ANL-6909.txt

ANL-6909 | ANL-6909

4Ad MASTER

Argonne JAational Laboratorp

CATALOG OF NUCLEAR REACTOR CONCEPTS

Part I. Homogeneous and
Quasi-Homogeneous Reactors

Section Il . Reactors Fueled with Homogeneous
Aqueous Solutions and Slurries

by

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

DISCLAIMER

This report was prepared as an account of work sponsored by an
agency of the United States Government. Neither the United States
Government nor any agency Thereof, nor any of their employees,
makes any warranty, express or implied, or assumes any legal
liability or responsibility for the accuracy, completeness, or
usefulness of any information, apparatus, product, or process
disclosed, or represents that its use would not infringe privately
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otherwise does not necessarily constitute or imply its endorsement,
recommendation, or favoring by the United States Government or any
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Government or any agency thereof.
DISCLAIMER

Portions of this document may be illegible in
electronic image products. Images are produced
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ANL-6909

Reactor Technology
(TID-4500, 32nd Ed.)
AEC Research and
Development Report

ARGONNE NATIONAL LABORATORY
9700 South Cass Avenue
Argonne, Illinois 60440

CATALOG OF NUCLEAR REACTOR CONCEPTS

Part I. Homogeneous and Quasi-Homogeneous Reactors
Section IT, Reactors Fueled with Homogeneous
Aqueous Solutions and Slurries

by

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

Technical Publications Department

July 1964

Operated by The University of Chicago
under
Contract W=31-109-eng-38
with the
U. S. Atomic Energy Commission
TABLE OF CONTENTS

PrEface ° © L] L4 [} o [ ] - - L] [ ] L] L J LJ L] »> L d - L] - L J L4 * - L-J L] (] L J

SECTION II. REACTORS FUELED WITH HOMOGENEOUS AQUEOUS SOLUTIONS
SLURRIES

Chapter 1. Introduction - - o =« o o « « o s« o« o o s o o o
Chapter 2. Research and Testing Reactors . ¢ o = o o o o
Chapter 3. Non-Boiling Reactors for Power and Breeding .

Chapter 4. Boiling Reactors for Power and Breeding . . .

145
11

PREFACE

This report continues the catalog of concepts for nuclear reactors
that was begun in ANL-6892. As in the previous report, the material is
divided into chapters, each with text and references, plus data sheets

that cover the individual concepts.

J.H.M,
July 1, 1964

iii
PART I HOMOGENEOUS AND QUASI-HOMOGENEOQOUS REACTORS

SECTION II REACTORS FUELED WITH HOMOGENEOUS AQUEOUS
SOLUTIONS AND SLURRIES

Chapter 1 Introduction

The use of aqueous solutions and slurries to provide an even dispersion
of fuel atoms among moderator atoms was one of the earliest suggestions for
nuclear reactors.l The concepts in this section follow the early idea and
are thus distinguished by a common characteristic: fuel dissolved or
suspended to form a homogeneous aqueous solution, colloidal dispersion, or
slurry.

These concepts have been investigated by many workers. Especially
valuable as sources of detailed information are reviews by Lane et al.ls?
For engineering details of specific reactors, see the tabulations referred
to in the General Introduction to the catalog.

The title of the section defines generally how concepts have been
chosen for inclusion, but some explanation may clarify why certain related
concepts were excluded.

The term "homogeneous aqueous'" would strictly mean that the core must
be an aqueous solution or suspension throughout, with no solid moderator.

A few concepts in which the fuel solution is in contact with such a
moderator are included here, however, because of their small number and
their close similarity to the other concepts included. 'Aqueous" includes
light water, heavy water, and any water-containing solution or suspension.
Molten salts and liquid-metal solutions or slurries are discussed in
Sections III and IV respectively. Solutions and suspensions are here
defined to jinclude particle sizes from those of true solutions to particles
that can be made into a slurry--up to 500 mu in diameter. Aqueous
suspensions and slurries are included in this section because their
properties and behavior in a reactor resemble those of solutions and
because a well-agitated suspension or slurry is virtually homogeneous.
Fuels for these reactors thus range from pure Newtconian liquids to paste
fuels.

As in other sections, subcritical assemblies, exponential

experiments, test loops, and similar concepts will not be treated in a

separate chapter because they do not strictly fit the definition of reactors.
They will, however, be discussed wherever pertinent., The various critical
experiments that have been part of the development of reactors are examples.

Some reactors that embody concepts that strictly fit the definition of
homogeneous aqueous solutions or suspensions are discussed elsewhere because
they have some other distinctive feature.

Examples are pulsed and excursion reactors. The Russian pulsed reactor
is a high-flux research reactor, with an aqueous solution of enriched
uranium salt as fuel. The Solution Type Pulse Reactor (STPR) of Atomics
International (AI) has highly enriched uranyl sulfate solution for fuel.
Al's Kinetic Experiment Water Boiler (KEWB) is, as the name designates, a
"water boiler,”" but it is also a pulsed reactor that can sustain large power
excursions. All of these reactors will be described in Part III of the
catalog.

The concepts in this section are discussed in the three remaining
chapters:

Chapter 2 Research and Testing Reactors

Chapter 3 Non-Boiling Power and Breeder Reactors

Chapter 4 Boiling Power and Breeder Reactors
Most of the research and testing reactors in Chapter 2 are of the misnamed
"Water Boiler" type--the water does not boil. They were so named because of
the boiling appearance caused by radiolytic gas bubbles. Some others,
however, are included. Chapters 3 and 4 cover reactors designed to yield
power and to breed fuel. They differ in whether or not heat is removed by
boiling in the core.

Obviously, divisions between chapters are not always sharp-—-as between
Chapters 2 and 3. To a large extent the divisions are according to the
purpose of the reactor.

