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ANL-7092
Reactof-Technology
(TID-4500, 46th Ed.)
AEC Research and
Development Report

 

ARGONNE NATIONAL LABORATORY
9700 South Cass Avenue
Argonne, Illinois 60440

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CATALOG OF NUCLEAR REACTOR CONCEPTS

Part 1. Homogeneous and Quasi-homogeneous Reactors
Section III. Reactors Fueled with Molten-
salt Solutions

by

Charles E. Teeter, James A. Lecky,
and John H. Martens

*,

Technical Publications Department

September 1965

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SECTION III. REACTORS FUELED WITH MOLTEN-SALT

SOLUTIONS. . .. .. e e e e e e e e e e e e e
Chapter 1. Introduétion ...........................
Chapter 2. One-region Reactors .. ... ... ... e
Chapter 3. Two-region Reactors . e

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PREFACE

This report is an additional section in the Catalog of Nuclear Reactor
Concepts that was begun with ANL-6892 and continued in ANL-6909. As in
the previous reports, the material is divided into chapters, each with text
and references, plus data sheets that cover the individual concepts. The
plan of the catalog, with the report numbers for the sections already issued,
is given on the following page.

Dr. Charles E. Teeter, formerly employed by the Chicago Oper-
ations Office at Argonne, Illinois, is now affiliated with the Southeastern
Massachusetts Technological Institute, New Bedford, Mass. Through a
consultantship arrangement with Argonne National Laboratory, he is con-
tinuing to help guide the organization and compilation of this catalog.

J.H.M. ,
- September, 1965

 
 

 

PLAN OF CATALOG OF REACTOR CONCEPTS

General Introduction ANL.-6892 S
Part I. Homogeneous and Quasi-homogene.ous Reactors ' .
Section I. Particulate-fueled Reactors ~ ANL-6892
Section II. Reactors Fueled with Homogeneous :
' Aqueous Solutions and Slurries ANL-6909
Section III. Reactors Fueled with Molten-salt
Solutions This report
Section IV, Reactors Fueled with Liquid Metals
Section V. Reactors Fueled with Uranium_Hexé.—
fluoride, Gases, or Plasmas
Section VI. Solid Homogeneous Reactors s

Part II. Heterogeneous Reactors

Section I,
Section II,
Section III.
Section IV.
Section V.
Section VI,
Section VII,
Section VIII.

Section IX,

Part III. Miscellaneous Reactor Concepts

Reactors Cooled by Liquid Metals
Gas-cooled Reactors

Organic—cooled Reactors

Boiling Reactors

Reactors Cooled by Supercritical Fluids
Water-cooled Reactors

Reactors Cooled by Other Fluids
Boiling-water Reactofs

Pressurized-water Reactors

 
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PART I. HOMOGENEOUS, AND QUASI-HOMOGENEOUS REACTORS

SECTION III. REACTORS FUELED WITH MOLTEN-SALT SOLUTIONS

 

Chapter 1. Introduction

The reactor concepts described in this section utilize a fluid fuel
consisting of a fissionable material dissolved in a carrier of molten salt.
Some such concepts also call for a molten salt as a primary coolant.

H. G. MacPherson has reviewed the technology of molten-salt
reactors.ls?

The fissionable compound usually chosen is uranium tetrafluoride.?
Uranium fluorides other than the tetrafluoride have the disadvantages of
higher volatility, instability, or corrosivity. Chlorides or fluorides as
solvents have been given the most consideration because of the need for
radiation stability and high solubility. The chlorides are used for fast
reactors, and the fluorides, because of their low cross sections for ther-
mal neutrons, are best for thermal and epithermal reactors.

Compounds other than fluorides and chlorides have been suggested,
including phosphates,® sulfates,? sulfides,* hydrosulfides,* and hydroxides.5’6
Molten fluoride mixtures, e.g. LiF-NaF, have the most desirable proper-
ties: they dissolve adequate amounts of fuel; they have satisfactory heat
transfer properties; they resist radiation; they can tolerate an accumu-
lation of fission products; the melting points are low enough that corrosion
problems of excessively high temperatures are avoided; and the vapor
pressures are low enough to permit low~-pressure operation.4’7’8 Although
fluorine itself has some moderating properties, a better moderator must
be present, either in the molten-salt mixture or as a separate structure,
to obtain a thermal reactor of reasonable size. The alkali metal fluo-
rides have been especially considered as solvents because they have low
melting points. Beryllium fluoride may be added to the mixture as a
moderator, and thorium tetrafluoride can be added for conversion.

Graphite and beryllium are commonly used as structured moder-
ators. Discussion of structured-moderated reactors in this section may
appear anomalous, in that most reactors in this first part of the catalog
have completely homogeneous cores. The fuel itself, however, is a horno-
geneous solution, and such structure-moderated reactors are otherwise
closely related to the more homogeneous ones.

Molten-salt reactors can be classified in several ways--by the
purpose of the reactor, by the use or nonuse of a separate moderator,
by the cooling method (internal or external), or by the core arrangement
 

(one- or two-region). For this catalog, the classification will be by one-
and two-region reactors. Chapter 2 will cover the first and Chapter 3

the second. According to MacPherson,7 the most attractive types are the
one-region, graphite-moderated reactor, and the two-region reactor. The
one-region reactor is simpler, and it is cheaper to construct and operate -
for small power stations. Most of the one-region reactors discussed in-

Chapter 2 are burners. The two-region reactor has better neutron econ--

omy, is better for breeding, and, in larger installations, gives higher con-

version ratio and lower fuel-cycle costs. |

o

The origin of the molten-salt concept is not clear. Early in the
1950's, however, molten salts were considered as reactor fuels to satisfy
the need for high temperature and extremely high power densities needed
for reactors intended for nuclear aircraft propulsion. Development work,
- particularly at Oak Ridge National Laboratory, resulted in several con-
cepts, and in 1954 the Aircraft Reactor Experiment (ARE), part of the
Aircraft Nuclear Propulsion Project {ANP), was operated. Since the
-cancellation, .as of October 1957, of work on circulating-fuel reactors
for aircraft,? work has continued aimed at developing power and breeder
reactors that utilize molten salts. | .

Molten-salt reactors are attractive concepts for several rea-
sons.1f2’7 They provide high temperatures in a low-pressure system
to produce steam at temperatures high enough to give high thermal-
cycle efficiencies. They are versatile because of the range of solubili-
ties of different compounds of fissionable elements in salts. The simple
ionic salts are stable under irradiation. Such reactors also have the
advantages of other fluid-fueled reactors; for example, they have high nega-
tive temperature coefficients of reactivity, fission products can be re-
moved continuously, fuel elements need not be fabricated, and make-up
fuel may be added as needed. According to Weinberg,!® a great advantage
is that the fissionable materials can be consumed at very high thexrmal
efficiencies and with extremely high burnup. The major problem with a
salt-fueled reactor is that all the salt in the system must be kept molten
at all times.

In 1959, the Fluid Fuel Reactor Task Force compared the aqueous
homogeneous, molten-salt, and liquid-metal fueled reactors.!! The prin-
cipal conclusions were: the molten-salt reactor had the best chance of
achieving technical feasibility; only with the homogeneous aqueous reactor
was there a possibility of achieving a reasonably short doubling time;
and the total power costs for the MSR were between those for the aqueous
reactor and the liquid-metal-fueled reactor.

x

Salt-fueled reactors are advanced concepts, and there is as yet no
adequate experience for building large-scale power plants, although many ?
concepts have been developed for such plants.” Current development is
represented by the Molten Salt Reactor Experiment, which achieved cri-
ticality at ORNL in mid 1965.

 
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References

H. G. MacPherson, Molten Salt Reactors, Chapter 21 in Reactor
Handbook, Vol. IV, Engineering, Stuart McLain and J. H. Martens,
eds., Interscience Division, John Wiley and Sons, New York, 1964.

James A. Lane, H. G. MacPherson, and Frank Maslan, Fluid Fuel
Reactors, Addison Wesley Publishing Co., Reading Mass., 1958.

W. S. Ginell, Molten Phosphate Reactor Fuel. Part I, NAA-SR-
5925, Atomics International, Sept. 30, 1961.

R. C. Crooks, M. J. Snyder, and J. W. Clegg, Fused Salt Mixtures
as Potential Liquid Fuels for Nuclear Power Reactors, BMI-864,
Del., Battelle Memorial Institute, Sept. 8, 1953. Decl. with del.,
Feb. 20, 1957, :

C. S. Dauwalter and J. Y. Estabrook, Circulating Potassium Hy-
droxide Reactors, Y-F¥8-9, Del., ORNL, Dec. 19, 1950.

E. M. Simons and J. H. Stang, Engineering Problems Pertinent to
the Use of Sodium Hydroxide in Reactors, Chemical Engineering -
Progress Symposium Series. Nuclear Engineering, Part I, No. 11,

 

Vol. 50, AIChE, New York, 1954, pp. 139-144.

H. G. MacPherson to R. W. Ritzmann, Molten-Salt Reactors: Report
for 1960 Ten-Year-Plan Evaluation, Unpublished report, ORNL,
July 25, 1960.

W. R. Grimes, F. F. Blankenship, G. W. Keilholtz, H. F. Poppendiek,
and M. T. Robinson, Chemical Aspects of Molten Fluoride Reactors,
Proc. 2nd U. N. Int. Conf. on Peaceful Uses of Atomic Energy,

 

Geneva, 1958, 28, pp. 99-111, United Nations, New York, 1958,

Aircraft Nuclear Propulsion Project. Quarterly Progress Report
for Period Ending December 31, 1957, A. J. Miller, Project Co-
ordinator, ORNL-2440, Del., ORNL, April 24, 1958. Decl. with
del., Nov 4, 1959 o o -

A, M, Wemberg, Adva.nc:ed Systems--—A Personal Appraisal, Nuclea:__
Eng. 5, No. 53, pp. 463-465, October 1960.

 

Report of the Fluid Fuel Reactors Task Force, TID- 850'7 USAEC,
Feb. 1959..

W. R. Grimes, Molten Salts as Reactor Materials, Nuclear News.

7, No. 5, pp. 3-8, May 1964.
 

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11

Chapter 2. One-region Reactors

Most of the one-region reactors discussed in this chapter are
burners, with highly enriched uranium in the molten fuel and no external
blanket. The converters have either partially enriched uranium or tho-
rium salts as fertile material in the fuel-salt mixture. One-region re-
actors have been considered both for aircraft propulsion and for the
generation of heat and electrical power. The ANP (Aircraft Nuclear
Propulsion) Project led to many concepts, and one reactor experiment,
ARE (Aircraft Reactor Experiment), was operated for a short time. As
part of the Aircraft Nuclear Propulsion Project, several designs for
molten-salt-fueled aircraft reactors were proposed both before and
after the operation of the Aircraft Reactor Experiment. In addition to
the studies at ORNL, work was carried out by such contractors as the
H. K. Ferguson Company and the Pratt & Whitney Aircraft Company.

In 1957, however, work on circulating-fuel reactors for aircraft was
discontinued. The MSRE (Molten Salt Reactor Experiment) is the cur-
rent major effort on this type of reactor, with criticality achieved in
mid 1965.

Reactors for Propulsion

Early Concepts

 

Concepts reported by H. K. Ferguson in 1950 and 1951 included
both reactors fueled with molten fluorides and a few fueled with suspen-
sions of uranium oxide in molten sodium hydroxide. The suspensions
were used because of the difficulty of dissolving uranium compounds in
NaOH. These suspensions are included here, although they canrnot be
defined as molten-salt: solutlons

In the Homogeneous Circulating Suspension Reactor, the fuel is
a suspension of 2.2 wt.% uranium oxide in sodium hydroxide, which acts
as moderator.! The fuel enters the top of the reactor, where a whirl-
ing motion is 1mparted to 1t by vanes, and leaves through the bottom ta
a heat exchanger.

In a similar concept,! the fuel is a coarse suspension of the oxide
in sodium hydroxide, which circulates through a spherical reactor. The
oxide is removed in a cyclone separator; the liquid passes through a
heat exchanger, and then picks up the oxide before returning to the core.

The Circulating-fluoride. Reactor concept includes beryllium
rods as moderator and a molten fluoride fuel, which circulates to a
wraparound heat exchanger.” This concept was studied as a variation
of a reactor fueled with molten metal.

 

 
 

 

 

 

 

12

circulating fuel, and to study the kinetic behavior of the reactor.

In another concept,’ the beryllium moderator is in different forms.
In one, it is distributed throughout the core; in the other, it is a reflec-
tor layer. |

™

The Circulating Moderator- coolant Reactor utilized uranium
tetrafluoride dissolved in a molten mixture of sodium and beryllium
fluorides.? Instead of a solid moderator, molten sodium hydroxide is
used as moderator, coolant, and reflector. The sodium hydroxide flows

‘downward through tubes in the core to cool the quiescent fuel salt. Sec-

ondary cooling is by exchange with sodium in a heat exchanger. The re-
flector is a jacket of the hydroxide around the core. The design power
for this reactor is 140 MW(t). A very similar reactor is the Circulating-
moderator ARE,?

-~ In 1952, Dayton and Chastain made calculations for the desigh of
one- and two-region circulating reactors using hydroxides as moderator-

‘coolants.* The compounds studied were sodium hydroxide, lithium-7

hydroxide, and lithium-7 deuteroxide. Uranium-233 oxide, suspended in
the hydroxide within the spherical reactor, was the fuel specified. For
the one-region reactors, thorium oxide was added to the fuel for breed-
ing. The conversion ratios were so low that the authors concluded that
internal breeding would not be feasible if a small critical mass were
required. If, however, cost and size are not considered, Li’OD appears
to be the most attractive of the compounds.

Two ORNL designs were for a 200-MW(t) Aircraft Reactor (1951)3
and a C;irculating-fuel Reactor for Direct Heat Transfer to Engines
(1953).

In the first reactor, the fuel--a molten mixture of beryllium,
sodium, and uranium fluorides--does not circulate. It is contained in
U-tubes, which are within coolant tubes through which sodium circulates.
Beryllium oxide is the moderator and reflector in the cylindrical reactor.
This was intended to be a full-scale reactor, and the ARE was to dupli-
cate it as far as possible in materials, temperature pattern, and kinetics,
but not in fuel circulation or power. In the second reactor, hot fuel cir-
culates from the core directly to the aircraft engine; otherwise this re-
actor is similar in many respects to the first.

The Aircraft Reactor Experiment

 

The ARE was built at Oak Ridge National Laborétory as a
circulating-fuel reactor of low power and high temperature, but materi-
als suitable for use in a reactor of high power were employed.”"? Ac-

-cording to Weinberg, "The purpose of this reactor experiment was simply

to gain experience in handling salts in a reactor at very high tempera-

tures, to see whether one could in fact contain the intensely radioactive
n10

 
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13

Before the circulating fuel was decided upon, solid fuel pins and
stagnant fluoride fuels were considered.!!

