% complete. The fabrication and installation
xii
of the mechanical components are nearing completion,
and the installation of the instruments and electrical
control equipment is imminent. Heat transfer tests, run
with water, on the concentric tube economizer for the
corrosion product cold trap indicate that the tempera-
ture of the salt supply to the cold trap can be lowered a
satisfactory 200°F when the cold trap is operated at a
maximum AT of 100°F.
Detailed assembly procedures were prepared for the
salt pumps to be used in the coolant-salt technology
facility and the gas system test facility. The refurbish-
ment of these pumps is under way. Plans were made to
perform water tests with the salt pump in the gas
system test facility.
The ALPHA pump, equipped with Viton elastomeric
seals in its lower shaft seal, has operated satisfactorily in
the MSR-FCL-2 for approximately 3900 hr. Data on
pump coastdown and lubricating oil flows and tempera-
tures were taken for use in setting operating set points.
The design of several improvements was completed.
The layout and design of a small molten-salt mixer for
laboratory use were completed.
4. Instrumentation and Controls
The hybrid computer model of the MSBR system has
been used to investigate thermal transients in the salt
systems for representative perturbations. A report
describing the model and some typical results is being
published.
5. Heat and Mass Transfer and Physical Properties
Heat transfer. Additional determinations of heat
transfer coefficient for a proposed MSBR fuel salt have
been made in the transitional range of Reynolds
modulus (3000—4000) at low inlet fluid temperatures
for which the temperature coefficient of viscosity is a
large negative value [-0.2 b ft! hr' (°F)7]. In
particular, for operation at Reynolds modulus 3161
without an adiabatic entrance length, the slowly varying
axial gradient in wall temperature downstream from the
heated length suggests the absence of eddy diffusion
and hence that the flow is laminar. The heat transfer
coefficient for this case is well below that predicted by
the Hausen correlation for the transition region. We
attribute the delayed transition to turbulent flow to the
influence of heating a fluid having a negative tempera-
ture coefficient of viscosity.
Wetting studies. The bubble-pressure technique has
been employed to determine the contact angle for the
salt mixture (LiF-BeF,-ThF4-UF4; 67.5-20-12-0.5 mole
%) on Hastelloy N as a function of time and tempera-
ture. It was found that low-oxygen-content salt (~85
ppm) does not wet a Hastelloy N surface upon initial
exposure at 1000°F, but gradually becomes strongly
wetting. At 1230°F the salt returns to the partial
wetting condition. The increase in wetting with time is
believed to be due to moisture in the helium purge gas
which results in formation of an oxide film on the
surface.
Mass transfer to circulating bubbles. The dependence
of the mass transfer Sherwood modulus for bubbles and
liquids in cocurrent turbulent flow on Reynolds modu-
lus has been examined theoretically for the flow
regimes distinguished by a bubble Reynolds modulus
less than or greater than 2. For the former case the
exponent of the Reynolds modulus is found to be 0.92,
whereas for the latter case the exponent is 0.66. The
experimentally observed value is 0.94. The experi-
mentally determined dependence of the Sherwood
modulus on bubble size is significantly greater than that
predicted for either regime.
PART 2. CHEMISTRY
6. Fission Product Behavior
The analysis of data from an experimental array
exposed in the MSRE core shows that on paired metal
and graphite surfaces, tellurium and molybdenum were
deposited considerably more strongly on metal than on
graphite. Niobium and ruthenium were less intensely
deposited, favoring metal surfaces only moderately
more than graphite. Such individual characteristics will
have to be considered in anticipating their behavior
under MSBR system conditions.