Table I3 lists the most significant homogeneous aqueous reactors with
thelr characteristics and applications. This section generally follows the

pattern used in the table.

Early History

Aqueous solutions of uranium salts were used to investigate atomic
fission not long after the discovery, by Hahn and Strassmann in 1938, that

neutron bombardment could break up the uranium nucleus, and after the

TABLE I Homogeneous Reactor Types and Applications

Reactor designation

Power level range,

Fuel solution or suspension

Application

Mw heat
Water boiler 0-0.05 Enriched U0,SO, or UO,(NO.,), in H,O University nuclear research
2 4 27 372 2
and training
Homogeneous 800--2000 Enriched UOZSO4 in DZO Nuclear research at ultra-
research reactors high thermal-neutron fluxes
U235 burners 40-500 Enriched UOZSO4 in HZO or DZO Small- to large-scale power
plants in high-fuel-cost
locations; mobile power
plants
LAPRE type power 1-100 Enriched UO., dissolved in 60 wt.Z Remotely located small- and
reactors phosphoric™acid intermediate-scale power
Enriched U0, dissolved in 95 wt.Z% plants
phosphoric acid
One-region power 500~1000 Slightly enriched UO3 in DZO Large-scale power production
converters
One-region Pu 1000-2000 Slightly enriched UOZSO4 in DZO Dual-purpose power plus plu-
producer [with or without added Lil(S0,)] tonlum production
Two-region Pu 500-1500 Enriched UOZSO4 in DZO (core) Dual-purpose power plus plu-
producer Depleted U0,S0, in D,0 (blanket) tonium production
One-region thorium 500~-1500 Enriched U235 or U233 oxide plus Large-scale power production
breeder ThO2 in DZO
Two-region thorium 200-1000 Enriched U232 or U233 as UOZSO4 Large-scale power production
breeder, solution in D.0 (core) plus ThO. in D.O and U233 breeding or u235
core 2 ¢ P 2 2 to U233 conversion
(blanket)
Two-region thorium 200--1000 Enriched U235 or U233 oxide plus Large-scale power production

breeder, slurry
core

ThO2 in DZO (core) plus ThO

in D20 (blanket)

2

breeding or U235
conversion

and U233
interpretation of the significance of their work by Meitner and Frisch."

Von Halban, Joliot, and Kowarski placed a uranium solution in a large vessel,
in the center of which was a source of neutrons. Around it they placed
devices to detect neutrons and count them as they were formed. When they
substituted for the uranium solution another that was very similar but
contained no uranium, they found that many more neutrons were produced when
the neutron source was placed in the solution of uranium.%,®

This experiment demonstrated that neutron multiplication would take
place in an aqueous solution of uranium salt; it was among the first sub-
critical or exponential reactors. Neutrons from a neutron source were
multiplied, but there was not a self-sustaining or chain reaction because
the neutron source was needed to maintain the reaction as well as to start it.

Von Halban and Kowarski continued their experiments in England using a
homogeneous suspension of uranium oxide in a vessel immersed in about a ton
of heavy mineral oil to serve as a reflector. Although this system was not
large enough to be itself capable of maintaining a chain reaction,
measurements made enabled the experimenters to show that a self-sustaining
system would be possible with between 3 and 6 tons of heavy water; they
also showed that, with enriched uranium, a self-sustaining reaction would
be possible in light water.

In this experiment, a suspension of uranium oxide and heavy water was
contained in a spinning sphere that kept the powder well mixed with the
fluid. It provided clear evidence that in a sufficiently large system a
chain reaction would be possible.’ Because the only feasible homogeneous
reactor using natural uranium would be one moderated with heavy water,
and sufficient amounts of D20 did not become available until 1943 for use
In reactors, early interest in homogeneous aqueous reactor systems was
purely academic. Even after the atomic energy program was well under way,
work was concentrated on heterogeneous reactors.! An important advantage
of heterogeneous reactors Is that fuel can be clad to prevent corrosion,
attrition, and escape of fission products.

Interest in homogeneous reactors was revived early in 1943, when a
supply of heavy water in U. S. and Canada was expected. However, slurry
reactors took precedence over the solution type because enriched uranium
was not then available in sufficient quantity. Furthermore, the only known

soluble salts of uranium of sufficiently low cross section to enable the

design of a reactor of feasible size and heavy water requirement were
uranyl fluoride and uranium hexafluoride. These compounds were considered
but rejected, principally because of hydrolysis of the UF6, corrosion,
meager solubility data, precipitation, and instability under radiation. A
second reason for the choice of a slurry reactor was the evidence that
decomposition of the heavy water would be more severe in a solution than
in a slurry. In the solution, fission fragments would be formed in
intimate contact with the water. They would, however, be formed inside a
solid particle in a slurry.! Finally, a slurry would have a higher fuel
loading than a solution. In spite of the emphasis on heterogeneous and
slurry reactors, however, many concepts for homogeneous aqueous reactors

were considered for research, power, and breeding.
References

J. A. Lane, H. G, MacPherson, and Frank Maslan, Fluid Fuel Reactors,

Addison-Wesley Publ., Co., Reading, Mass., 1958.

J. A. Lane, "Aqueous Fuel Reactors," Chapter 19 in Reactor Handbook,

Vol. IV, Engineering, 2nd Ed., Stuart McLain and J. H. Martens, eds.,

Interscience Division, John Wiley and Sons, New York, 1964.

Ref. 1, p. 12.

Gordon Dean, Report on the Atom, 2nd Ed., Knopf, New York, 1957,
pp. 242-83.

H. Von Halban, Fredric Joliot, and L. Kowarski, "Liberation of Neutrons

in the Nuclear Explosion of Uranium," Nature, 143, No. 3620, pp. 470-71,
March 18, 1939.