Stagnant fluoride fuel was used in the first design.!! The core
would have a cylindrical moderator matrix, containing vertical holes
into which small fuel tubes would be placed. This assembly would be

‘within a cylindrical pressure shell, through which sodium would be pas-

sed. A slab of boron carbide would be placed at the top of the lattice,
and the fuel tubes would extend through the slab, which would act as a
"neutron curtain” for control. Severe problems, however, made this
concept less attractive than a circulating-fuel reactor. The molten salts
have poor thermal conductivity. The extremely large thermal gradiant
at reasonably high power levels would make the temperature of the fuel
near the center of the tube prohibitively high. There would be difficul-
ties in loading the reactor. Also, during loading, large control rods
would be needed, and the heat cycling and draining of coolant between
fuel additions would pose many problems. Expansion during melting
might be a problem relative to possible deformation (expansion) or
rupture.

The final fuel was uranium tetrafluoride dissolved in a mixture
of sodium and zirconium fluorides. Beryllium oxide was the moderator
and reflector; blocks of it were stacked around the fuel tubes, tubes for
reflector cooling, and control assemblies that passed vertically through
the core. The fuel took a serpentine path through parallel circuits to
the outside of the core, finally leaving at the bottom. It circulated to an
external heat exchanger then back to the core. Sodium passing through
tubes in the moderator cooled it. Barren fuel salt also had been sug-
gested as coolant. The maximum design power was 2.5 MW{(t).'*’1* The
reactor became critical on November 11, 1954, and it was shut down
the following evening. It had demonstrated the feasibility of using a
high-temperature fluoride fuel in a circulating—fuel reactor.'®

A reactor with a tandem heat exchanger was suggested as a
modification of the ARE.!® The moderator - reflector is water or sodium
hydroxide, with the core and the tandem heat exchanger being surrounded
by a layer of water at 300-350 psi. The fuel enters the reactor at the
top, makes a loop through fuel tubes in the reactor, and exits to heat ex-
changers. The moderator enters the lattice around the periphery at the
rear of the reactor and flows to an outlet at the forward end.

In another modification, the fuel-coolant salt circulated through
the coolant tubes in the BeO reflector of the ARE.!® In this modification,
there would be a larger 1eakage of high-energy neutrons and of gamma
radiation than with barren fuel salt as moderator coolant.
 

 

14

'The Fireball and Related Concepts

 

A éoncept that originated before the operation of the ARE and was
scheduled for use in later developments was a circulating-fuel, reflector-
moderated reactor, known originally as the Fireball.!"!® In this reactor :

‘there are no fuel tubes. The fuel circulates in an annulus betwee'n a cen-

tral island of berylliurm moderator and an outer shell of moderator and
reflector. In the earliest design the central island was spherical, but
later development resulted in a vase-shaped island.’® This shape, ac-

~cording to Fraas, reduces the critical mass, improves power distribu-

tion, and hydrodynamically gives the best passage for fuel.’® Cooling is"
by circulating the fluid salt to a circumferential wraparound heat exchanger
between the moderator shell and the pressure shell. There heat transfer
to sodium or sodium-potassium takes place. The fuel passes downward

in the annulus, then outward. The moderator is cooled by sodium flowing
downward between the beryllium and the enclosing shell.

_ Fraas and Savolainen have discussed eight core designs for the
spherical reflector-moderated reactor, with a wraparound heat exchanger,
that are related to the Fireball.!? .

The simplest design is the core in which a thick spherical mod-
erator shell surrounds a spherical chamber for liquid fuel. The shell
has top and bottom ducts for fuel to pass in and out of the core. Because
of absorption of neutrons near the fuel-reflector interface, power density
decreases to a comparatively low value near the center. Also, the flow
pattern is indeterminate. Vanes or screens at the inlet might improve
the flow.

Adding a central island reduces critical mass and gives a more
uniform power distribution. The moderator, however, must be cooled in
this design. Liquid bismuth or lead could be circulated between the fuel
and moderator regions to remove the heat.

Graphite was suggested in two other modifications. A block of
graphite containing parallel passages for fuel flow is placed in the central
zone to give a nearly homogeneous mixture of fuel and graphite in the
core. Concentric shells of graphite could be used as moderator and as
guides for the fuel flow. The authors concluded that these designs were
little better, from the nuclear standpoint, than the simple core with no
moderator structure.

Three designs were for the use of molten sodium hydroxide as a
liquid moderator. In one, the moderator passes through coiled tubes in
the core. The tubes would both serve as moderator in the core and im-
prove distribution of fuel velocity. It would, however, be difficult to , “
avoid local hot spots, and the structural material would capture a high

 

 
 

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15

percentage of neutrons, so that the critical mass would be increased. In
two other modifications, fuel passes through tubes and the hydroxide mod-
erator circulates through spaces between fuel passages. In one design,
the tubes are straight. In the other, they are curved to fit the shell con-
tours, in order to reduce the volume of header regions and to give a more

‘nearly spherical core shape, with lower shield weight.

Some concepts that have employed the Fireball design with little
change are the Aircraft Reactor Test,?® The Circulating Fluoride- fuel
High Flux Reactor,?! and a modified Fireball concept of General Electric.
In the High-flux :Reactor;, which includes a central island of graphite, use
of a layer of bismuth between the central island and the fuel layer was
suggested by the author to resist flow of thermal neutrons from the central
island, where they are created, to the shell, where they are absorbed. It
would also protect the internal island from gamma radiation.

22

Pratt & Whitney concepts were originally variations and develop-
ments of the Fireball design, with changes in the dimensions and other
factors to give reactors of different power 1eve1_s,23’24 In two simplified
versions?® the annulus of the Fireball is replaced by five tubes in the
center of the core. In one, the core is a graphite cylinder and in the
other it is a beryllium sphere. The design was intended to give more
structural stability and to alleviate problems of flow separation. Graph-
ite was used to obtain more favorable critical mass and power distribu-
tion. The beryllium-moderated concept was developed into a more complete
design.?® In another modification, the island is cylindrical, rather than
vase shaped as in the original Fireball.?” Other variations considered
briefly were the use of beryllium, beryllium oxide, or graphite for the
island and for the reflector, and using and not using a reflector. Alterna-
tive fuels considered were lithium or beryllium fluorides as bases for
fuels, and slurries of uranium dioxides in alkali metals.?® Lithium-7 was
considered as a coolant,*® and zirconium hydride as a moderator.

The use of zirconium hydride as a moderator was Jincorporated in
one concept.’® The core consists of fuel tubes and zirconium hydride
rods, with lithium-7 flowing parallel to the tubes. Two arrangements of
fuel tubes were studied. In one, a multitube arrangement; the fuel flows
down through the inner tubes then returns through the outer tubes. In the
other, a thimble-tube design, the fuel flows through an inner tube and re-
turns through the annulus between this tube and an enclosing outer tube.
Lithium-7 could circulate directly to the engine radiator. The hydride
moderator, which would be clad with molybdenum, is claimed to permit

high tempeérature without the néeed for excessive cooling or structural

support.

In 1955, staff members at the Walter Kidde Nuclear Laboratories,
Inc., published designs for reactors for aircraft propulsion.??
 

16

One reactor was similar to an earlier concept of H. K. Ferguson
Co. The fuel is stationary, within tubes. A great many tubes are re-

‘quired because they must be small to avoid excessive internal tempera-

tures. Heat removal would be difficult because of the poor conductivity
of the molten salts.

Three were circulating-molten-salt reactors, two closely re-
sembling earlier concepts. They included a reactor having the Fireball
structure, with a modification in which the core is cylindrical rather
than spherical; a circulating-fluoride reactor with rods of beryllium
moderator in the core; and a circulating~fluoride reactor with molten
sodium hydroxide as moderator and reflector.

An ORSORT concept, the Screwball (1953), differed from the Fire-
ball chiefly in that helical fuel tubes replaced the annular passage be-
tween the central island, which was eliminated, and the moderator shel
Thus it somewhat resembles the Pratt & Whitney concept described in
Ref. 26. The moderator is circulating NaOD. According to the authors,
problems with the Fireball design indicated need for changes. The Screw-
ball design is claimed to eliminate or alleviate all but one, namely,
pressure surges. Use of fuel tubes reduces uncertainty of sustained in-
stabilities in fuel region. "Self-shielding" should be less with the tubes
than with the spherical fuel annulus of the Fireball. Lower power den-
sity (2.5 'kW/cm3) was chosen to aid problems of questionable fuel density.
Removal of the solid island eliminates the need to cool it. The pressure-
surge problem was not alleviated. Surges may be larger with Screwball
because of the tortuous expansion path out of the core, but pressure is
not expected to be great enough to cause difficulty. Substitution of circu-
lating NaOD leads to problems with corrosion; stagnant or low-velocity
layers of NaOD next to hot fuel tubes or containing shell, therefore, must
not occur.

1030

Reactor for Ship Propulsion

ORSORT students designed a molten-salt reactor for ship propul-
sion, which contained many of the features of the aircraft reactor con-
cepts.’! A compact, high-performance reactor was sought in order to
reduce weight. In the cylindrical reactor core, the moderator is an
array of Inconel-clad beryllium oxide rods tipped with poison material
to reduce end leakage and fissioning in the entrance and exit plenums for
the fuel. The fuel flows up through the central core region, then down
through an annular downcomer at the core periphery. There it enters a
wraparound heat exchanger, which is cooled by molten salt. This
125-MW(t) reactor has a nickel reflector, as well as extensive shielding.

An advanced design of this type of reactor was conceived by the
same group to improve reactor performance and reduce weight. To im-
prove moderation in the core, the designers substituted zircohium hydride

 

 
. _

17

for beryllium oxide as moderator and used a beryllium-containing fuel.
Nickel-molybdenum cladding was substituted for Inconel cladding in the
core because thinner cladding could be used to reduce the amount of
poison in the core. In the resulting design, dimensions and power level
were reduced.

Reactors for Electrical Power and Heat
Even before the Aircraft Reactor Experiment was carried out,
some one-region reactors had been proposed for purposes other than
aircraft propulsion. Development of such reactors has continued. Theyvy

include some designs that generally resemble the ARE in core structure.

Farly Concepts

 

 

 

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An early concept, on which little has been published, is for a
thorium converter for electrical power production, which was developed
by Davidson and Robb.** In this converter, the fuel is a solution of
93 percent enriched U?*®F, in molten fluorides containing thorium fluo-
ride for conversion.

Four concepts, by students at the Oak Ridge School of Reactor
Technology, are intended for power and heat, and one is a breeder.

The Fused Salt Reactor for Power and Heat, proposed in 1953,
could be either a burner or converter, with thorium fluoride added to
the molten salt if conversion is desired.?® The core is a graphite
sphere, through which fuel circulates from the bottom through chan-
nels cut into the graphite. The graphite is clad with an Inconel shell
and is contained in an Inconel pressure vessel. The reactor is designed
for installation at remote locations to produce power--about 5 MW(t)-~-
and heat. The heat is produced by water heated by low-pressure waste
gas from the turbine.

In the Fused Salt Breeder Reactor (FSBR),** the fuel is a solu-
tion of uranium-233 fluoride and thorium fluoride in fused lithium-7
fluoride and beryllium fluoride. Stacked graphite blocks form the core,
which has a spherical top, a flat bottom, and cylindrical sides. The fuel
passes through passages in the graphite and through heat exchangers
in the graphite surrounding the core. The fuel passes to intermediate

heat exchangers cooled by sodium. Two sizes were considered, one for
125 MW{e) and one for 250 MW({e). . -

The 600-MW Fused Salt Homogeneous Reactor Power Plant?®
utilizes a fuel salt of fluorides of uranium-235, zirconium, and sodium.
The reactor is stated to be self-moderating because of the fluorine.

 
 

oo e b ke s

18

The reactor vessel is a vertical cylinder, with a dished bottom, of stain-
less steel. The fuel circulates through the multipass vessel upward to

a sodium loop within the vessel, to an annular U-tube heat exchanger,
and to a secondary heat exchanger. The design of this burner could be
altered so as to breed uranium-235.

~ The Fused Salt Reactor for Process Heat (1956)36 has two features
that differ somewhat from previous designs. The aim was to generate as
much heat as possible in the regions of the reactor not bearing fuel and-
to use a ceramic as both moderator and heat-exchange medium for high-
temperature heat exchangers. Magnesium oxide was chosen as the best
available material to meet both requirements, even though it is not a
very good moderator. The reactor consists of a cylindrical matrix of
hexagonal magnesium oxide blocks penetrated by Inconel tubes in a tri-
angular array. Fuel circulates through these tubes. Beryllium oxide

‘reflector surrounds the core, and a steel pressure vessel is the con-

tainer. Of the 400 MW(t) produced by the reactor, 35 Mw is generated

in the MgO moderator and is used to heat steam from heat exchangers.
The moderator is perforated to allow passage of steam. The remaining
heat is removed by the fuel circulating to external heat exchangers. The
design is for four reactors to be used in conjunction with a coal-
hydrogenation plant. The high-temperature steam from the moderator
would be used to gasify coal to produce hydrogen for the hydrogenation.
The remaining steam, at a lower temperature, would be used to drive
turbines. | |

The use of natural convection has been suggested to eliminate the
problem of providing reliable, long-lived pumps for fuel circulation.??:?®
This advantage is at the cost of a greater fuel volume. The fuel circu-
lates by natural convection through the core, vertical convection risers,
and primary heat exchangers. In the design suggested, the core is spher-
ical, with the primary heat exchanger above the core. The heat-exchange
medium may be either molten salt or helium. Either would give a power
of 22 MW(e). This system may be attractive for some applications because
of easy maintenance and good reliability.

Molten lead is suggested as coolant in an ORNL concept of 1958.%?
The only moderator is the fluorine in the fuel salt--NaF-ZrF,~-UF,. The
molten lead circulates the fuel salt by direct mixing with a jet pump.
Heat exchange is rapid and no primary exchanger is needed. The lead
is separated from the salt downstream by a pipeline separator; the fuel
goes to the core, the lead to a heat exchanger. The design power is

194 MW (e). | ‘

The MSRE and Related Concepts

Before the design of the MSRE was decided upon, other concepts
originated from the Molten Salt Reactor Project at Oak Ridge National

 
[RVIPTRRIFIpE ST

Ll s R T e

 

 

19

Laboratory. Two that were developed furthest were the Slightly Enriched,
Fused-Salt-Fueled Reactor and the Experimental Molten-Salt-Fueled
30 MW(t) Power Reactor.

Predeces sdrs to MSRE

 

The Slightly Enriched, Fused-Salt-Fueled Reactor*™* is a conver-
ter fueled with slightly enriched uranium tetrafluoride dissolved in molten
lithium fluoride and beryllium fluoride. It was proposed in 1958, and a
preliminary design was published in 1959. The cylindrical core is of un-
clad graphite, which serves as moderator. The fuel flows upward through
holes in the graphite. Highly enriched uranium is added as makeup fuel.
The reactor produces 315 MW(e).