The presence of superficial grain boundary cracks on
Hastelloy N samples from the MSRE fuel system has
prompted a joint investigation by the Reactor Chemis-
try Division and the Metals and Ceramics Division of
those fission products which might have contributed to
the process. Preliminary tests have exposed specimens
of Hastelloy N, nickel, and type 304L stainless steel to
elemental vapors of tellurium, selenium, sulfur, iodine,
cadmium, arsenic, and antimony at 650°C for 1000- to
2000-hr periods. Only those specimens of Hastelloy N
that were exposed to tellurium alone showed effects
similar to that found in the MSRE fuel system. Nickel
specimens exposed to tellurium showed much less
severe attack, and all others showed little if any
deleterious effect.
Tellurium forms the volatile fluorides TeF, and
TeF4, both of which are readily reduced to elemental
Xiii
tellurium by UF; (or by stronger reducing agents).
Accordingly, a simple method for controlled generation
of these materials would permit corrosion testing of
metal specimens in molten-salt mixtures to which
known and controlled additions of tellurium are being
made. Experimentation has shown that the reaction of
tellurium with anhydrous CoF; affords such a conve-
nient generation method. Reaction of tellurium with
CoF; when mixed as powders is rapid, but cannot be
well controlled. When these reagents are placed in a
compartmented cell, such that they can mix only upon
vaporization of tellurium, TeF¢ appears as the cell is
heated to 225°C; increasing pressures of TeF are ob-
served as the temperature is raised to 275°C. A small
quantity of TeF, appears in the TeF¢ at 300°C, and no
other vapor species appear below 400°C. Apparatus for
using this method to introduce tellurium into molten-
salt systems is under construction.
7. Behavior of Hydrogen and Its Isotopes
Hydrogen and helium solubilities in Li, BeF, have
been determined at 600°C over the pressure range 1 to
2 atm. As expected, in both cases the solubility varied
linearly with saturation pressure over the range of
pressure investigated.
Chemical events associated with birth of tritons
through neutron capture in lithium and with thermali-
zation of the energetic triton have been considered.
Path length for thermalization is such that in an MSBR
core only a few percent of the tritons should reach
graphite before thermalization. Production of atomic
fluorine during thermalization of the triton adds an
insignificant fraction to the (low) concentration of F°
due to thermalization of the fission fragments. Overall,
it appears that previously assumed equilibria involving
U3, U™, TF, and tritium should be applicable.
Deuterium permeation through Hastelloy N has been
measured over the pressure range from 30 torr down to
less than 1072 torr. Although earlier workers have
uniformly reported variance from a dependence as the
14, power of pressure at pressures below 10 to 100 torr,
no such departure was found in the present work.
Accordingly, the extrapolations which have been used
within MSRP for calculation of tritium permeation
appear correct.
Initial experiments on type 304L stainless steel
showed that an oxide film decreased the permeation
flow and that such a film resulted in deviation from
dependence on 4 power of the pressure. Presently
available (and very simple) models of film behavior are
inadequate to explain the observations.
Measurements of hydrogen diffusing out of closed
capsules containing fluoroborate from a convection
loop confirmed analyses of substantial (26 to 40 ppm)
concentrations of hydrogen within the salt. The mea-
surements also provided an estimate of the equilibrium
quotient of the reaction
OH™ + 3 Cr(s) = 5Cr® + 0% + Y, H, (g)
in molten fluoroborate at 535°C.
Infrared absorption techniques are valuable in quanti-
tative study of BF;O0H™ and BF3;0D™ ions in molten
fluoroborates. These studies have now been greatly
facilitated by design and fabrication of an LaFj;-
windowed cell and furnace assembly. Operation of the
system with NaF-NaBF, mixtures to 450°C appears
quite satisfactory.
Continued study, by infrared absorption, of NaF-
NaBF, mixtures indicates, as before, the band at 3645
cm ™ due to the stretching mode of OH in BF;OH™
and that at 2690 cm™' due to the OD stretch in
BF3;0D". These studies have, in addition, revealed that
simple bubbling with D, of a clean NaF-NaBF, melt
containing BF;OH™ does not lead to appreciable
exchange of deuterium for hydrogen in the BF;OH™.