L. Bertin, Atom Harvest , Secker and Warburg, London, 1955, pp. 42-44.
Ref. 6, p. 57.

Chapter 2 Research and Testing Reactors

The reactors taken up in this chapter are characterized by their
purpose—-—-to give experimental data, primarily through the nuclear radiation
they produce. They are thus distinguished from those experimental reactors
that are used primarily to obtain data for the design of a power reactor,
Such experimental reactors will be described in connection with the
pertinent power reactors. The USAEC Division of Technical Information has
defined a research reactor as '"any reactor whose nuclear radiations are
used primarily as a research tool for basic or applied research regardless
of operating power level. May include facilities for testing reactor
materials.”! Most of these research reactors are of the "water-boiler"

type, but there are several others.

Zero-Power Reactors

"zero power" deserve to be

A few reactors of extremely low power or
treated separately, rather than with power reactors as the subcritical
assemblies and critical experiments are discussed. They are reactors
useful in their own right rather than as part of the development of another
reactor. Zero-power reactors have been defined by the National Research
Council as experimental nuclear reactors operated at low neutron flux and
at a power level so low that not only is no forced cooling required but

also fission product activity in the fuel is sufficiently low to allow the

fuel to be handled after use without serious hazard.?

NASA Zero-Power Reactor

This reactor was planned for performing critical experiments,
measuring reactivity effects, serving as a neutron source, and being a
training tool. Varying core configurations of solutions of highly
enriched uranyl fluoride in light water would operate at a maximum power
of 10 watts.3 Two versions, NASA ZPR-I and NASA ZPR-II have been listed,
with the second described as an enlarged version of the first.“ The
power levels, however, apparently are the same. In December 1963, both
were being used for critical experiments, with 1959 being given as a
startup date for ZPR-I and 1963 for ZPR-II.! ZPR-I was described as
"eritical experiments for the NASA Test Reactor," ZPR-II as "NASA
Zero-Power Reactor II, solution type critical." Plans were for operating

both in the same building, but with only one reactor operating at a time,
Some of the data obtained from this reactor were used in a computer
program to compute the age of fission neutrons in water.?
The design of this reactor utilized earlier work at Oak Ridge on the

criticality of aqueous solutions of uranium-235 salts.5s7

Highly Enriched Homogeneous Reactor

This reactor,8 described by Brown, et al, of Westinghouse APD, is of
lower power than that of NASA. It contains an extremely dilute aqueous
solution of uranyl nitrate, in which the enrichment is greater than 90 percent.
It is contained in a stainless-steel tanmk, 72 inches high and 36 inches in

diameter.

Proserpine

This French "reactor" is actually a critical facility, but is in the
form of a small water boiler and thus is discussed here. It has been
described in several references.?»10 It is designed to study criticality
problems of plutonium solutions, and it utilizes a fissionable solution of
, in 0.5N H

plutonium sulfate, Pu(SO SO4 in light water. The reflector

4)2 2

is beryllium oxide and graphite.

Water Boilers

A "water boiler" is a small reactor with a core consisting of a
container of fuel solution. It is equipped with controls, reflectors,
heat-transfer system, and other auxiliaries as required. The fuel is a
soluble salt of uranium or plutonium; the moderator-solvent is light or
heavy water.

The core is a stainless-steel sphere, about one foot in diameter,
containing the fuel solution.!! Because a natural uranium light water
solution cannot be made critical, the uranium is enriched in U235. The
fuel solution is cooled by water circulating through coils inside the core.
Normally this cooling keeps the solution temperature below about 80°C and,
because the operating pressure is near atmospheric, no actual boiling
occurs. One water boiler, SUPO, was, however purposely operated under
boiling conditions in a stability study.!2

A fuel-handling system adds fuel solution to the core and recovers
solution expelled by any operation above design power. Hydrogen and

oxygen evolved in operating at relatively high power are catalytically
recombined and the water is returned to the core. The core is surrounded

by a graphite or beryllium oxide neutron reflector with a concrete shield
around the reflector. Control is by neutron-absorber rods usually containing
boron or cadmium, with passages or thimbles for the rods in the core or
reflector. Usually the reactors have experimental facilities, such as
thermal columns and beam tubes.l!!

The water boiler has evolved through many stages, 3-18 and various

types are now marketed by several manufacturers. 19520

The LOPQO Series

This water boiler, or LOPO-HYPO-SUPO concept, is the prototype for all
water-boiler reactors. It was important in World War II in that it
provided vital data for the atomic bomb project.

In all these reactors, power and breeding are not primary objectives.
LOPO originally was built '"to gain experience in the operation and control
of a chain-reacting assembly while using a minimum of active material...
Its main job was the mass production of neutrons..."?l

The "water boiler" at Los Alamos was planned in 1943 to test certain
theoretical calculations!3 when an enriched uranium supply seemed possible.
R. F. Christy was responsible for the critical-design calculations for this
reactor.1*»15,22 Christy's reactor design consisted of a noncorrosive
stainless-steel shell, one foot in diameter, containing enriched uranium
or plutonium as a water solution of a suitable salt, such as uranyl sulfate
or nitrate, with a neutron reflector of beryllium oxide bricks occupying
a cube of about four feet on an edge. The fuel solution circulated by
convection, and about 10 kw(t) was extracted by a coiled tube carrying
cold water through the inside of the reactor. Control was exercised through
a vertically moving control rod.

Preliminary plans and calculations for such an operation were completed
by September 1943, and a critical experiment, LOPO-CX, was carried out in
1943-44, Ten kilowatts was arbitrarily chosen as the operating level. It
appeared, however, more advisable to construct first a reactor of much lower
power, known as the Low Power Water Boiler (LOPO), because it eliminated
heavy shielding requirements and minimized the problems of uranium compounds
going out of solution, gas evolving from decomposition of the water by
fission fragments, and fission fragments contaminating the solution. The

plans for LOPO were completed in November 1943, assembly proceeded through
the spring of 1944, and the assembly went critical in 1944.13,2!