The Experimental Molten-Salt-Fueled 30-MW(t) Power Reactor*:*?
is a burner fueled with highly enriched uranium tetrafluoride dissolved in
molten lithium fluoride and beryllium fluoride. There is no other mod-
erator. The core is a sphere of INOR-8. The fuel circulates to a heat
exchanger. Barren molten salt--lithium fluoride-beryllium fluoride--is
the secondary coolant.

MSRE

The Molten Salt Reactor Experirnent'?“*"‘r6 is the first of the three
stages in the development of molten-salt reactors discussed by MacPherson
in 1960.%7 After this one-region reactor, a two-region reactor experiment
was planned, and a high-power prototype of a molten-salt reactor would
follow. ‘

The MSRE has the objectives of showing that a circulating molten-
salt-fuel system will operate successfully and demonstrating a reactor
type that can be developed into an advanced converter or thermal breeder.
It also is intended to demonstrate that unclad graphite is a satisfactory
moderator that can be used in contact with molten salts for extended
periods and that on-site hydrofluorination processing ¢can clean up con-
taminated fuel.

The MSRE is a converter, with the possibility of internal breeding.
In structure it resembles the Aircraft Reactor Experiment more closely
than such later aircraft reactor developments as the Fireball. The mod-
erator consists of vertical stringers of graphite, which form a cylindrical
core within a reactor vessel. The fuel passes downward in an annulus
bétween the graphite cylinder and the containing vessel. It then flows
upward in channels formed between the stringers, out the top to a heat
exchanger (in which the intermediate coolant is LiF-BeF,), and back to
the core.
A P

Cogelos Ao k- G bl

 

 

20

Advanced Concept

The use of molten fluorides as fuel in a 10-MW(e) fast reactor
for spacecraft was considered by Allen.*® Calculations were made to
estimate the size and weight of such a reactor and to compare them with
those of a reactor fueled with uranium carbide. Highly enriched (93.5
percent) uranium tetrafluoride was the fuel, with fluorides of sodium,
beryllium, lithium, and zirconium considered as solvents. The fuel
mixture for which calculations were made is 70 percent UFy, 30 percent
NaF. Lithium is the coolant, and the reactor is unmoderated, with an

~inner reflector of zirconium and an outer reflector of beryllium. The

author concluded that uranium fluoride does not have an apparent ad-
vantage over uranium carbide; it does not achieve the full potential of
liquid fuels for the purpose; and other liquid fuels, such as liquid met-
als, should be studied.

Status

The Molten Salt Reactor Experiment apparently is the only cur-
rent active program for developing molten-salt reactors. Construction
is in progress; and criticality was achieved in 1965.%* The one-region
reactor is meant as an intermediate step in developing a large two-
region breeder, but one-region reactors have also been suggested for
uses in their own right, especially in smaller power stations.*’

i~

)

 
21

 

DATA SHEETS

ONE-REGION REACTORS

‘

T T - gy e A T oy e

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TN TORTER U R T T

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R T R T T T T T T T Y
22

 

 

 

 

 

 

 

 

 
 

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23

No. 1. Homogeneous Circulating Suspension Reactor

H. K. Ferguson Co.

Reference: Unpublished report, H. K. Ferguson Co., Dec. 12, 1950.

Originators: Staff of Atomic Energy Division, Karl Cohen, Director.

Status: Design, 1950.

Details: Thermal neutrons, steady state, burner. Fuel-coolant: suspension
of 2.2 wt % U23502_ in NaOH. Moderator: NaOH. Fuel suspension circulates
to external heat exchanger; secondary coolant: Na. Reflector: 6 in. Inconel
in 3 laminated shells around core. Core: cylinder, within spherical pressure
vessel. Fuel suspension enters top of reactor, where whirling motion is im-
parted to it, and leaves through opening in bottom of core, to shell-and-tube
heat exchanger. Shielding: tank of borated H,0O; Pb; plastic. Control: nega-
tive temperature coefficient; shim rod--3 concentric cylinders of boron
steel, H,O filled, located in nickel-plated Zr thimble through vertical axis

of core. Power: 270 MW(t). Problems: high freezing point of NaOH delays
startup and shutdown; NaOH corrosive; stability of slurry under irradiation
questionable. | '

Code: 0313 17 31312 44. 637 711 . 84677 921 101

.. 81111

No. 2 Homogeneous Circulating Fuel-moderator Slurry Reactor
H. K. Ferguson Co.
Reference: Unpublished report, H. K. Ferguson Co., 1950.

Originators: Staff of Atomic Energy Division, Karl Cohen, Director.
Status: Preliminary study 1950; abandoned.

Details: Similar to concept in Data Sheet No. 1. Coarse suspension of UO;

in molten NaOH circulates through spherical reactor. Fuel is removed in
cyclone separator. Liquid goes through heat exchanger, then picks up fuel
and returns to core. Secondary coolant: Na, NaK, or NaOH.
Code: 0313 17 . 31312 44 637 711 84677 921 101

| ' 81111

 

 
 

e i | seumba i

 

 

 

 

24

No. 3 Circulating-fluoride Reactor

 

H. K. Ferguson Co.

Reference: Unpublished report, June 1, 1951.
Originators: Staff of Atomic Energy Division, Karl Cohen, Director.

Status: Design, 1951.

Details: Thermal neutrons, steady state, burner. Fuel-coolant: UK, dis-
solved in molten NaF-BeF, 3.4 wt % U. Moderator: Be rods, 2 in. diameter,
clad with stainless steel. Fuel circulates to wrap around heat exchanger.
Core: cylinder, 42 by 42 in. Reflector: 2 in. of moderator rods. Control:
negative temperature coefficient, vertically moving control rods; each rod
filled with molten Pb~Cd alloy; rods in thimbles. Power (max.): 152 MW(t).
Code: 0313 15 31211 44 627 711 81112 921 104

' — . 84679

- No. 4 Circulating-fuel, Dispersed-moderator Reactor

 

H. K. Ferguson Co.

Reference: Unpublished report, H. K. Ferguson Co., 1950.

Originators: Staff of Atomic Energy Division, Karl Cohen, Director.
Status: Preliminary study, 1950; abandoned.

Details: Thermal neutrons, steady state, burner. Variation of a liquid-
metal-fueled reactor. Fuel-coolant: molten mixture of U, Na, and Be
fluorides. Moderator: Be or BeO distributed throughout core. Fuel cir-
culates through reactor to heat exchanger cooled by Na, K, or Li. Power
(for liquid-metal fuel): 200 MW(t).

Code: 0313 15 31211 4X 627 @ 7XX 84679 9X 104

No. 5 Circulating-fuel, Reflector-moderator Reactor

 

H. K. Ferguson Co.

Reference: Unpublished report, H. K. Ferguson Co., 1950.
Originators: Staff of Atomic Energy Division, Karl Cohen, Director.

Status: Preliminary study, 1950; abandoned.
Details: Same as concept in Data Sheet No. 4 except that Be or BeO mod-

erator is reflector layer in the spherical reactor.
Code: 0313 15 31211 4X 627 TXX 84679 921 104

 
 

S oo i

 

 

 

i, s e ol o B e e | s 5 s ik G G SN i e | e LGRRRETE .. i

 

25

No. 6 Circulating Moderator-coolant Reactor

 

H. K. Ferguson Co.

Reference: HKF-112.
Originators: Staff of Atomic Energy Division, Karl Cohen, Director.
Status: Design, August 1951.
Details: Thermal neutrons, steady state, burner. Fuel: UF, (18 wt % U235)
dissolved in molten salt--40 mol % NaF-60 mol % BeF,. Moderator-coolant:
molten NaOH. Reactor: vertical cylinder, 32 in. diameter, 32 in. high, with
small, closely spaced Inconel tubes in triangular-pitch lattice. Fuel-
expansion chamber connected to top and bottom of core. Reflector: 4-~in.
jacket of NaOH around core; 10 in. NaOH in top and bottom headers. NaOH
coolant flows downward through tubes in core. Fuel is in space between
tubes. NaOH goes to heat exchanger for heat exchange with Na. Control:
shim control--negative temperature coefficient; fine control--single rod,
along axis of core, consisting of vertical annular rod filled withmolten Pb-Cd.
Power: 140 MW(t).
Code: 0313 . 17 31112 44 627 711 84677 921 106

81112

No. 7 Circulating-moderator ARE

 

H. K. Ferguson Co.

Reference: HKF-112.

Originators: Staff of Atomic Energy Division, Karl Cohen, Director.

Status: Design, August 1951. |

Details: Closely resembles "Circulating Moderator-coolant Reactor,"” Data

Sheet No. 6. Thermal neutrons, steady state, burner. Fuel: molten UFy,-
NaF-BeF; containing 3.40% uranium. Moderator~coolant: circulating molten

NaOH. Reflector: NaOH, 8 in. thick top and bottom, 4 in. thick at circum-
ference. Reactor: cylinder, 32 by 32 in. Quiescent fuel is on shell side of
coolant tubes, through which NaOH flows at 890 gpm. Inlet temperature:
1409°F; outlet: 1419°F (average). NaOH goes to heat exchangers. Control:
shim--negative temperature coefficient, varying level of fuel; fine--axial

" control rod. Power: 1 MW(e).

 

Code: 0313 17 31112 44 627 711 84677 921 106
L 8111X
ki

 

26

No. 8 Unreflected Homogeneous Reactor Moderated by
Sodium Hydroxide

 

 

Battelle Memorial Institute

Reference: BMI-T746.

Originators: R. W. Dayton and J. W. Chastain.

Status: Design calculations, 1952.

Details: Steady state, thermal neutrons, converter. Fuel-moderator-coolant:
homogeneous mixture of U***0,, NaOH, and ThO,. Mixture presumably cir-
culates to external heat exchanger. Core container: zirconium sphere. Core
temperature: 1100°F. No reflector. Calculations indicated that internal
breeding not feasible if small critical mass is required.

Code: 0311 17 31312 45 637 756 84677 91 101

No. 9 Unreflected Homogeneous Reactor Moderated by
Lithium-7 Hydroxide

 

 

| Battelle Memorial Institute

Reference: BMI-T746.

Originators: R. W. Dayton and J. W. Chastain.
Status: Design calculations, 1952.

Details: Same as concept in Data Sheet No. 8, except that Li’OH is used in-

stead of NaQOH.
Code: 0312 17 31312 45 637 756 84677 91" 101

No. 10 Unreflected Homogeneous Reactor Moderated by
Lithium-~7 Deuteroxide

 

 

Battelle Memorial Institute

Reference: BMI-746.
Originators: R. W. Dayton and J. W. Chastain.
Status: Design calculations, 1952.

Details: Same as concept in Data Sheet No. 8, except that Li‘OD is used in-

stead of NaOH.
Code: 0312 17 31312 45 637 756 84677 91 101

 
 

 

 

 

27

No. 11 200-MW(t) Aircraft Reactor
ORNL

Reference: Unpublished report, ORNL, 1951.

Originators: Staff of ORNL ANP Project.
Status: Preliminary design, 1951. Project cancelled, 1957,
Details: Intermediate neutrons, steady state, burner. Fuel:: molten BekF,-

NaF-UF,. Moderator and refiector: BeO. Coolant: Na, primary and secon-

dary. Core: 3 ft square cylinder with ellipsoidal ends. Inconel pressure
shell. 2268 parallel coolant tubes, spaced by perforated and dimpled disks,
run along long axis of core. In each coolant tube are 3 U-tubes (0.100 in.
diameter) containing fuel. Legs of each U-tube connected to separate inlet
and outlet headers. Fuel does not circulate. Control: negative temperature
coefficient; shim control by varying volume of fuel. Power: 200 MW (t).
This reactor intended to be full-scale reactor for which the ARE was to
duplicate, as far as possible, materials, temperature pattern, and kinetics.
Code: 0213 15 31103 44 627 711 84679 921 106

83189

 

No. 12 Circulating-fuel Reactor for Direct Heat Transfer to Engine

ORNL

Reference: Unpublished report, 1953.

Originators: R. W. Schroeder and B. Lubarsky.

Status: Preliminary design, 1953. Project cancelled, 1957.

Details: Intermediate neutrons, steady state, burner. Fuel-coolant: molten
fluorides containing UF,. Moderator: BeQO. Reflector: BeO, cooled by cir-
culating barren molten fluorides. Core: parallel tubes arranged in concen-
tric circles within a cylinder, 40.4 in. diameter, with conical and truncated
ends. Each core tube is surrounded by hot-~pressed BeO. Around core is
BeO reflector. Core has two manifolds to allow two-pass flow of fuel. In-
conel structural material for all metallic parts in contact with fuel or with
moderator coolant. Fuel flows in through center of inlet, passes through
tubes, and leaves through annuli between fuel inlet and wall. Control: nega-
tive temperature coefficient; power demand; fuel drainage for shutdown.
Shielding: Pb, plastic, H;O. Power: 640 MW(t). Direct flow of hot fuel to
engine eliminates liquid~liquid heat exchanger but makes problems in
shielding.

Code: 0213 15 31211 44 627 711 84679 921 104
 

28

No. 13 Aircraft Reactor Experiment with Stagnant Fuel
ORNL

 

Reference: Nuclear Sci. and Eng. 2, No. 6, pp. 804-825, Nov. 1957.
Originators: E. S. Bettis, R. W. Schroeder, G. A. Christy, H. W. Savage,

 

R. G. Affel, and L. F. Hemphill.

Status: Design; abandoned for circulating-fuel reactor.

- Details: Thermal and intermediate neutrons, steady state, burner. Fuel:

molten-salt mixture containing UF,;. Moderator: BeO. Coolant: Na. Core:
cylindrical matrix of BeO with small vertical holes for fuel tubes. Core
within pressure shell. Fuel remains within tubes. Na pumped through pres-
sure shell. Control: poison rods within matrix; slab of B4C at top of lattice.
Fuel tubes are not completely filled at design pressure and zero power, but
fuel level extends above B4,C. Increase in temperature expands more fuel
above B4C for fast control. Difficulties: at reasonably high power, high . -

. thermal gradient in molten salt causes fuel in center of tubes to reach pro-

hibitively high temperatures; difficulties in loading fuel at room tempera- "
ture--phase changes of fuel during heating might rupture fuel tubes, large
control rods would be needed, and heat cycling and draining of coolant be-
tween fuel additions would pose many problems.
Code: 0413 15 31103 44 627 711 8111X 921 106

- ' - 83189

 
ol o

e Bk R

 

 

. i S s

 

29

No. 14 Aircraft Reactor Experiment (ARE)

 

ORNL

References: ORNL-1845, Del.; Nuclear Sci. and Eng. 2, No. 1, Feb. 1957,
pp. 797-853; CF-53-12-9.
Originators: First suggested by R. C. Briant. Project directed by him
until his death, and subsequently directed by W. H. Jordon and S. J. Cromer.
Status: Critical, October 1954. Final shutdown, November 1954.