Reaction of the admitted D, to produce D™ prior to
the exchange appears necessary, though the species
responsible for the oxidation of D, to D™ has not been
certainly identified.
Hitec, a commercial heat transfer medium consisting
of NaNOQO;, NaNQ,, and KNOj;, has been considered as
a coolant for an MSBR since it would certainly oxidize
and retain tritium. Thermal stability of this mixture to
600°C and of the individual components to 400°C has
been examined, using a mass spectrometer to identify
the gaseous species evolved.
8. Fluoroborate Chemistry
Measurements of the solubility of BF; in LiF-BeF,
were continued, using an improved more-rapid tech-
nigue. The measurements continue to show that over
the composition range 50 to 85 mole % LiF, BF,
solubility varies almost linearly with the thermo-
dynamic activity of LiF. A study of the solubility of
BF; in MSBR fuel solvent indicated that BF3 gas is
potentially useful in a reactivity control system.
The formation free energies of NaNiF; and NaFeF;
have been estimated from heterogeneous-equilibria mea-
surements. This permitted estimation of the most
oxidizing conditions which can exist in the presence of
Xiv
molten mixtures of NaF and NaBF4without significant
oxidation of nickel.
Approximately 1550 1b of NaBF,-NaF (92-8 mole %)
was prepared for the Coolant Salt Technology Facility.
Analyses of representative salt samples show that
sodium fluoroborate mixtures of sufficient purity can
be prepared from commercially available materials by
evaporation of water vapor during the melting process.
9. Protactinium Chemistry
Studies of various equilibria involving protactin-
ium(IV) and (V), MSBR fuel solutions, and solid oxides
have been completed. Protactinium(IV) and (V) con-
centrations have been measured as a function of
temperature at oxide saturation, and the potential of
the Pa*/Pa’* couple has been estimated. From these
results it is evident that protactintum could be precipi-
tated in a process side stream as Pa®* oxide while
avoiding oxide precipitation in the main fuel circuit,
where the Pa** state is maintained by control of the
redox potential (U*/U%).
Equilibrium measurements have been performed for
the distribution of Pa** between molten MSBR solvent
salt and solid solutions of Pa0,-ThO,. The equilibrium
quotients obtained, together with similar data for other
tetravalent actinide ions, yield a correlation with the
lattice parameters of the pure actinide dioxides.
The activity coefficients of ThO,, Pa0,, and UQ, in
the ternary oxide solid solutions which can be precipi-
tated by oxide from MSBR fuels have been estimated
from a previous correlation for binary oxide solid
solutions. At 600°C, 7p,q, is predicted to vary in the
range 1.0 to 1.4, and the Pa/U ratio in the oxide solid
solution precipitated from an MSBR fuel will be ~Y% of
the corresponding ratio in the fuel.
10. Development and Evaluation
of Analytical Methods
for Molten-Salt Reactors
Automated in-line measurement of U(III) concentra-
tions in a thermal convection loop continued to show
excellent reproducibility and to demonstrate the ad-
ventitious oxidation of the U(IlI) by contaminants
introduced during the insertion of corrosion specimens
and the recovery of the U(III) by oxidation of the
chromium from the structural metal and specimens. A
comparison of the experimental voltammetric waves for
U(II) with those computed from theoretical considera-
tions has shown that, while the reduction of chromium
makes a significant contribution to the early part of the
wave, an excellent match between theoretical and
experimental data is obtained in the region of the peak
reduction current. Moreover, the presence of chromium
does not introduce any significant error in the potential
at which the maximum in the derivative wave occurs.
The location of the derivative maximum is used to
compute the U(III) concentration. Under the reducing
conditions at which the loop has been operated,
chromium is the only corrosion product that is present
in significant concentrations. Although the voltam-
metric determination of chromium is complicated by its
proximity to the larger uranium wave, we have found it
possible to extract a measurable chromium wave from
the reoxidation curves by a stripping technique. Calibra-