LOPO was one of the first aqueous-solution reactors using enriched fuel.
It used 14.5 percent enriched uranium (565 grams of uranium-235 as uranyl
sulfate) dissolved in ordinary water. The solution was contained in a
Type-347 stainless-steel sphere one foot in diameter, with a beryllium oxide
reflector and a cadmium control rod. The lack of a shield and cooling system
limited the heat power level of LOPO to 50 milliwatts.

LOPO was later modified to become the High Power Water Boiler (HYPO),
which went critical in December 1944, It operated at a peak power of about
6 kw. Major modifications included the addition of a shield and an internal
cooling system consisting of a single coil, which carried cooling water
through the core. Additional experimental holes were added, the uranyl
sulfate fuel solution was replaced by uranyl nitrate because an extraction
method to remove fission products was known only for the nitrate solution
at that time, and the control system was made more elaborate.21,23,2"4

Because higher neutron fluxes and more research facilitles than were
avallable from HYPO were desired, the reactor was further modified
starting in April 1949. Completed in March 1951, it became SUPO, the Super
Power Water Boiler. Major changes included an increase in operating power
level from 5.5 kw to a minimum of 45 kw by replacing the cooling coil by
three stainless-steel tubes, replacing the beryllium portion of the
reflector by graphite, changing the reactor solution from 15 percent
enriched uranyl nitrate to one of 88.7 percent enrichment, and adding new

experimental holes and vertical control rods.?3,25,26,27

Atomics International Reactors

Atomics International's 5-watt Water Boiler Neutron Source (WBNS) was
built and first operated at Downey, California; it went critical in 1952.
It was later converted to a 2.5-kw reactor, AE-6. The AE-6 was used as a
source of thermal neutrons to study nuclear behavior of heterogeneous
subcritical assemblies and the irradiate small samples in neutron fluxes
up to 1011 ne.utrons?./c:m?'/sec:.28"33

The fuel solution was contained in a one-foot diameter stainless-steel
sphere, initially with np cooling system (as in LOPO) because of the very
low power level, 1 to 2 watts, but cooling coils were provided in case of

higher-power operation. They were used in the later 2-kw operation. In the

loading sequence of the special fuel system, the total amount of distilled

10
water was added to the core tank, then the first aliquot of fuel (one-half
of critical mass). Next, gas was bubbled through the solution to mix the
. fuel and distilled water, and a part of the resulting solution was raised
into a mixing tank. The next measured aliquot of fuel was added and the
process repeated. The fission gases were disposed of by a special gas-
disposal systefi with accumulator, gages, and valving. A graphite
reflector two feet thick, control rods, and a concrete shield were used.
Other Atomics International reactors similar to AE-6 include the
Livermore Water Boiler Neutron Source,3“»35 starting in 1953; the
University of California at Los Angeles (UCLA) Medical Reactor and the
Armour Research Reactor, which went critical in 1956; the Danish Reactor-l
and the Japan Research Reactor of 1958, the Italian CESNEF Reactor of 1959; the
Walter Reed Army Medical Center Homogeneous Reactor, Washington D.C., which went
critical in 1961; and the Los Alamos Medical Reactor, which was scheduled
for construction starting in 1958 but apparently never went critical,1©s,36,37
The Armour Research Reactorl®232 was the first operating industrial
reactor in the U.S. It was operated by Illinois Institute of Technology in
Chicago and financed by 24 industrial users. It went critical on June 24,
1956. The reactor operated until 1958 at 10 kw power, an interim power
limit "pending the results of kinetic experiments ... carried out on an
identical machine by Atomics International for the AEC.3® The design for
the NAA Medical Reactor3? was similar to that of the Armour Research
Reactor, as were those of the AL Models L-54 and L-55.1°
Other reactors resulted from another Atomics International water
boiler type, the "Laboratory Reactor," beginning with the L-47 prototype,!®
a predecessor to the L-77. Reactors of this type are the University of
Wyoming Research Reactor and the Puerto Rico Reactor at the Puerto Rico
Nuclear Center, both of which went critical in 1959.
These L-77-type reactors are stainless-steel spherical vessels
containing a critical mass of 20 percent enriched uranyl sulfate fuel and
are controlled by cadmium-lead rods working through vertical thimble
tubes; lead is used as a reflector.16532,38,40 Flora comments that the
"Los Alamos 'Water Boiler' was chosen as a starting point since this
type of research reactor has more years of successful operating history
il

to its credit than any other type of research reactors.,

. The unit was so designed that the factory '"package" could be shipped

11
and assembled without difficulty and could be used in existing facilities
and without special staff. One L~-77, known as the DEMAG Reactor, was
operated at the Second Geneva Conference in 1958 as part of the commercial
exhibit., Another 1L-77 was included in the U. S. Atomic Energy Commission's
exhibit at Beirut, Lebanon, in its Atoms for Peace Program.l+2
Various modifications of this reactor were made, such as the 5-Watt

Laboratory Reactor“l»%3 or the Small 5-Watt Laboratory Reactor.*!