 

Details: Thermal and intermediate neutrons, steady state, burner. Fuel-

coolant: circulating mixture of 93.4% enriched U5 a5 UF, in NaF and ZrFy.
Moderator and reflector: BeO blocks stacked around fuel tubes, reflector
cooling tubes, and control assemblies. Innermost section: about 3 ft diam-
eter x 3 ft high cylindrical core. BeO in form of small hexagonally machined
blocks, split axially. Fuel circulated in closed loop through 6 parallel cir-
cuits at inlet fuel header at top of reactor core; each circuit makes 11 series
of passes through core, starting at core axis and progressing in serpentine
fashion to periphery of core, finally leaving at the bottom of the core. Fuel
circulated to external heat exchanger and back to core. Reflector coolant:
Na, passed up through reflector tubes, cooling reflector and Inconel pressure
shell and filling moderator interstices before leaving core. Na also helps
transfer heat readily from moderator to fuel stream. Control: one regula-
ting and 3 vertical shim rods of slugs of hot-pressed B,C clad in stainless
steel; negative temperature coefficient. Maximum power: 2.5 MW(t).
‘Code:' 0413 15 31211 44 627 711 81161 921 104

84679

 

No. 15 Aircraft Reactor with Tandem Heat Exchanger

ORNL

Reference: ORNL-1227.

Originators: ANP staff.
Status: Preliminary design, 1952; discontinued.

Details: Thermal and intermediate neutrons, steady state, burner. Fuel:

molten fluorides. Moderator-reflector: H,O or NaOH. Cooling: circula-
tion of fuel to heat exchanger. Reactor and heat exchanger in tandem.
Reactor horizontal cylinder. Core and heat exchanger surrounded by 1/2-»in,
layer of H,O at 300-350 psi. Fuel enters reactor at top rear, makes a com-
plete loop through fuel tubes in core, and discharges to heat exchanger.
Fuel tubes: stainless steel, l%- in. ID, 0.015 in. wall thickness. Moderator
enters active lattice around periphery at rear of reactor and flows toward
outlet at forward end. If H,0O is moderator, double-wall construction is used.
Reactor vessel surrounded by about 4 ft H,0. Control: 2 curtains of 50 Cd
rods mounted on two endless tracks; curtains move from reflector intoactive
lattice; each rod is cylinder, 12 in. long, 1/2 in. OD, 1/4 in. ID. Power:
400 MW(t).
Code: 0413 13 31211 44 627 711 84677 921 104

17 81212

 
 

 

30

No. 16 Fireball, Early Design

 

ORNL

Reference: Y-F10-104.

Originators: Suggested by R. C. Briant to A. P. Fraas, who worked out
details.
Status: Preliminary design calculations; project cancelled 1957.

Details: Thermal and intermediate neutrons, steady state, burner. Fuel-

coolant: -molten fluorides containing uranium. Moderator-reflector: BeO,
graphite, or circulating moderator of NaOD. Fuel circulates to external
heat exchanger. Structure: 3 concentric spheres of BeO or graphite. BeO:
(90% BeO, 10% Na) central sphere, 14 in. diameter; fuel-coolant shell, 4 in.
thick, 22 in. OD. Reflector: 1 {t thick, 46 in. OD. Graphite: central sphere
18 in. diameter; fuel-coolant shell: 6 in. diameter. Reflector: 12 in. diam-
eter. Graphite reduces moderating power, with more leakage of fast neu-

“trons. Deuterium-bearing compound could be added to graphite to make it

equivalent to BeO, or circulating reflector of NaOD might be used. Island

. decreases critical mass and increases uniformity of fissioning density.
- Control: negative temperature coefficient.

Code: 0413 12 31211 44 627 711 84677 923 104
15
17

 

 
 

e b o e A A i ve B e b e e o

e bRl SRS L. L

e oo g e it

 

 

 

31

No. 17 Reflector-moderated Circulating-fuel Reactor (Fireball)
ORNL

References: Y-F10-104; ORNL-1515.
Originators: Suggested by R. C. Briant to A. P. Fraas, who worked out

details.
Status: Design incorporated in Aircraft Reactor Test; project cancelled
1957.

Details: Thermal and intermediate neutrons, steady state, one-region

burner. Fuel-coolant: molten fluorides containing about 2 mol % uranium.
Moderator: Be shell. Fuel circulates to heat exchanger; secondary ccolant:
NaK; moderator cooled by Na. Reflector: Be moderator shell, 12 in. thick.
Around reflector is 1 in. boron carbide. Reactor vessel: four concentric
shells. Two inner shells surround core region and separate it from vase-
shaped island of Be in center of reactor. Quter Be moderator-reflector shell
surrounded by main pressure shell. Fuel-region diameter: 21 in. for

200 MW (t) reactor. Primary construction material: Inconel. Fuel circulates
downward through annulus between two innermost shells, where fission oc-
curs, then downward and outward to circumferential sphericalheat exchanger
that is between moderator shell and pressure shell. Fuel flows upward in
heat exchanger to top, where it enters the top of the annular passage leading
back to the core. Moderator cooled by Na flowing downward in annulus be-

between Be and enclosing shells and back upward through passages in the

Be to external heat exchangers. Vase-shaped island used because it reduces
critical mass, improves power distribution in fuel region, and hydrodynam-
ically gives best passage for fuel. Shielding: pressure shell, Pb shielding,
thermal insulation, and borated water in self-sealing rubber tank. Control
(suggested): fine--one or two rods in central region or in reflector; coarse
shim--negative temperature coefficient. Power: 200 MW (t).
Code: 0413 15 31211 44 627 . 711 8111X 923 104

‘ _ - 84679

No. 18. Fireball without Central Island
| - ORNL |
Refefence: Unpubliéhed report, ORNL, Dec. 3, 1954.

Originators: A. P. Fraas and A. W. Savolainen.
Status: Part of general design survey; project discontinued, 1957,

Details: Same as Fireball, but without central island.
Code: 0413 15 31211 44 627 711 8111X 921 104

84679
 

32

E No. 19 Fireball Cooled by Lead or Bismuth
ORNL

 

Reference: Unpublished report, ORNL, Dec. 3, 1954.
Originators: A. P. Fraas and A. W. Savolainen.
Status: Part of general design survey; project discontinued, 1957.
Details: Same as Fireball, except that liquid Pb or Bi coolant circulates
between fuel and moderator regions.
Code: 0413 15 31211 44 627 711 8111X 923 104

- ' 31105 84679

- 31106

No. 20 Graphite-moderated Fireball

 

ORNL

Reference: Unpublished report, ORNL, Dec. 3, 1954.

Originators: A. P. Fraas and A. W. Savolainen.

Status: Part of general design survey; project discontinued, 1957.

Details: Same as Fireball, except that graphite is moderator, either as

block in central zone with holes for fuel passage, or as concentric shell.

Code: 0413 12 31211 44 627 711 8111X 923 104
84679

No. 21 Fireball Moderated with Sodium Hydroxide
ORNL

 

Reference: Unpublished report, ORNL, Dec. 3, 1954.
Originators: A. P. Fraas and A. W. Savolainen.
Status: Part of general design survey; project discontinued, 1957.
Details: Same as Fireball, except that molten NaOH is moderator. NaOH
either circulates through coiled tubes in core or through spaces between
fuel passages. Tubes are either straight or curved to fit shell contours.
Code: 0213 17 31211 44 627 711 8111X 923 109

84679

 

 
 

N e e ki

 

. kol ...

33

No. 22 Aircraft Reactor Test

 

ORNL

Reference: ORNL-1835.
Originators: R. C. Briant, A. P. Fraas, et al.
Status: Design, 1953; project cancelled 1957.
Details: Thermal and intermediate neutrons, steady state, burner. Fuel-
coolant: molten NaF-ZrF-UF, (50-46-4 mol %) or NaF-KF-LiF-UF,
(11-42-44-3 mol %); enrichment: 93.5% U**®. Secondary coolant: NaK.
Moderator-reflector: Be, 12 in. thick. Reflector coolant: Na. Structure of
reactor same as for Fireball, with spherical central island of Be and outer
Be reflector, and out of core. Fuel circulates downward between inner Be
island and outer Be reflector, and out of core. Then fuel flows upward
through heat exchanger region, which is around spherical core. Fuel is dis-
charged downward into core. Heat is transferred in heat exchangers to the
secondary coolant. Reflector cooled by Na flowing downward through pas-
sages in Be andupwardthrough annular space between Be and enclosing
shells. Central Be island cooled similarly except that the Na enters through
bottom of island and returns to top of reactor through cooling passages in
main pressure shell. Core diameter: 21 in.; island diameter: 11 in. Fuel-
region thickness: 4.5 in. Reflector thickness: 72 in. Shielding: 7 in. bor-
ated H,O; 31 in. H,O. Control: negative temperature coefficient; shim
control: one rod, 5% Ak/k. Power: 60 MW(t)
Code: 0413 15 31211 44 627 711 84679 923 104

81X1X

No. 23 Circulating Fluoride-fuel High-flux Reactor
ORNL

 

Reference: CF-56-~6-9 Rev. 2.

Originator: W. K. Ergen.

Status: Preliminary design data, 1956.

Details: Thermal and intermediate neutrons, steady state, burner. Fuel:

solution of 0.167 mol % U235F4 in NaZrFs. Moderator-reflector: graphite.

Coolant: presumably, circulation of fuel solution to external heat exchanger.
Co_r_'e_:' spherical shell embedded in infinite moderator. Central island of
graphite. Core radius: 50 cm. Flux at center of sphere: 3 x 107*% neutrons/

| ‘sz Control: negative temperature coefficient; prepoisoning or control
‘rods to control eXCess react1v1ty Layer of Bi between central island and

fuel layer suggested to resist flow of neutrons from central island, where
they are created, to shell, where they are absorbed. It would also serve as
a gamma shield for protecting the internal column. Power: 444 MW(t).
Code: 0413 12 31211 44 627 711 84679 923 104

815XX

81X1X

 
 

 

34

No. 24 Reflector-moderated LF-1 through LF-6

 

General Electric. Company

Reference: APEX-135.

Originators: W. C. Cooley and C. Hussey, ANP Project.

Status: Design study, 1953.

Details: Fireball modification. Thermal and intermediate neutrons, steady
state, burner. Fuel-coolant: mixture of fluorides, either "Fulinak" (Li, Na,
K, and U fl_uorides) or "Fubeli" (Be, Li, and U fluorides) containing 93.4%
enriched U?® a5 UF,. Moderator-reflector: Be. Annular core region be-

- tween 9-in. OD internal Be moderator island and 18-in. ID external Be re-
flector. Reflector: 12-in. thick, with 48 in. diameter spherical contour,

around which intermediate heat exchanger is located. Secondary coolant:
Na or NaK. Structural material: Inconel. Fuel flows downward through
annular core passage and returns upward outside the NaK tubes in the inter-

“mediate heat exchanger to the pump section. Fuel inlet temperature:

1000°F; outlet: 1500°F. Reflector and moderator cooled by Na. Control:
negative temperature coefficient; control rod in the central Be island for
fine control; an additional rod may be installed. Concept is identical for
LF-1, -2, -3; -4, -5, and -6, except for power produced: LF-1, 76 MW;
LF-2, 101 MW; LF-3, 115 MW, LF-4, 135 MW; LF-5, 180 MW; and LF-6,
232 MW.
Code: 0413 15 31211 44 627 711 81111 923 104

- 84679

No. 25 300-MW Circulating-fuel Reactor

 

Pratt & Whitney Aircraft Division, United Aircraft Corp.

References: Unpublished reports, 1953-54.

Originators: Staff members.

Status: Design calculations; work discontinued 1957.
Details: Modification of Fireball. Intermediate neutrons, steady state,

burner. Fuel-coolant: molten NaUFjy - NaZrF; (3.26 mol % UF,). Moder-
ator: Be. Fuel circulates to circumferential heat exchanger. Fireball
structure with internal island of Be and Be moderator-reflector. Core
volume:. 127 liters. Be island: 6.69 in. OD. Inconel-clad pressure shell:

~ 34.3 in. OD. Shield: Inconel, with boron thermal shield. Control: probably
- negative coefficient of reactivity and probably vertically moving control rod
 in moderator, as with Fireball. Power: 300 MW(t).

Code: 0213 . 15 31211 44 627 711 84679 923 104
» o - 81X1X

 

 
gt eisian, o o oo i,

L e BANREL

AT W B e

i - i

 

 

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35

No: 26 Core-moderated Reactor (PWAR-7)

 

Pratt & Whitney Aircraft Division, United Aircraft Corp.

Reference: Unpublished report, 1957.
Originators: Staff members.
Status: Preliminary design, 1956; work discontinued, 1957.

Details: ‘Fireball type. Intermediate neutrons, steady state, burner. Fuel-

coolant: NaF-ZrF-UF, containing 3 mol % UF,. Moderator: Be. Fuel
circulates to circumferential NaK heat exchanger. Be island. Reflector:

7 in. Be. Core diameter: 29 in.; length: 37 in. Pressure shell: overall
height, 5 ft; OD, 62 in. Inconel structure. Max. fuel temperature: 1600°F.
Control: presumably negative temperature coefficient. Power: 190 MW(t).
Design for twin reactors. Suggestions to increase power and decrease
weight. Fuels: LiF- or BeFj-based fluorides, or slurries of uranium oxide
in alkaline-earth metals. Coolant: Li’. Structural materials: Mo; Mo- .
0.45% Ti; Nb-0.65% Zr; FP-16 (23% Ni, 14% Mo, 8% Cr, 1% Al, 2.5% Ti).
Code: 0213 15 31211 44 627 711 84679 923 104

No. 27 Modified Fireball (Early Concept for Core-moderated Reactor)

 

Pratt & Whitney Aircraft Division, United Aircraft Corp.

References: Unpublished reports, 1954.
Originators: Staff members.

Status: Design calculations, 1954; project discontinued, 1957.

Details: Same as Fireball, except that fuel annulus is replaced by five tubes
with end diffusers, in a circular pattern, which pass through the core. In
one, the tubes pass through a graphite cylinder within a Be core. Graphite

is believed to give better critical mass and power distribution. In the other
design, the fuel passages are through the spherical Be core, which is within
an Inconel pressure vessel. This modification of the fuel passage is believed
to reduce flow-separation problems and to give more structural stability.

Code: 0213 15 31211 44 627 711 81X1X 923 104

84679
 

36

No. 28 Circulating Fuel Core-moderated Reactor (CMR)

 

Pratt & Whitney Aircraft Division, United Aircraft Corp. ‘ *

Reference: PWAC-186.

Originators: Staff members.

Status: Design, 1956; project discontinued, 1957.

Details: Intermediate neutrons, steady state, burner. Fuel-coolant:

- enriched UF, in molten NaF-ZrF4.. Moderator: Be. Fuel circulates to

wraparound NaK heat exchanger between reflector and pressure shell.
Reflector: Be. Core: essentially cylinder of Be (29.1 in. diameter, 37 in.
high) through which 30 straight parallel tubes (3 in. ID) pass. Fuel enters
core from plenum chamber above, passes downward through fuel tubes, -
enters plenum at bottom, and flows on shell side of heat exchanger. Fuel
enters core at 1200°F; leaves at 1600°F. Core and reflector cooled by Na
flow. Internal shield: cermet of BiOC and Cu, clad with Inconel, between
reflector shell and reflector support shell. Control: central vertical rod,
in thimble, for reactor shim and control-~rod is clad cermet of BiOC and
Cu cooled with He; negative temperature coefficient. Power: 190 MW(t).
Two reactors to be used in tandem. Use of moderator in core expected to
lower fuel requirements. Using tubes instead of annulus for fuel expected
to eliminate unpredictable and possibly unfavorable flow and heat-transfer
characteristics of the original Fireball core. A cylindrical rather than ‘
spherical core should simplify structural and fabrication problems.
Code: 0213 15 31211 44 627 711 81111 923 104

84679

No. 29 Circulating Fuel, Reflector-moderated,
Epithermal, Aircraft Reactor

 

 

Pratt & Whitney Aircraft Division, United Aircraft Corp.

Reference: PWAC-189.
Originators: Staff members.

Status: Design, 1956; discontinued, 1957.

Details: Fireball structure. Intermediate neutrons, steady state, burner.
Fuel-coolant: NaF-ZrF,-UF, (56.3, 37.2, 6.5%); uranium 93.5% enriched.
Moderator: Be. Core components cooled by Na. Reflector: Be, 48 in. OD.
Fuel annulus: 6 in. thick. Inner island is cylindrical, 8 in. diameter.
Pressure shell: sphere, 70 in. diameter. Fuel enters core at 1200°F, cir-
culates downward then outward and upward and leaves at 1600°F to NaK
wraparound heat exchanger. Internal shielding: shells containing B,
Control: central vertical rod, cooled by He, consisting of rare earth oxides
in Ni matrix, clad with Hastelloy-X; negative temperature coefficient.
Power: 194 MW(t). Two reactors to be used in tandem for aircraft
propulsion. ' -
Code: 0213 15 31211 44 627 711 81114 923 104 |

84679

 
 

 

i, AL Al i el m b e e A, SLAM G i

i,

 

37

No. 30 Direct-circulating, Zirconium Hydride-moderated Reactor

 

Pratt & Whitney Aircraft Division, United Aircraft Corp.

Reference: Unpublished report, 1956.
Originators: Staff members.

Status: Design, 1956; project discontinued, 1957.

Details: Thermal and intermediate neutrons, steady state, burner. Fuel:
uranium salt in molten fluorides, possibly LiF - or BeF-based. Moderator:
ZrHy. Coolant: Li’. Core: fuel tubes and ZrH rods, clad with Mo. Cool-
ant flows parallel to tubes. In one design, the fuel flows down through tubes
in center, goes to a header, and returns through outer tubes. In alternative,
fuel flows down through inner tube and up through annulus formed by enclos-
ing inner tube with an outer one. Li’ coolant could flow directly to an engine
radiator. Use of ZrHX permits high temperature without need for excessive
cooling or structural support.

Code: 0413 17 31106 44 627 711 84679 9XX 106

 

No. 31 Stationary Fluoride Fuel, Sodium Cooled,
Reflector-moderated Reactor

Walter Kidde Nuclear Laboratories, Inc.

Reference: WKNL-42.

Originators: Staff members:

Status: Design for evaluation, 1955. Project discontinued, 1957.

Details: Similar to H. K. Ferguson concept in Data Sheet No. 6. Therrnal
neutrons, steady state, burner. Fuel: either U2 or U?® in molten KF-
NaF-ZrF,-UF, mixture (4 mol % UF,). Stationary fuel is in tubes. Using
U** instead of U%% increases coolant volume and reduces fuel required,
but increases average heat flux excessively. Increasing U?*® concentration
to 27.5 mol % also would increase fuel volume, but because of decreased
thermal conductivity would require more than three times as many fuelele-
ments, 68,400 instead of 20,400. Supporting so many with parallel flow of
coolant would be very difficult. Coolant: Na. Difficulties: difficult heat
removal because of poor thermal conductivity of the molten salts; because
tubes must be small to avoid excessive internal temperatures, many tubes
are required. '

Code: 0313 1X 31103 44 627 711 8XXXX 9XX 106
oo Rt g . AL decia

 

38

No. 32 Beryllium Core Moderated, Circulating Fluoride-fuel Reactor

 

Walter Kidde Nuclear Laboratories, Inc.

Reference: WKNI.-42.

Originators: Staff members.

Status: Design for evaluation, 1955. Project discontinued.

Details: Thermal neutrons, steady state, burner. Fuel-coolant: molten
fluorides--50.8 mol % NaF, 46.8 mol % ZrF,, 2.4 mol % UF,. Moderator:
1630 Inconel-clad Be rods, 0.88 in. OD, 42 in. long, in core. Annular heat
exchanger. Reactor container: 64-in. ID multiwall Inconel shell with dished
bottom; cooling channels in wall. Core: circular cylinder, 72 in. by 72 in.
Fuel flows downward parallel to rods, upward through Na-cooled heat ex-
changer, and back to core through pumps. Reflector: 4-in. blanket of 885 Be

- rods cooled by fuel. Fuel inlet temperature: 1500°F, outlet (max), 1150°F.

Control: negative temperature coefficient; 13 borated steel control rods in
Inconel thimbles extending from top of reactor shell into core, replacing like
number of moderator rods. Power: 300 MW(t).
Code: 0313 15 31211 44 627 711 81111 921 104

| 84679

No. 33 Beryllium Reflector Moderated, Circulating Fluoride-fuel Reactor

Walter Kidde Nuclear Laboratories, Inc.

Reference: WKNIL-42.

Originators: Staff members.

Status; Design for evaluation, 1955; project discontinued.

Details: Thermal and intermediate neutrons, steady state, burner. Fuel-
coolant: molten fluorides--50 mol % NaF, 46.2 mol % ZrF, 3.8 mol % UF,,
Moderator~reflector: clad Be. Fireball structure. Island: 12 in. OD;
Reflector: 24.4 in. ID, 48.4 in. OD; Inconel pressure shell: 69 in. OD. Fuel
circulates from core annulus, up through wraparound Na-cooled heat ex-
changer, and back to top of core. Island and reflector cooled by separate Na
system. Control: negative temperature coefficient; vertical rod through
center of island. Power: 300 MW(t). Alternative design: cylindrical re-
flector to avoid difficulties of fabricating spherical reflector. Moderator
shell, island, and fuel annulus as in spherical design.

Code: 0413 15 31211 44 627 711 8111X 923 104

84679

 

 
 

ren e o o e S

ot a1

 

o

39

No. 34 Sodium Hydroxide Core Moderated,
Circulating Fluoride-fuel Reactor

 

 

Walter Kidde Nuclear Laboratories, Inc.

Reference: WKNL-42.

Originators: Staff members.

Status: Design for evaluation, 1955. Project discontinued.

Details: Thermal neutrons, steady state, burner. Fuel-coolant: molten

fluorides-~51.1 mol % NaF, 47.0 mol % Zr¥, 1.0 mol % UF,. Fuel circulates

to Na-cooled heat exchanger. Moderator: molten NaOH in 98 Inconel U-tubes,
2 in. OD, 42 in. long. Moderator circulated to external Na-cooled heat ex-
changer. Reflector: 4 in.thick annulus containing NaOH. Core: circular
cylinder, 42 by 42 in. Pressure shell: 70 in. OD. Fuel inlet temperature:
1150°F; outlet (max) 1765°F. Control: negative temperature coefficient,

4 vertical borated steel control rods. Power: 300 MWI(t).

‘Code: 0313 17 31211 44 627 711 81111 921 104

84677
 

BT R o R i i R WSR2t

 

 

i .

oo BT BERERG .

40

No. 35 Reflector-moderated, Circulating-fuel Aircraft Reactor (Screwball)
| | ORSORT

Reference: CF-53-9-84.

Originators: J. H. MacMillan. et al.. _
Status: Preliminary design study for feasibility; ORSORT term paper, 1953.

Details: Modification of Fireball. Thermal and intermediate neutrons,

steady state, burner. Fuel: molten mixture of 50 mol % NaF, 47 mol %

ZrFy, and 3 mol % enriched UF,. Moderator: circulating molten NaOD.

Coolant: NaK. Reflector: spherical Be shell, 11 in. thick, 2 in. OD.
Structure: spherical, with Be shell, pressure shell, and helical fuel tubes

in annular form. Inconel structural material; clad with nickel where in con-
-~ tact wit}h‘Na.OD, Helical tubes used to reduce uncertainties of unstable flow

in reactors of high power density. Fuel enters core at north pole and flows
downward through six 3.5 in. ID Inconel tubes. Five tubes are wrapped in a
variable-pitch helix to form a spherical annulus of fuel. Sixth passes through
center of sphere, forming smaller-diameter helix. In returning, fuel flows
over primary NaK-cooled heat exchanger. NaK coolant in circumferential
heat exchanger, spherical shell. Fuel and coolant flow countercurrently.
NaOD, which cools reflector, flows downward through spherical cavity in
reflector and surrounds fuel tubes; acts as moderator "island." Returns to
top through holes in reflector, through spherical cavity outside reflector,

through a shell-and-tube heat exchanger and back to core. NaOD cooled in

exchanger by exchange with NaK that is returning to reactor. NaK then goes
to primary heat exchanger to cool fuel. Control system needed to keep temp-
erature of returning NaOD constant to prevent corrosion from over-heating or
freezing from cooling. Heat-exchange system uses only one intermediate
heat-transfer medium and eliminates need for additional radiators to cool
part of NaK, as proposed for Fireball. Control: negative temperature coef-

ficient; shim control--adding enriched fuel; variable by-pass in xenon sepa-

rator would add xenon to provide fine adjustment in reactivity between
additions of enriched fuel. No rods. Basically, power extracted is deter-
mined solely by demand of propulsion system. Power: 200 MW(t).

- Code: 0413 17 31204 44 627 711 84677 923 107

83789
81596

 
b e ot i 3, s

 

 

 

41

No. 36 High-performance Marine Reactor

ORSORT

 

Reference: Unpublished ORSORT term paper, August 1957.
Originators: K. H. Dufrane, T. G. Barnes, C. Eicheldinger, W. D. Lee,
N. P. Otto, C. P. Patterson, T. G. Proctor, R. W. Thorpe, and R. A. Watson.

Status: Design and feasibility study, 1957.
Details: Thermal and intermediate neutrons, steady state, burner. Fuel-

coolant: U?* in molten-salt mixture containing 49 mol % NaF, 45% ZrF,,
6% UF,. Moderator: cylindrical Be rods, clad with Inconel, equally spaced
throughout core region in triangular pitch array and extending length of core.
Rods tipped with poison material (BeO plus B!) to reduce end leakage, as
well as to reduce fissioning in exit and entrance plenums for fuel. Fuel cir-
culates to secondary heat exchanger. Reflector: Ni blanket, 6 in. thick,
around core. It is surrounded by 5%—in. thick region of cylindrical rods,
3/4-in. thick, containing mixture of BeO and B'°. Boron-bearing Inconel
rods in interstices of cylinders. Reactor vessel: cylinder, 80 by 80 in.,
containing expansion tank for fuel and coolant coils in head for removing
internally generated heat by flow of portion of fuel. Shielding: primary--
structural steel plus 5 in. Pb plus 39 in. H,0O; secondary-—4—6«%~ in. Pb;
thin slab of B,C in Cu matrix surrounds this region. Primary construction
material: Inconel. Fuel flows up through central core region (75 cm diam-
eter, 80 cm high) and down through annular downcomer at periphery con-
taining primary (fuel-to-secondary fluid) wraparound heat exchanger cooled
by molten salt (30 mol % NaF, 20% LiF, 50% BeF,). Heat exchanger of once-
through, shell-and~tube, counterflow design; fuel goes to shell side, coolant
to tube side. Fuel returns to core, coolant to steam-generating equipment.
In a modification, an intermediate heat exchange with a tertiary fluid sug-
gested to reduce shielding weight. Control: negative temperature coefficient;
varying coolant flow; single vertical controlrod (Inconel -BeO-Ni-1 vol. %
B'%) in thimble extending length of core at core centerline for reactor shut-
down, change in mean temperature, and fuel burnup; rod thimble about 4 in.
diameter, with gap for cooling by molten salt or metal; provision for emer-
gency fuel dumping. Power: 125 MW(t).
Code: 0413 15 31211 44 627 711 81111 921 104

| ' | | 83789

84679
ik i i i

 

 

42

No. 37 Modified High-performance Marine Reactor
ORSORT

 

Reference: Unpublished ORSORT term paper, August 1957.

Originators: K. H. Dufrane, T. G. Barnes, C. Eicheldinger, W. D. Lee,

N. P. Otto, C. P. Patterson, T. G. Proctor, R. W. Thorpe, and R. A. Watson.
Status: Preliminary study, 1957. |

Details: Modification of design in Data Sheet No. 36 to give advanced design

- that would reduce weight and improve performance. Fuel-coolant: molten

salt, 42 wt % BeF-38% NaF-20% UF, Be salt to increase moderation in core.
Moderator: ZrHY rods, 0.5 in. diameter, instead of BeO to increase modera-
tion. Ni-Mo cladding (0.01 in.) used on rods instead of Inconel. Ni-Mo cor-
rosion resistance permits thinner cladding on moderator and thus less poison
in core. Core: 40 cm diameter by 78 c¢m high. Fuel circulates to Na-cooled
U-tube heat exchanger. Fuel enters core at 1100°F, leaves at 1300°F. Inter-
mediate heat exchanger used to reduce amount of shielding necessary. Shield-
ing: primary--1 in. structural steel just outside insulation of core vessel;
next 15.7 in. H;Oand 6 in. Pb; finally 70 in. H,O in l/Z—in. thick steel vessel.
Power: 100 MW(t). Other details same as original design.
Code: 0413 17 31211 44 627 711 81111 921 104

83789

84679

No. 38 Molten-salt Thorium Converter for Electrical Power Production

 

Knolls Atomic Power Laboratory

Reference: KAPL-M-JKD-10.

Originators: J. K. Davidson, W. L. Robb, L. Bernath, W. H. Horton,
R. P. Schuman, R. H. Simon, and A. D. Tevebaugh.

Status: Informal reactor evaluation, 1952-53; no further work.