Other Water Boilers

Water-boiler research reactors, in addition to those of Atomics
International, include those designed at universities and miscellaneous
types. North Carolina State College at Raleigh was the first university
to design (R. L. Murray), build, and operate a reactor.“*" A water boiler
was chosen because details of this type were the first to be released by
the AEC, The North Carolina State College Reactor (NCSR or NCSCR), also
known as the Raleigh Research Reactor (RRR-1), went critical in 1953.32,%43
NCSCR-1,32 NCSCR-2,32 and NCSCR-4“46 were all the water boiler type; the
NCSCR-3, however, was a 10~kw pool type completed in March 1960. NCSCR-4
still is in use at North Carolina State, and a similar type is in use
at the University of Wyoming."7»>*8 The NCSCR reactors are substantially
the same as the original LOPO-HYPO-SUPO, except that a cylindrical core
is used instead of a sphere. The Utah Water Boiler reactor3? is a reactor
designed at the University of Utah and described by Borst and Mong, #9550
It is intended for a university wishing to operate with the utmost
simplicity and safety. More information on university reactors is given
in the records of the University Reactor Conferences.*’

Two French homogeneous aqueous reactors for special-purpose testing
have recently been reported.®! Few details are given. Alecto I and
Alecto II, at Saclay, France, operate at a few watts, with a maximum
thermal flux of 108 neutrons/cmzlsec. Alecto I, which became critical
in 1961, uses uranium nitrate solution in light water. Alecto II,
which became critical in 1962-63, uses a solution of plutonium nitrate,
also in light water. Both were designed, built, owned, and operated
by CEA (Centre d'Etudes Nucléaires de Saclay).

Miscellaneous water boilers on which there is little information are

the Chinese Homogeneous Research Reactor at Tientsin,>?:°3 and that of

the Gamma Corp., Mansfield, Massachusetts.5%

12
A modified form of the water boiler is the Test Tube Reactor developed
at Los Alamos.’® This reactor, so-called because of its cylindrical shape,
is related to the LAPRE-1 (Los Alamos Power Reactor Experiment Number 1)
which has also been called a '"test—-tube reactor.'" The fuel solution,
highly enriched uranyl sulfate or phosphate in light water, moves by
convection up a hot central tube and returns through the surrounding
annulus, which contains coolant coils. This reactor has a power rating
of 100 kw.

Babcock and Wilcox also designed a similar modified water boiler for
nuclear and engineering research.’® The design power was 50 kw, with
enough heat-transfer surface in the core to make higher power possible.

The solution-~20 liters of uranyl sulfate in light water--is in a

cylindrical container with a hemispherical bottom.

Irradiation and Testing Reactors

In addition to water boilers, other aqueous solution reactors have
been designed for diverse irradiation and testing purposes. Some are
operated at higher power than water boilers, and two are not strictly
homogeneous, in that they contain solid moderators.

Several aqueous—solution reactors were conceived as neutron
producers. An exploratory study of reactors for gamma irradiation in
food sterilization®’ indicated that homogeneous reactors have no advantage
over other types as irradiation sources for food sterilization if they
are used only for this purpose. If, however, they can be used both for
irradiation and power production, they have several economic advantages,
A meat-irradiation facility has been tentatively suggested in another

>8 This facility would be a 15-Mw reactor with a solution of

report.
uranyl sulfate and copper sulfate in light or heavy water. It resembles
the Homogeneous Reactor Test. (See Chapter 3.)

The Homogeneous Research Reactor, also known as the Aqueous Homogeneous
Research Reactor, is a single-region reactor with 8 percent enriched uranyl
sulfate solution in heavy water as fuel-moderator-coolant. It was
Intended to operate at 500 Mw(t) or 125 Mw(e) and to produce a thermal

53,60 Tt was concluded that the design

flux of 5 x 10%° neutrons/cmz/sec.
could proceed on the basis of the assumed parameters. Successful operation
was eXpected, with no technological breakthroughs required but with many

attendant problems.61

13

A later study at Aeronutronic Systems, Inc., considered both
heterogeneous and homogeneous reactors for a flux of 2 x IO]'5 neutrons/cmz/
sec. A homogeneous Advanced Engineering Test Reactor (AETR) based very
largely on the design of the HRE, HRT, and the Homogeneous Research Reactor
was selected as preferable.®2 Removal of fission products and addition of
fuel would be continuous. To maintain the total contained excess reactivity
at a constant level, the solution, with 10 percent uranium-235, would be
contained in a stainless-steel sphere 8 feet in diameter.

A reactor suggested for high-flux irradiation in chemical processing
employs a subcritical zone for exterior circulation of fuel.®3 A
homogeneous aqueous solution of uranium phosphate is one form of fuel
specified. The fuel flows from the reactor to a cooling zone, and then
back to the reactor. The passages into and out of the reactor contain
neutron-absorbing material, like boron, to keep the material subcritical
outside of the reaction zone. This arrangement permits very high fluxes—-
above 1013 neutrons/cmZ/sec--and high heat generation--more than 20 Mw
per pound of fissile material. Several variants are described, with such other
fuels as gases, liquid metals, molten salts, or fluidized solids.

A high~-flux homogeneous reactor described by Hibshman utilizes
enriched uranyl sulfate in light water to give a power of 28 Mw.%" The
fuel is fed to the core and withdrawn through many inlet and outlet tubes.

Two reactors with aqueous solutions and solid moderators have been
suggested.

The ASTRA (Advanced Scientific Techniques Research Associates)

Advanced Engineering Test Reactor also used ideas from the Oak Ridge
Homogeneous Reactor Project. It, however, has a structural core. This
reactor contains highly enriched uranyl sulfate in heavy water solution,

and a moderator-reflector of graphite blocks.®>

The High Flux Research Reactor for Large-Volume Irradiations is an
engineering test reactor with a fuel-coolant sclution of uranyl sulfate
in light water. The solution is circulated through the graphite moderator

in parallel tubes. The reactor is designed to produce 120 M. 66

Evaluation and Status

As the water boiler is the most common research reactor, its future

possibilities have received the most attention. Breazeale®’ has compared

14

water-boiler and the swimming-pool reactors—-the two designs suitable for a
low-power research reactor. The water boiler, using a light-water

solution of uranyl sulfate and surrounded on all sides by a good reflector
like carbon or beryllium, will have a critical mass of less than 1 kg of
uranium-235. The swimming-pool type with 1l-kg critical mass would be
inflexible and, because of its small heat-transfer surface, could be operated
only at low power. But at 2-kg critical mass, the swimming-pool type has
greater power capability and flexibility than the water boiler of 1 kg.
Presumably, 2 kg enriched uranium is available for this extra flexibility;
moreover, it is easier to use a water shield with the swimming-pool reactor
than with the water boiler, which needs a shielded tube to carry off the
radiocactive gases.