Details: Thermal and intermediate neutrons, steady state, converter.

Fuel: solution of 93% U?*®F; or UF, in LiF-BeF,-ThF,. Core: Inconel,

5.77 ft radius. Up to 85% average conversion of Th®* to U?33 is possible;
conversion ratio would be 0.73 to 0.905. Fuel would have to be added
periodically. ‘

Code: 0411 1X 31211 44 627 746 84679 XX 104

 

 
 

e

i ik

 

 

4

43

No. 39 Fused Salt Reactor for Power and Heat
ORSORT

 

Reference: CF-53-10-26
Originators: Theodore Jarvis et al..
Status: Conceptual design, 1953; ORSORT term paper.
Details: Thermal and mixed neutrons, steady state, burner or converter.
Fuel: U?®F, (0.3 mol %); thorium fluoride might be added for conversion,
Critical mass U?*®: 2.7 kg. Moderator: graphite, of high density to avoid
penetration of fuel. Reflector: graphite sphere. Containment: graphite
sphere in Inconel shell, surrounded by Inconel pressure vessel. Fluoride
fuel circulates from bottom through many parallel channels cut directly into
the graphite moderator out the top of the reactor to external heat exchangers,
in which nitrogen is the heat-exchange medium. Channels of different length
so that core approaches shape of sphere. No provisions for cooling channels
in graphite moderator because its temperature will not exceed 2000°F at
maximum power level and with surface cooling. Reactor pressure shell,
heat exchanger, and piping are Inconel. Control: negative temperature coef-
ficient; dumping as additional emergency control; no control rods. Power:
5.3 MW(t). Reactor de signed for installation at remote location to produce
electrical power, as well as power for station heating. Heat provided by
water heated by low-pressure waste gas from the turbine exhaust.
Code: 0411 12 31211 44 627 711 84679 921 104

0413 746 83189

No. 40 Fused Salt Breeder Reactor (FSBR)
ORSORT

 

Reference: CF-53-10-25.

Originators: D. B. Wehmevyer et al.

Status: Design study, 1953; ORSORT term paper.

Details: Thermal and intermediate neutrons, steady state, breeder. Fuel-

coolant-fertile material: solution of U?**F, and ThF, in molten Li'F and

BeF,. Intermediate coolant: Na. Moderator-reflector: stacked graphite
blocks, which contain the salt. Fuel is pumped through graphite passages
and through heat exchangers located in the graphite mass surrounding core.
Critical radius of core: 160 cm. Core has spherical top, flat bottom, and
cylindrical sides. Average operating temperature: 1400°F. Control: nega-
tive temperature coefficient; control of fuel concentration. Breeding ratio:
1.0. Two sizes of reactors: 310 MW(t), 125 MW(e); 616 MW(t), 250 MW(e).
Code: 0412 12 31211 45 627 746 83789 921 104
| 84679
 

44

No. 41 600-MW Fused Salt Homogeneous Reactor Power Plant
ORSORT

Reference: CF-56-8-208, Del.

Originators: R. W. Davies et al.

Status: Design and feasibility study, 1956; ORSORT term paper.

Details: Intermediate neutrons, steady state, burner. Fuel-coolant: solu-
tion of highly enriched U?**¥F, in molten NaF and ZrF. No moderator other
than fluorine in solution. Reflector: core vessel. Core: cylinder, 2 ft diam-
eter by 10 ft high, surrounded by tube bundles of primary heat exchanger.
Reactor vessel: multipass cylindrical container. Fuel flows up through
core and U-tube heat exchangers and through annular downcomer at periphery
of the vessel. Eight pumps circulate the fuel. Inlet temperature: 1050°F;
outlet: 1200°F. Control: negative temperature coefficient. Intermediate
coolant: Na. Ni-Mo alloys for construction materials. Reactor designed to
burn U2¥, but design could be altered to breed U2, Power: 600 MW(t).
Code: 0213 17 31211 44 627 711 84677 921 101

No. 42 Fused Salt Reactor for Process Heat
ORSORT

Reference: CF-56-8-211.

Originators: J. T. Roberts et al.

Status: Conceptual design, 1956; ORSORT term paper.

Details: Thermal and intermediate neutrons, steady state, burner. Fuel-
coolant: 93% enriched UF, in NaF-ZrF,. Moderator: MgQO ceramic.
Reflector: BeO. Fuel circulated at 17.4 £t/sec through 90 2-in. ID Inconel
tubes in triangular array on 12-in. centers. Tubes are in a 10-ft-diameter
by 12-ft~high cylindrical matrix of hexagonally shaped blocks of MgO. Mod-
erator perforated to allow passage of steam through tubes inserted in holes.
35 MW (t) is generated in the MgO and used to heat steam from 2040°F to
3000°F. Annulus between tubes and walls of MgO may be filled with aluminum
silicate "wool" as a thermal barrier. Fuel heated from 1050°F to 1200°F in
a single pass upward through the core by internal heat generation; delivers
the heat to Na in an external heat exchanger. Core reflected on sides with
8-in. BeO and contained in a 12-ft-diameter by 18-ft-long steel pressure
vessel. Control: rods included, but no details given; presumably also neg-
ative temperature coefficient. Design includes 4 such reactors for an
integrated coal-hydrogenation plant. Each produces 400 MWI(t) 123 MW(e).
Code: 0413 17 31211 44 627 711 8 1XXX 921 104

84679

 

 
 

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i o BRSO M B8, i 2N e DR i A B i M .

 

ek Kb e

 

45

No. 43 Molten Salt Natural Convection Reactor

American Standard and ORNL

References: CF-58-2-46; Trans. ANS, 1, No. 1, p. 64.

Originators: F. E. Romie (American Standard) and B. W. Kinyon (ORNL).

Status: Proposal, 1958.

Details: Intermediate neutrons, steady state, presumably burner. Fuel-
moderator-coolant: solution of UF, in LiF-BeF,. Core vessel: sphere,

8 ft diameter. Fuel solution circulates by natural convection through the
core from the bottom, through vertical convection risers, to primary heat
exchangers above the core. Maximum fuel temperature: 1225°F; minimum:
975-1025°F. Primary heat exchanger may be cooled by either molten salt
or helium: Control: presumably negative temperature coefficient. Power:
60 MW(t); 22 MW (e).

Code: 0213 17 31211 44 627 711 84677 9X 101

No. 44 Homogeneous Fused Salt Power Reactor

ORNL

 

Reference: Trans. ANS, 1, No. 1, pp. 63-64.

Originators: W. A. Box et al.

Status: Feasibility study, 1958.

Details: Intermediate neutrons, steady state, burner. Fuel-moderator:
highly enriched UF, dissolved in molten NaF-ZrF, (57 mol %-43 mol %).
Coolant: molten Pb. Core: cylinder, 10 ft diameter, 10 ft high. Pb cir-
culates molten salt by direct mixing by means of jet pump. Heat exchange
is very rapid; no primary heat exchanger needed. Fuel and coolant sepa-
rated downstream from the jet pump by pipeline separator. Fuel goes to
core, Pb to steam generator. Heat extracted from Pb by six heat-transfer
loops that supply turbine generators. Pressure: 140 psi. Control: negative
temperature coefficient. Power: 600 MW(t); 194 MW(e).

Code: 0213 17 31106 44 711 84677 9X 105

 

 
b bl B Barah e

 

46

No. 45 Slightly Enriched, Fused-salt-fueled Reactor
ORNL

 

References: ORNL-2684, pp. 18-22; CF-58-10-60; CF-59-1-26.

Originators: H. G. MacPherson and C. E. Guthrie.
Status: Proposal, 1958; preliminary design, 1959.
Details: Near-thermal neutrons, steady state, converter. Fuel-coolant:
slightly-enriched 1.3-1.8% U235F4 in molten LiF-BelF,;. Moderator:
graphite. Unclad graphite moderator core, 12;}: ft in diameter and height,
is contained in a cylindrical INOR-8 vessel. Fuel flows at 35,470 gpm from
bottom to top of core through 3.6 in. holes on 8-in. centers in the graphite.
Highly enriched uranium added as make-up fuel. Control: controlling tem-
perature and fuel concentration. Initial conversion ratio of U to Pu: about
0.79. Power: about 775 MWI(t); 315 MW(e).
Code: 0311 12 31211 42 627 743 84679 922 104

83789

No. 46 Experimental Molten-salt-fueled 30 MW (t) Power Reactor

 

ORNL

References: ORNL-2796; ORNL-2684, pp. 3-17.
Originators: L. G. Alexander et al.

Status: Preliminary design, 1960.
Details: Intermediate neutrons, steady state, burner. Fuel-moderator-

coolant: solution of 90% enriched U235F4 in molten LiF-BeF,. Core: sphere,
6 ft diameter. Secondary coolant: molten LiF-BeF,. Single structure con-

~tains core, heat exchanger, expansion tank, and fuel pump volute. Fuel cir-

culates at 1480 gpm through bottom of core, flows upward at walls of core to
heat exchanger and back to core, where it flows down. Steam produced at
1000°F and 1450 psi. Exit temperature of fuel: 1235°F. Structural material:
INOR-8. Control: negative temperature coefficient; control of fuel concen-
tration. Power: 30 MW(t); 10 MW(e).
Code: 0213 17 31211 44 627 711 84677 9XX 101

83789

 
 

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47

No. 47 Molten Salt Reactor Experiment (MSRE)
ORNL

 

References: Proc. Symp. Power Reactor Experiments 1, IAEA, Vienna,
1962, pp. 247-92; Nucleonics. 22, No. 1, pp. 67-70, January 1964.
Originators: MSR project, ORNL.
Status: Conceptualdesign, 1960. Criticality achieved in 1965.
Details: Thermal neutrons, steady state, converter (possibly with internal
breeding). Fuel-coolant: solution of 93.5% enriched U?*F, in LiF-BeF,-
ZrF,-ThF,. Intermediate coolant: Li’F-BeF,. Moderator: unclad graphite,
1064 stringers (2 x 2 x 63 in. long) loosely pinned to restraining beams at
core bottom; form cylindrical core (about 5 ft diameter by 7.5 ft high) con-
tained in a reactor vessel. Annulus between this inner cylinder and outer
shell provides cooling for the shell. Flow is laminar at 1200 gpm; fuel,
blanketed by helium gas, enters top of core at 635°C, flows in a spiral path
downward in annulus along the wall to the bottom where a dished head re-
verses the flow. It then flows up through channels in the graphite core
matrix formed by machining the faces of the stringers. All components in
contact with the fuel are of INOR-8. Exiting at 663°C the fuel enters the
sump-type pump from which it is discharged through the shell side of the
heat exchanger back to the core inlet. Control: negative temperature coef-
ficient; 3 shim control rods--thin-walled cylinders of B4,C made of stacked
short sections clad in INOR-8. Power: 10 MW(t).
Code: 0311 12 31211 44 627 746 84679 923 104

0312 81111

No. 48 Uranium Fluoride-fueled Fast Reactor for Spacecraft

Jet Propulsion Laboratory, Calif. Inst. Tech. .

Reference: JPL-TR-32-198.
Originator: I.. 8. Allen.

Status: Calculations, 1962.

Details: Fast neutrons, steady state, burner. Fuel: uranium as 93.5% en-
riched UF, dissolved in molten-salt mixture--70% UF,, 30% NaF. No mod-
'eré.tor.': Coolant: Li, which flows to engine. Core: sphere, 21 in. diameter,
40 liters vol. Core composition: 42% UF,, 18% NaF, and 10% Zr by volume.
Reflector: internal, Zr; external, 3 in. Be. Control: presumably negative
temperature coefficient. Power: 10 MW(t). \

Code: 0113 11 31106 44 627 711 84679 923 109
 

 

48

10.

11.

12.

13.

14.

15.

16.

17.

References

Unpublished report, H. K. Fergfison Company, Dec. 12, 1950.

Tbid., June 1, 1951.

Circulating.Moderator—-cAoola.nt Reactor for Subsonic Aircrait, HKF-112,
H. K. Ferguson Co., Aug. 29, 1951. Decl. July 8, 1964.

R. W. Dayton and J. W. Chastain, Hydroxides as Moderator coolants in
Power-breeder Reactors, BMI-746, Del., Battelle Memorial Institute,
May 26, 1952.

Unpublished report, ORNL, 1951. |
R. W. Schroeder and B. Lubarsky, unpublished report, ORNL, 1953.

E. S. Bettis, W. B. Cottrell, E. R. Mann, J_. L. Meem, and G. D; Whitman,
The Aircraft Reactor Experiment--Operation, Nuclear Sci. Eng. 2,
No. 6, pp. 841-853, Nov. 1957.

 

R. C. Briant and A. M. Weinberg, Molten Fluorides as Power Reactor
Fuels, Ibid., pp. 797-803. .

W. K. Ergen, A. D. Callihan, C. B. Mills, and Dunlap Scott, The All‘Cl‘aft
Reactor Experiment~-Physics, Ibid., pp. 826-840. .

A. M. Weinberg, Some Aspects of Fluid Fuel Reactor Development,

Ibid., 8, pp. 346-360, Oct. 1960.

E. S. Bettis, R. W. Schroeder, G. A. Christy, H. W. Savage, R. G. Affel,
and L. F. Hemphill, The Aircraft Reactor Experiment--Design and
Construction, Ibid., 2, No. 6, pp. 804-825, Nov. 1957.

W. B. Cottrell, H. E. Hungerford, J. K. Leslie, and J. L. Meem, Opera-
tion of the Aircraft Reactor Experiment, ORNL-1845, ORNL, Sept. 6,
1955. Decl. with del., Feb. 12, 1959.

W. B. Cottrell, ARE Design Data, CF-53-12-9, ORNL, Dec. 1, 1953.
Decl., Oct. 9, 1959.

E. S. Bettis and W. K. Ergen, Aircraft Reactor EXperiment, in Fluid
Fuel Reactors, J. A. Lane, H. G. MacPherson, and Frank Maslan, eds.,
Addison-Wesley Publishing Co., Reading, Mass., 1950, pp. 673-680.

Aircraft Nuclear Propulsion Project Quarterly Progress Report for

Period Ending March 10, 1952, ORNL-1227, W. B. Cottrell, ed., ORNL,
May 7, 1952.

C. B. Mills, ARE with Fuel-coolant in the Reflector, Y-F10-91, ORNL,
Feb. 27, 1952. -

C. B. Mills, The Fireball, A Reflector Moderated Circulating-fuel
Reactor, Y-F10-104, ORNL, June 20, 1952, *

 
 

e Ak e e A B st N b, o

e en A e e

e e Bk B Xl e

b it - 4 ek

o b i i e B M L e e it e,

IR

i
i
4
i
1
!
1

 

18.

19.
20.

21.

22.

23.

24.
25.
26.

27.

28.

29.

30.

31.

32.

49

A. P. Fraas, Design Characteristics of the Reflector-moderated Reac-
tor, in Aircraft Nuclear Propulsion Project Quarterly Progress Report
for Period Ending March 10, 1953, W. B. Cottrell, ed., ORNL-1515,
April 16, 1953.

A. P. Fraas and A. W. Savolainen, Unpublished report, ORNL, 1954.

W. B. Cottrell, W. K. Ergen, A. P. Fraas, F. R. McQuilkin, and J. L.
Meem, Aircraft Reactor Test. Hazards Summary Report, ORNL- 1835,
ORNL, Jan. 19, 1955. Decl., Oct. 9, 1959.

W. K. Ergen, Preliminary Design Data for a Circulating Fluoride-fuel
High-flux Reactor, CF-56-6-9, Rev. 2, ORNL, Jan. 28, 1958,

W. C. Cooley, Nuclear Aircraft Power Plants Utilizing Liquid Circu-
lating Fuel Reactors, APEX-135, ANP Project, General Electric
Company, June 1, 1953. Decl., June 9, 1961.

Unpublished reports, Pratt & Whitney Aircraft Division of United Air-
craft Corporation, 1953-54.

Ibid., 1957.

Ibid., 1954.

Reactor Study, PWAC-186, Pratt & Whitney Aircraft Division of United
Aircraft Corporation, July 15, 1957.

The P & WA Circulating Fuel Reflector-moderated Reactor, PWAC-189,
Pratt & Whitney Aircraft Division of United Aircraft Corporation,
Nov. 15, 1957.

Unpublished report, Pratt & Whitney Aircraft Division of United Aircraft
Corporation, 1956.

K. H. Puechl, Alternate Reactor Concepts for Aircraft Propulsion,
WKNL-42, Walter Kidde Nuclear lLaboratories, Inc., Feb. 1, 1955.
Decl. Jan. 19, 1965. |

J. H. MacMillan, C. B. Anthony, K. Guttmann, C. P. Martin, J. L.
Munier, and R. D. Worley, A Reflector Moderated, Circulating Fuel,
Aircraft Reactor, CF-53-9-84, ORSORT, Aug 14, 1953,

K. H. Dufrane, T. G. Barnes,_C Elcheldlnger W. D. Lee N. P. Otto,
C. P. Patterson, T. G. Proctor, R. W. Thorpe, and R. A. Watson,
Unpublished ORSORT term paper, Aug. 1957.

J. K. Davidson and W. L. Robb A Molten- salt Thorium Converter for
Power Production, KAPL-M-~JKD- 10 KAPL, Oct. 31, 1956. Decl.,
Feb. 26, 1957.
. b nibicheds:

 

50

33.

34.

35.

36.

37.

38.

39.

40.

4],

42.

43.

44.

Theodore Jarvis, L. L. Brown, D. L. Conklin, F. O. Ewing, R. O. Lowrey,
C. M. Rice, G. E. Tate, and R. E. Wascher, Reactor Design and Feasi-
bility Problem. Fused Salt Package, CF-53-10-26, ORSORT, Aug. 1953.
Decl., Jan. 20, 1964.

D. B. Wehmevyer, J. A. Bara, Jr., D. J. Cockeram, R. B. Donworth,

L.. B. Holland, R. S. Hunter, P. J. Mraz, W. Park, and W. L. Webb,
Study of a Fused Salt Breeder Reactor for Power Production, CF-53-10-
25, ORNL, Sept. 1953. Decl., July 5, 1957. | |

R. W. Davies, D. H. Feener, W. A. Frederick, K. R. Goller, I. Granet,
G. R. Schneider, and F. W. Shutko, 600 MW Fused Salt Homogeneous
Reactor Power Plant, CF-56-8-208, Del., ORNL, Aug. 1956. Decl. with
del., March 4, 1957.

J. T. Roberts, J. S. Lagarias, F. J. Remick, R. W. Ritzmann, J. O.
Roberts, W. J. Roberts, J. E. Schmidt, and P. R. Kasten, Reactor Pro-
ducing 3000°F Steam for Process Heat, CF-56-8-211, ORSORT, Aug. 6,
1956. Decl., Sept. 15, 1959.

. E. Romie and B. W. Kinyon, A Molten Salt Natural Convection Reac-
tor System, CF-58-2-46, ORNL, Feb. 5, 1958.

F. E. Romie and B. W. Kinyon, Design Study of a Molten Salt Natural
Convection Reactor, Trans. ANS, 1, No. 1, p. 64, 1958.

W. A. Box, C. S. Barnett, R. Bean, S. J. Ditto, ¥. A. Hazenkamp, L. R.
Pollack, and M. L. Winton, Feasibility Study of a Homogeneous, Fused
Salt, Molten Metal Cooled, Power Reactor System, Trans. ANS, 1, No. I,
pp. 63-64, 1958.

H. G. MacPherson, Survey of Low-enrichment Molten-salt Reactors,
CF-58-10-60, ORNL, Oct. 17, 1958. |

H. G. MacPherson, A Preliminary Study of a Graphite Moderated Molten
Salt Power Reactor, Unpublished report, ORNL, Jan. 13, 1959.

Molten Salt Reactor Project Quarterly Report for Period Ending Jan. 31,
1959, ORNL-2684, ORNL, March 17, 1957.

L. G. Alexander, B. W. Kinyon', M. E. Lackey, H. G. MacPherson, J. W.
Miller, ¥. C. VonderLage, G. D. Whitman, and J. Zazler, Experimental
Molten-salt-fueled 30-MW Power Reactor, ORNL-2796, ORNL, March 24,
1960.

A. L. Bock, E. S. Bettis, and W. B. McDonald, The Molten-salt Reactor
Experiment, Proc. Symp. on Power Reactor Experiments, 1, IAEA,
Vienna, 1962, pp. 247-292.

 

 
o e R s e B i it -

i .

 

e e h aih, abh  R

i
1
i
i
i
|
q
{
1
1
§

 

45.

46.

47.

48.

49.

H. G. MacPherson, Molten Salt Reactor Program Quarterly Progress
Report for Period Ending July 31, 1960, ORNL-3014, ORNL, pp. 1-23,
Dec. 22, 1960.

E. S. Bettis and W. B. McDonald, Molten-salt Reactor Experiment,
Nucleonics, 22, No. 1, pp. 67-70, Jan. 1964.

H. G. MacPherson, Molten-salt Reactors: Report for 1960 Ten-year-
plan Evaulation, Unpublished report, ORNL, July 25, 1960.

L. S. Allen, A Parametric Survey of Criticality-limited Fast Reactors
Employing Uranium Fluoride Fuels, JPL~TR-32-198, Calif. Inst. Tech.,
March 15, 1962.

W. R. Grimes, Molten Salts as Reactor Materials, Nuclear News, L
No. 5, pp. 3-8, May 1964.

51
 

52

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[ (T A P

BB o el e e

e G AR .

53

Chapter 3. Two-region Reactors

 

In the development of practical molten-salt-fueled reactors, two-
region reactors represent a more advanced stage of development than
one-region reactors. A two-region power breeder is the goal. Work on
two-region MSR's has been nearly all concentrated at Oak Ridge National
Laboratory. Most reactors described in this chapter have graphite-

‘moderated cores and fluoride fuels, although two concepts utilize chloride

fuels and no moderator.

In 1960, MacPherson described three possible types of reactor
construction for breeding.’

The first, the unit fuel tube, was believed to be the most practical.
The fuel passes through the reactor in graphite tubes, which are enclosed
by graphite moderator. The blanket, which contains a thorium salt, sur-
rounds the core. The blanket salt also passes through small passages in
the graphite as a coolant.

The graphite core shell consists of three blocks of graphite--a top
header, a center section, and a bottom header. The blocks should be of
graphite that is nearly impervious to the fuel solution. The three blocks
are either clamped, cemented, or held together by posts.

In the internally cooled reactor, the fuel is in graphite tubes ex-
tending through the moderator into the blanket. The tubes are connected
at each end to a header system, so that the fuel can be slowly circulated
to keep it uniform, to remove gaseous fission products, and to allow for
the fuel concentration to be adjusted for burnup. The heat is transferred
through the tube wall to the blanket salt, which thus acts as a coolant.
Graphite inserts in the tubes force the fuel to the outer portion of tubes
to increase heat transfer. A problem with this structure is that it is un-
likely that all of the 10,000 tubes needed for a 200 MW(e) reactor would
keep their integrity for a long reactor lifetime.

Hydroxide-moderated Reactor

 

In considering hydroxide-moderated two-region reactors, Dayton
and Chastain emphasized lithium-7 hydroxide and lithium-7 deuteroxide.?

- The spherical reactor, consisting of concentric shells of zirconium, has an

outer annulus, which contains a suspension of thorium oxide in heavy water.
This suspension serves both as a breeding blanket and as a reflector. The
specified core temperature of 1100°F was assumed to be high enough for
efficient power production but low enough to avoid excessive corrosion,
Breeding gains greater than 0.2 were postulated for reactors with either
lithium compound. The authors concluded that the reactor moderated with
 

54

1i"OD would be preferable. The reactor moderated with Li’OH has a lower-

cost moderator, smaller core radius, and smaller fuel requirements. These
advantages, however, are more than offset by those of the reactor moderated
with Li’OD. This reactor, because of its larger core radius, would offer a
larger breeding gain and would have improved heat-removal properties.
Also, 'although more fuel is required, it is a much smaller percentage of

the total mass than is required with the reactor moderated with Li’OH,

MIT Fast Reactor

 

A reactor designed in early 1952 that differed from most two-region
reactors was the MIT Fluid Fuel Fused-salt Reactor reported by Goodman
__(—?_,_1“:_?:L_l.3’4c Instead of fluorides, the fuel is uranium tetrachloride, which is
dissolved in a molten mixture of lead chloride and sodium chloride. The
higher thermal neutron-absorption cross section of the chlorides requires
this to be designed as a fast reactor. The blanket consists of uranium
tetrachloride containing depleted uranium. A lead reflector around the
semispherical core is used for shim control.

ORSORT Fast Breeder

 

A similar design for a fast reactor was employed by Bulmer, et al.,
students at the Oak Ridge School of Reactor Technology, in 1956.° Again
the fuel salt was a chloride instead of a fluoride, with both plutonium and
depleted uranium chlorides dissolved in sodium and magnesium chlorides.
The depleted uranium permits internal breeding. The blanket, a paste of
depleted uranium oxide powder in sodium, is divided into two regions by
a graphite moderator, which makes possible a smaller blanket. A molten-
lead reflector is between the core and the blanket. A graphite reflector
surrounds the second blanket region. A power of 260 MW(e) was given,

Internally Cooled Molten-~salt Reactor

 

Liackey calculated the nuclear characteristics for this reactor, which
uses molten fluorides of uranium, thorium, lithium, and beryllium as fuel.
The blanket and coolant are the molten-salt solutions Withouturaniunn.6
Graphite is the moderator. Cylindrical passages for coolant are in a tri-
angular lattice within the graphite, with the fuel in the annular space be-
tween the coolant tube and the graphite. The fertile salt is in a blanket
around the graphite. A different core arrangement, to reduce doubling
time, would use two concentric tubes with fuel in the annulus and coolant
flowing through the inner tube and outside the outer tube. This design was
for a reactor of high power--1563 MW(t).

 

 
g i s St i, AR e i i < o s e

i o S e it i ) it o - e SR il o Bt B o B aladi i, i bR |t v bR s B b ) S WAL e B, i R MM

55

Thermal Convection Reactor

 

Zasler has proposed a reactor, ' for which few details are given,
in which circulation of fuel could be either by natural convection for a
5-MW experimental reactor, or by fuel pumps for a 50-MW(t) pilot plant,
In the spherical reactor, which is surrounded by a blanket, fuel flows
from the reactor to a shell-and-tube heat exchanger.

Molten Salt Reactor (MSR)

 

 

 

et i b

This ORNL concept? ™! has been given many names: MSR; Interim
Design Reactor; Molten Fluoride Converter; Molten Fluoride Power Re-
actor; Reference Design Reactor; RDR;  MSR Reference Design; and
Homogeneous, Two-region, Molten-fluoride-salt Reactor. It underwent
several modifications, particularly in the core design. The fuel is highly
enriched uranium tetrafluoride in a eutectic of lithium, beryllium, and
thorium fluorides. Plutonium-239 has been suggested as an alternative
fuel.'?> A fertile blanket of the same composition as the fuel salt but with-
out the uranium surrounds the fuel region. The core has the shape of an
inverted pear. The fuel enters the bottom of the reactor, passes to the
top, goes to the primary heat exchangers, and returns to the reactor. The
reactor, designed for central-station power, would produce 260 MW(e).

Two-region, Graphite-moderated, Molten-salt Breeder Reactor

 

‘Members of the MSR Project staff selected a typical design for
calculation of reactor performance.'> The core is a graphite shell struc-
ture with a single cylinder of graphite as the main part and two graphite
end pieces., The three pieces are held together by rods of INOR-8 or other
alloy. The fuel, uranium-233 in a molten fluoride carrier containing
thorium, passes downward through channels in the core then back upward
to an external heat exchanger. The blanket, which contains thorium in
molten salt, is kept at a slightly higher pressure than the core. The power
would be 125 MW(t).

" Molten Salt Breeder Reactor (MSBR)

 

A reactor considered in 1961 in a comparison of homogeneous re-
actors for thorium breeding was the MSBR,*® which was based on a
design by MacPherson in 1959.! In both the 1959 and 1961 designs, the
fuel is a solution of enriched uranium tetrafluoride in LiF-BeF,, the blan-
ket surrounding the core is molten salt containing thorium, and the fuel
passes through tubes in the graphite moderator. In the 1959 design, the
core is of the unit-fuel-tube construction, in which one-pass graphite tubes
go through the graphite moderator. In the 1961 design, the core is con-
structed of graphite prisms, machined at the corners to form vertical
 

 

56

passages, into which two-pass bayonet graphite tubes for fuel flow are
inserted. Both designs are for use in groups of reactors for power
production. The 1959 concept is for three reactors producing 333 MW(e)
each, the 1961 concept for two reactors producing 500 MW(e) each.

 

 
1 T TR Y S T e Y TS TR T Sy ey 1 Y e s oy ey ey - - v T e T T T ORI R Y ey T

57

DATA SHEETS
TWO-REGION REACTORS

¥ w - Y ) 3 L

T TR TR T e e TUUTIR T T T TN YT Y T TR T W TR
 

58

 
e Dot i il R ey e e e i R SO b A it RACT R

|
|
i

 

59

No. 1 Reflected Homogeneous Reactor Moderated by
Lithium-7 Hydroxide

 

 

Battelle Memorial Institute

Reference: BMI-746.
Originators: R. W. Dayton and J. W. Chastain.

Status: Design calculations, 1952.