Lane®® has pointed out that water boilers have limited use as research
reactors, although a few years ago they seemed to be promising competitors
to the swimming-pool and tank-type reactors. Some of the difficulties are:

a. The use of water as a fuel-bearing medium limits operating

temperature and neutron flux.

b. Severe corrosion, both from circulating solutions and radiation,

can damage equipment, resulting in such conditions as loss of
radioactive solution from a reactor tank. Such leakage
occurred in the Raleigh Research Reactor.

c. Unsafe conditions may result from excessive reactivity changes.

d. Hazards arise from the explosive hydrogen-oxygen mixture produced

by radiolytic decomposition of water.

e. Highly trained technical workers, especially chemists, are required.
Water boilers and similar solution reactors have many of the advantages and
disadvantages discussed by Lane in a recent work on aqueous reactors.®®

Solution reactors have generally fallen out of favor for research uses,
with few new ones being built. However, the water-boiler reactor at Walter
Reed Hospital went critical as late as 1961 and was in operation by the fall
of 1962.70s71 Also, in early 1963, the AEC announced that it planned to
issue a permit to the University of Nevada for construction of an Atomics

International Model L-=77 reactor.72

15
DATA SHEETS

RESEARCH AND TESTING REACTORS
No. 1 Zero-Power Reactor

. (ZPR-I and ZPR-II)
NASA Lewis Research Center

References: NACA-RM-SE-57-F-28; TID-8200 (9th rev.); AEC Press Release,
Mar, 23, 1962,

Originators: Laboratory staff,.

Status: December 1963; critical experiments being carried out.
Details: Two reactors, ZPR-II stated to be enlarged version of ZPR-I, but
power is given as the same; otherwise also apparently the same, Thermal
neutrons, steady state, burner. Fuel-moderator-coolant: solution of
highly enriched uranyl fluoride in HZO' Concentration will change according
to experiment. Reflector will also vary. Solution added as needed through
line from solution room. Core vessel: c¢ylinder, open at the top. Shielding:
54-in, thick concrete. Control: amount of solution in reactor and inherent
negative temperature coefficient of the solution. Control rod used for
safety only. Emergency dumping into safe-~geometry storage tanks provided
as additional safety measure. Power: normally 1/10 watt, with maximum
of 10 watts; neutron flux up to 108/cm2/sec.
Code: 0313 13 31201 44 624 74 83779 921 101

84677

811XX

No., 2 Highly Enriched Homogeneous Reactor

WAPD

Reference: WAPD-128.

Originators: J. R. Brown, B. H. Noordhoff, and W. O. Bateson.

Status: Critical experiments performed, May 1955.
Details: Thermal neutrons, subcritical reactor. Fuel-moderator-coolant:

235 235

solution of uranyl nitrate in H,0; more than 907 U ; ratio of H to U y

approximately 1700, Reflector:2 Lucite plastic at bottom of tank, which is
a stainless-steel cylinder 36 in. diameter, 72 in. high, with 1/8 in. thick
walls, Reactor tank inside a concentric cylinder, 52 in. diameter, 70 in.
high, filled with water, Control: two safety and one regulating rod. All
move vertically. Each is hollow Cd cylinder between concentric tubes of
stainless steel. Emergency dumping provided as additional safety measure.

. Code: 033X 13 31201 44 624 711 81112 921 101

84677

83779

19
No. 3 Proserpine Critical Experiment

Commissariat & 1'Energie Atomique, France

Reference: Proc. 2nd U.N. Int. Conf., 12, pp. 539-562.

Originator: J. Bertrand et al,

Status: Critical in 1958.

Details: Thermal neutrons, critical experiment. Fuel-moderator-coolant:
concentrated solution of plutonium sulfate in HZO' Reactor is reflected, single-
region. Reflectors: 27.5-cm thick layer of BeO next to the core tank and 50-cm
thick layer of graphite next to the BeO. Core vessel: cylindrical stainless-
steel tank, 25 cm in diameter and 30 cm high, with a wall thickness of 1 mm.
Outside the core tank is one of aluminum designed to retain solution leaks.
Control: parallel horizontal regulating rods. One, for fine regulations, is a
stainless-steel cylinder. The other, for compensation, 1s a stainless-steel
cylinder lined with Cd. Rods are driven by a mechanism outside the pile. Two
horizontal safety rods, externally tangent to the tank, are of an alloy of Al
and Mg. They are lined with Cd and filled with boron carbide beads.

Lode: 0313 13 31201 46 624 711 81212 921 101

No. 4 LOPO, Low Power Water Boiler

LASL

References: AECD-3063 (LADC-819); AECD-3059, (Rev. Sci., Instruments, 22, No. 7,
pp. 489-499, July 1951); Proc. 1st U.N., Int. Conf., 2, pp. 372-391, Lane et al,
Fluid Fuel Reactors, pp. 341-346.

Originators: D. W. Kerst. Calculations of critical mass first performed at

Los Alamos by R. F. Christy.