Details: Thermal neutrons, steady state, converter. Fuel-moderator-
coolant: Homogeneous suspension of U?**0, in LiOH. Minimum fuel for
criticality: about 2.5 kg U**’. Reflector-breeding blanket: circulating
suspension of ThO, in D,0, about 1 g/cm3, in spherical annulus. Blanket
thickness: 60 cm. Core contained by 2 concentric Zr shells, 0.5 cm

thick. Shells separated by air space 1 c¢m thick, which is thermal barrier.
Zr foil in air space to improve thermal barrier. Core temperature: 1100°F.
Blanket temperature: 250°F. Breeding gain: more than 0.2.

Code: 0311 17 31312 45 637 756 . 84677 941 101

No. 2 Reflected Homogeneous Reactor Moderated by
‘Lithium-7 Deuteroxide

 

 

Battelle Memorial Institute

Reference: BMI-T746.

Originators: R, W. Dayton and J. W. Chastain.

Status: Design calculations, 1952.

- Details: Same as Data Sheet No. 1, except that Li’OD is used instead of

Li’OH.
Code: 0311 17 31312 45 637 756 84677 941 101

No. 3 MIT Fluid Fuel Fused-salt Reactor
MIT

 

References: MIT-5000, pp. 12-19, 25-150; - MIT-5001, pp. 11-37.
Originators: Clark Goodman et al. " |

- Status: Design adopted for study of nuclear problems of nonagqueous

fluid-fuel reactors, 1952.

- Details: Fast neutrons, steady state, converter. Fuel-coolant: mixture of
molten UCl, (22% U235), PbCl,, and NaCl. No moderator. Reflector: Pb,

immediately surrounding semispheri¢al core. Fertile material: blanket
of depleted U (containing 0.3% U??*%) as UCl,, around reflector. Fuel-coolant
is circulated to external heat exchangers. Structural materials: Ni or Fe.
Control: reflector for shim control; negative temperature coefficient.
Conversion ratio of U?*® to Pu®*®*?: about 1.15. No power given.
Code: 0111 11 31211 43 627 735 84679 941 108

82X88

 
 

60

No. 4 Fused Salt Fast Breedér Reactor
ORSORT

 

Reference: CF-56-8-204, Del.
Originators: J. J. Bulmer et al.

- Status: Design and feasibility study, 1956. ORSORT term paper.

Details: Fast neutrons, steady state, breeder. Fuel coolant: circulating
mixture of molten NaCl, MgCl,, PuCl;, and U**8Cl,; PuCl; is assumed to
be in solution with UCl;; depleted U is for internal breeding. Fertile
material: blanket of paste of depleted UO, powder in Na. Fuel contained
in a 72.5 in. ID, nearly spherical, vessel, tapered at the top and bottom to
24 in. for pipe connections. Vessel is of Ni-Mo alloy-clad stainless steel.
Reflector: 1 in. liquid Pb immediately surrounding core. Around reflector
is blanket, under 100 psi pressure. Stationary blanket is divided into
2 regions by a 54-in. stainless steel-clad graphite moderator. An 8 in.
graphite reflector surrounds second blanket region. Use of graphite mod-
erator in middle of blanket reduces size of blanket necessary. Blanket
cooled by Na flowing through tubes. Fuel enters reactor core at bottom
at 1050°F and leaves through the top at 1350°F, then is pumped through an
external loop and tube side of Na heat exchanger. Control: mainly through
negative temperature coefficient. Shim control by changing level of Pb
reflector. Estimated breeding ratio: 1.09. Power: 700 MW(t), 260 MW(e).
Code: 0112 11 31211 46 627 755 84679 941 108

82188 '

No. 5 Internally Cooled Molten-salt Reactor

 

ORNL

Reference: CF-59-6-89.

Originator: M. E. Lackey.

Status: Calculations of nuclear characteristics, 1959,

Details: Thermal neutrons, steady state, breeder. Fuel: U233F4 in

molten mixture-- ThF,-LiF-BeF,. Two compositions: ThF;--7, 13%;
1iF--67.25, 71%; BeF,~-25.75, 16%. Coolant: molten fuel carrier.
Moderator: graphite. Fertile material: blanket, 2.5 ft thick, of molten
fuel carrier. Core: cylinder, 19.15 ft diameter, 19.15 ft high, surrounded
by blanket, in INOR-8 pressure shell, 1.5 in. thick. Cylindrical passages
for fuel and coolant in moderator, parallel to core axis, in triangular lattice.
Coolant tube in each passage. Fuel in annular space between coolant tube
and moderator. No reflector. Control: negative temperature coefficient.
Power: 1563 MW(t). Doubling times: with 7 mol % Th?32, 27.5 years; with
13 mol %, 22.5 years. The Pa®®*® formed could be diluted by mixing the fuel-
carrier salt with the blanket and coolant salts. This should reduce doubling
times. Further reduction possible by different core arrangement: 2 con-
centric tubes, with fuel in the annulus and coolant flowing through the inner

tube and on outside of outer tube.
Code: 0312 12 31211 44 627 746 84679 931 | 106

 
A e

 

 

61

No. 6 5 MW (50 MW) Thermal-convection Molten Salt Reactor
ORNL

 

References: ORNL-2551, pp. 49-51; ORNL-2626, p. 48.

Originator: J. Zasler.

Status: Proposal for preliminary design, 1958.

Details: Thermal neutrons, steady state, breeder. Spherical core, 5 ft
diameter, surrounded by 0.5 ft blanket. Fuel-coolant: flows at 158 gpm
through reactor and shell-and-tube heat exchanger, cooled by Na, entering
the heat exchanger at 1210°F and leaving at 1010°F. 5-MWI{t) experimental
reactor could be converted to 50-MW(t) pilot plant by adding fuel pumps
(to increase flow to 1515 gpm) and increasing capacity of the fuel pumps.
Code: 0312 1X 31211 4X 627 XX 84679 9XX 106

No. 7 Molten Salt Reactor (MSR; Interim Design Reactor; Molten
Fluoride Converter; Molten Fluoride Power Reactor;
Reference Design Reactor, RDR; MSR Reference
Design; Homogeneous, Two-region Molten-
fluoride-salt Reactor)

ORNL

 

 

 

 

 

References: ORNL-2751; Proc. 2nd U.N. Int. Conf. 9, 1958, pp. 188-201;
Lane; et al., Fluid Fuel Reactors, pp. 657, 681-696.
Originators: H. G. MacPherson et al..

 

 

- Status: Conceptual design, 1959;_w-o_rk continuing.

 

Details: Intermediate neutrons, steady state, converter. Fuel-coolant-
moderator: mixture of 90% enriched U**®F, with LiF-BeF,-ThFy; alternate
fuel: Pu®®? instead of U?*®. Neutrons with this fuel are thermal.

Fertile material: 2-ft-thick blanket, of ThF,; as eutectic of LiF-BeF,-ThFy,,
completely surrounding fuel region. Both fluids circulated to external shell-
and-tube heat exchangers, where heat is transferred to Na. Several core
configurations considered: 1) straight-through flow, 2) concentric inlet
(inner pipe)-outlet (outer annulus) flow, 3) concentric inlet (outer annulus)-
outlet (inner pipe) flow. Core: inverted-pear-shaped, Ni-Mo alloy, 8-ft in
diameter (6-ft in early designs), surmounted by expansion chamber con-
taining fuel pump. Fuel temperature: approximately 1200°F. Control: neg-
ative temperature coefficient. Blanket produces U?*? at conversion ratio of
0.6 to 0.8, depending upon processing. Reactor designed as central-station
power plant to produce 600 MW(t), 260 MW(e).

- Code: 0211 17 312 44 627 746 84677 931 101

 

0311 46
 

 

62

No. 8 Two-region, Graphite-moderated Molten-salt Breeder Reactor

ORNL

 

Reference: ORNL-~2799.

- Originators: MSR project staff.
Status: Typical design for calculation of performance, 1959, |
Details: Thermal neutrons, steady state, breeder. Fuel-coolant: U233,

presumably as UF,, dissolved in molten LiF-~BeF,-ThFy,.

- Moderator: graphite. Reflector: graphite. Fertile material: 30-in.

blanket of 13 mol % Th in molten LiF-BeF,-ThF,. Core: main part
made from single cylinder of graphite, about 5 ft diameter, 5 ft long,

and two end pieces. Three pieces held together by rods of INOR-8 or
other alloy. Parallel vertical channels for fuel flow, 1/2 in, max. radius.
Molten fuel salt passes downward through core then back upward to ex-
ternal heat exchanger. Blanket, which surrounds core, is kept at slightly
higher pressure. Pressure: 1000 psi max. Control: presumably neg-
ative temperature coefficient. Power: 125 MW(t).

- Code: 0312 12 31211 45 627 746 84679 941 104

No. 9 Molten Salt Breeder Reactor (MSBR)
ORNL

 

References: CF-61-3-9, pp. 68-81; CF-61-8-86.
Originators: H. G. MacPherson, L. G. Alexander, et al.

Status: Reference design for evaluation of thorium breeder reactors, 1961.
Details: Near-thermal neutrons, steady state, breeder. Fuel-coolant: UFy

in LiF and BeF,. Moderator: graphite. Fertile material: blanket of ThFy,
in LiF and BeF,. Reflector: graphite. Core cylinder 7%-ft x 7%-ft. of
graphite moderator prisms, 7.5 in. square by approx. 7 ft long. Corners
machined to form vertical passages of about 5 in. diameter circular cross-
section.  Fuel flows through 90 2-pass bayonet graphite tubes inserted into
these vertical passages. Core surrounded on all sides by 3-ft blanket, which
is then surrounded by the 1-ft graphite reflector. Heat exchanger and fuel
pump mounted directly above the reactor. Maximum fuel temperature:
1300°F; minimum 1125°F. Operating pressure: 100 psi. Control: through
negative temperature coefficient. Power station would include 2 such re-
actors to produce 2364 MW(t) or 1000 MW(e)--1182 MW (t) and 500 MW(e)

“each.

Code: 0312 12 31211 44 627 746 84679 941 104

 
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63

No. 10 Graphite-moderated, Circulating Fuel, Molten Salt Reactor Plant
ORNL

 

Reference: CF-59-12-64 Rev.

Originator: H. G. MacPherson.

Status: Reference design, 1959.

Details: Thermal neutrons, steady state, breeder. Fuel-coolant: U%°F,
in LiF-BeF,, at minimum temperature of 975°F and maximum of 1300°F,
circulating at 20 fps. Moderator: graphite. Fertile material: blanket
of Th, presumably as ThF,, in LiF-BeF,. Core: 5.3 ft in diameter;
blanket: 2.5 ft thick. Graphite moderator core contains graphite tubes
through which fuel passes (unit fuel tube construction). Blanket salt also
passes through small passages in moderator for cooling. Fuel inlet and
outlet at bottom of core, although straight-through flow is possible.
Control: presumably by negative temperature coefficient. Possible
breeding ratio: 1.06. Each of 3 reactors in plant produces 815 MW(t),
333 MW(e). -

Code: 0312 12 31211 44 627 746 84679 931 104
 

64

10.

11.

12.

13,

References

H. G. MacPhe.rson, Molten-salt Breeder Reactors, CF-59-12-64 Rev.,
ORNL, Jan. 12, 1960. |

R. W. Dayton and J. W. Chastain, Hydroxides as Moderator-coolants in
Power-breeder Reactors, BMI~746, Del., BMI, May 26, 1952,

Clark Goodman, J. L. Greenstadt, R. M. Kiehn, Abraham Klein,

M. M. Mills, and Nunzio Tralli, Nuclear Problems of Nonaqueous
Fluid-fuel Reactors, MIT-5000, MIT, Nucl. Eng. Project, Oct. 15, 1952.
Decl., Feb. 28, 1957.

George Scatchard, H. M. Clark, Sidney Golden, Alvin Boltax, and
Reinhardt Schuhmann, Jr., Chemical Problems of Nonaqueous Fluid- -
fuel Reactors, MIT-5001, MIT, Nucl. Eng. Project, Oct. 15, 1952.
Decl., Oct. 7, 1959.

J. J. Bulmer, E. H. Gift, R. H. Holl, A. M. Jacobs, S. Jaye, E. Koffman,
R. L. McVean, R. G. Oehl, and R. A. Rossi, Fused Salt Fast Breeder,
CF-52-8-204, Del., ORSORT, Aug. 1956,

M. E. Lackey, Internally Cooled Molten-salt Reactors, CF-59-6-89,
ORNL, June 22, 1959,

Molten-salt Reactor Program Quarterly Progress Report for Period
Ending June 30, 1958, ORNL-=-2551, ORNL, Sept. 24, 1958.

Molten-~salt Program Quarterly Progress Report for Period Ending
October 31, 1958, ORNL-2626, ORNIL, Jan. 23, 1959.

L. G. Alexander, D. A, Carrison, H. G. MacPherson, and J. T. Roberts,
Nuclear Characteristics of Spherical, Homogeneous, Two-region,
Molten-fluoride-salt Reactors, ORNL-2751, ORNL, Sept. 30. 1959.

H. G. MacPherson, H. W. Savage, R. C. Briant, W. H. Jordan,

L. G. Alexander, W. F. Boudreau, E. J. Breeding, W. G. Cobb,

B. W. Kinyon, M. E. Lackey, L. A, Mann, W. B. McDonald,

J. T. Roberts, A. W. Savolainen, E. Storto, F. C. VonderLage,

G. D. Whitman, and J. Zasler, Molten Fluoride Power Reactor,
Proc. 2nd U.N. Int. Conf. on Peaceful Uses of Atomic Energy, Geneva,
1958, 9, pp. 188-201, United Nations, New York, 1958,

L. G. Alexander, B. W. Kinyon, M. E. Lackey, H. G. MacPherson,
L. A. Mann, J. T. Roberts, F. C. VonderLage, G. D. Whitman, and
J. Zasler, Conceptual Design of a Power Reactor, in Fluid Fuel

Reactors, J. A. Lane, H. G. MacPherson, and Frank Maslan, eds.,
Addison-Wesley Publishing Company, Reading, Mass., 1958, pp. 681-696.

Ibid., p. 657.

Molten-salt Reactor Program Quarterly Progress Report for Period
Ending July 31, 1959, ORNL-2799, ORNL, Oct. 8, 1959.

 
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65

14. L. G. Alexander, W. L. Carter, R. H. Chapman, B. W. Kinyon,

15.

J. W. Miller, and R. Van Winkle, Thorium Breeder Reactor
Evaluation. Part I. Fuel Yield and Fuel Cycle Costs in Five Thermal
Breeders, CF-61-3-9, ORNL, May 24, 1961,

W. L. Carter and L. G. Alexander, Thorium Breeder Reactor
Evaluation. Part I. Fuel Yields and Fuel Cycle Costs of a Two-
region, Molten Salt Breeder Reactor, CF-61-8-86, ORNL,

Aug. 18, 1961.
 

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