Status: Startup, May 1944, Replaced by HYPO November 1944,

Details: First "water boiler"--so-called because of gas bubbles formed by radio-

lytic decomposition of water--and first reactor to use enriched fuel. Thermal

neutrons, steady state, burner, some conversion. Fuel-moderator-cooclant:

235) in H20. Solution contained in a thin-

walled stainless-steel sphere enclosed in thermostatted housing at 39 + 0.01°C.

solution of uranyl sulfate (14.5% U

Control: Cd-wrapped brass cylinder, with another brass tube fitting snugly over
the Cd and moving vertically in the reflector. Inherent safety feature: negative
temperature coefficient. Emergency dumping of solution provided as additional
safety measure. Over-all reactor size, 3 ft square by 4 ft high. Power:
50 milliwatts.
Code: 0311 13 31201 43 624 744 81152 921 101

84677

20

No. 5 HYPO, High Power Water Boiler

LASL

References: AECD-3065 (LA-394); AECD-3059 (Rev. Sci. Instruments, 22, No. 7,
July 1961, pp. 489-499); Lane et al, Fluid Fuel Reactors, pp. 341-6.

Originators: L. D. P. King, advised by E. Fermi on basis of LOPO.

Status: Critical, December 1944; Alterations began April 1949 to replace HYPO
with SUPO. Dismantled, 1950.
Details: Same as LOPO except: (1) 1l4%Z enriched U235 as uranyl nitrate instead
of sulfate; (2) natural convection plus internal cooling coil with circulating
water; (3) thicker sphere with horizontal l-in. pipe or ''glory hole" for
access to highest possible neutron flux; (4) two additional Cd control rods;
(5) addition of graphite thermal column; (6) inclusion of y-ray and neutron
shields; and (7) elimination of hydrostatic control system and storage
reservoir for solution. Solution temperature: 85°C, Peak power of 6 kw.
Code: 0311 13 31201 43 624 744 81152 921 105

84677

No. 6 SUPO, Super Power Water Boiler

LASL

References: LA-1301; AECD-3287 (LADC-1081); Lane et al, Fluid Fuel Reactors,
PP. 341-6.

Originators: L. D. P. King, group director.

Status: In operation, June 30, 1961, Modifications (of HYPO) first phase

completed February 1950. Modifications second phase completed March 1951.
Details: Modified HYPO First Phase: Fuel enrichment increased to 88.7% U235
(as uranyl nitrate). Be reflector replaced by graphite. Interior of sphere
rebuilt to include 3 stainless-steel coolant coils. Coolant temperature: 130°F;
atmospheric pressure. Second thermal colum added. Two more control rods of
sintered BlO in thimbles added, 3 Cd sheets tangent to sphere included. Reactor
size (outside of shield): 15 ft x 15 ft x 11 ft high. Power increased to 30 kw.
Reactor was run for 10,000 kw-hr before second phase modifications were begun.
Power then increased to 45 kw. Other alterations include rebuilding of original
thermal column, construction of recombination system to handle offgases, and
construction of shielded solution-handling system. Estimates also given for
intermediate and fast neutroné.
Code: 0313 13 31201 44 624 711 81111 921 105

81112

21
No. 7 Supo Model II

LASL

References: U.S. Patent 2,961,391; Proc. lst U.N. Int. Conf., 2, 1955,
pp. 372-91.
Originator: L. D. P. King.

Status: Patent granted, November 1960.
Details: Same as SUPO with following changes: core shape changed to
66-in.,~high, 12-in.-diameter cylinder; single control rod (presumably Cd);
gas-circulation rate in recombination system increased to 200 ml/min; sizes
of catalyst chamber and heat exchanger increased to produce higher power
level of 400 kw. Lower part of the cylinder is critical region, filled
with solution to a height equal to its diameter (12 in.). TFuel is
preferably uranyl nitrate, but sulfate may be used.
Code: 0313 13 31201 44 624 711 81112 921 105

84677

No. 8 WBNS, Water Boiler Neutron Source

Atomics International, A Division of North American Aviation, Inc.

References: Science, 119, pp. 9-15, Jan. 1, 1954; NAA-SR-839; AECU-2900,
pp. 183-191.
Originators: R. Chalker and M. E. Remley.

Status: Put into operation April 1952. Dismantled July 1956.

Details: Based on data from LOPO and HYPO. Thermal neutrons, steady state,
burner, Fuel-moderator-coolant: solution of 90% enriched U235 as uranyl
nitrate in distilled HZO' Cooling coils provided (but not used because of
low power) in l-ft-diameter stainless-steel spherical core tank. Reflector:
cylindrical, of graphite blocks adjacent to sphere; also serves as thermal
column. Design pressure: 300 psi. Control: 2 safety rods—-strips of
boral attached to Al channel to form I-beams; gravity-actuated with
horizontal movement, and 1 fine (pipe) and 1 coarse (also I-beam) Cd

control rods. Reactor housed in tank 6 ft high, 5 ft in diameter. Operating
power: 1 watt.

Code: 0313 13 31201 44 624 711 81211 921 101

81212

22
No. 9 AE-6 Water Boiler Neutron Source (Model L-6)

Atomics International, A Division of North American Aviation, Inc.

References: TID-2503, pp. 69-76; IAEA Directory of Nucl. Reactors, 2,
1959, pp. 181-6,
Originators: G. L. Blackshaw and C. H. Skeen.

Status: Critical, November 1956; in operation, June 30, 1961.

Details: Basically a conversion of WBNS to power of 2.5 kw. Changes
include 93.11% enriched U235
system employed with distilled water (inlet temperature, 42°F; outlet,

as uranyl sulfate instead of nitrate; cooling

52°F). Two safety rods--strips of boral attached to an Al I-beam, 1 shim
rod of Cd attached to Al I-beam, 1 regulating rod of Cd-filled stainless-
steel tube. Horizontal movement: tangential to core. Gas-handling
system and heavier permanent shielding around core and reflector added.
(AE-6 [or L-6] was the prototype for L-55 operating as Danish Research
Reactor No. 1.)
Code: 0313 13 31201 44 624 711 81211 921 105

81212

84677

No. 10 Livermore Water Boiler Neutron Source, LIWB or North American

Aviation Research Reactor Model L-3

(also Cal. Research & Development CR&D Reactor;

Livermore Research Reactor LRL)
Livermore Research Laboratory

References: NAA-SR-MEMO-784; NAA~AER-1023 (extended version¥*),
Originator: NAA staff.

Status: Startup, November 5, 1953. 1In operation, June 30, 1961.
Details: Higher-power version of WBNS. Same except for fuel, uranyl
sulfate instead of nitrate. Coolant: distilled water. Sphere: 12-1/2-in.
diameter. Two safety rods and 1 coarse control rod of flattened stalnless-
steel tubing packed with boron carbide. One fine control rod of boron
carbide. Power: 500 watts. Unlike other models of WBNS, LIWB has
completely closed-cycle sweep gas and recombination system.

*Extended and revised version produces more power (50 kwt), has larger

thermal column, and uses bismuth instead of lead for the gamma shield.
Code: 0313 13 31201 44 624 711 81211 921 105
81221

23
No., 11 Armour Research Reactor, (ARR)

(Atomics International Model L-8)
Armour Research Foundation*

References: Proc. 2nd U.N. Int. Conf., 10, pp. 393-403; Proc.

2nd U.N. Int. Conf., 10, pp. 404-418; IAEA Directory of Nuclear Reactors,
2, pp. 165-70.

Originators: Staff of Atomics International, a Division of North American

Aviation, Inc.

Status: Critical, June 1956; 10 kw operation, December 1956; 100 kw
operation, January 1959. In operation, June 30, 1961.

Details: Thermal neutrons, steady state, burner. Fuel: 88.147 enriched
U235 as uranyl sulfate in light water. Coolant: distilled light water
circulating through cooling coils inside core. Working pressure below
atmospheric. Reflector: graphite blocks. Stainless-steel spherical core
vessel 12.4 in. diameter. Four control rods of boron carbide clad with
stainless steel, 3 electrically driven, (rack & pinion type) 1 servo-system
operated, moving vertically. Because of urban location (Chicago), system
of barriers (shields) arranged: primary, closed core vessel and gas-
handling system of stainless steel; secondary, core vessel and steel

lining of sub-pile room, gas-tight doors, in Al cylinder; tertiary,

reactor room, air-lock, gas-tight doors, sealed condults and water-
floodable ducts for isolation of room ventilation system. Can run at 50 to
100 kw. ARR is the prototype for Atomics International Model L-54. (First
industrial research reactor). Vertical and horizontal experimental
facilities.

Code: 0313 13 31201 44 624 711 81111 921 105

*Now Illinois Institute of Technology Research Foundation.

24

No. 12 Medical Reactor

NAA

Reference: NAA-AER-1180.
Originator: NAA staff,

Status: Conceptual design, 1955; design used for the LASL Medical Reactor;
similar to L-8 (ARR) design.

Details: Thermal, steady state, burner. Fuel-moderator solution is
enriched UOZSO4 2

cooling coils in the core. Spherical core, 1-ft diameter, is surrounded

in light water, cooled by chilled H,O circulating in

by a graphite reflector (5 ft x 5 ft x 8 ft long). The core has 5 re-
entrant thimbles, 4 for control and safety rods of BAC pressed into a
vertical cylinder and 1 as an exposure facility. A secondary Al enclosure
around the stainless steel pressure vessel insures against any leaks.
Maximum fuel-solution temperature at full power (50 kw) is 80°C. The
design includes extensive exposure facilities for medical research.

Code: 0313 13 31201 44 624 711 81111 921 105

No. 13 Research Reactor Model L-54

Atomics International, A Division of North American Aviation, Inc.

Reference: Proc. 2nd U.N. Int. Conf., 10, pp. 404-18,
Originator: AI staff.

Status: Concept used for five operating reactors: JRR-1, Japan; FRB
(BER), Berlin, Germany; FRF, Frankfurt, Germany; CESNEF, Milan, Italy;
and Walter Reed Medical Research Reactor.

Details: Revision of ARR (L~8) design, using 207 enriched uranium in
uranyl sulfate. Primarily burner, but some internal conversion. Core
size increased to 15.75 in. diameter to contain larger amount of fuel
solution. Wet~type gas-handling system. Other aspects indentical with
ARR (L-8). Operating power: 50 kw.

Code: 0311 13 31201 43 624 744 81111 924 105

25

No. 14 Research Reactor Model L-55

Atomics International, A Division of North American Aviation, Inc.

Reference: Proc. 2nd U.N. Int. Conf,, 10, pp. 404-18.

Originators: J. W. Flora et al.

Status: Concept used for one reactor, the Danish DR-l.
Details: Variation of AE-6 (L-6). Thermal neutrons, steady state, burner.

Fuel can be uranyl sulfate (207 to 90% U235

) in light water. Coolant:
distilled water. Reflector: graphite. Core vessel: stainless-steel
sphere, 12.5 in. diameter. Four horizontal control rods (2 safety, 1 shim,

1 regulating) of flattened stainless-steel tubes filled with B,C powder,

4
tangent to core vessel. Vapor-pressure-type recombiner. Thermal column,
5-1/2 ft square vertical. Power: up to 1.5 kw.

Code: 0313 13 31201 43 624 744 81221 921 105

b4 711

No. 15 Laboratory Reactor Prototype L-47

Atomics International, A Division of North American Aviation, Inc.

Reference: Proc. 2nd U.N., Int. Conf., 10, 1958, pp. 404-18.

Originator: Al staff,.

Status: Startup, August 29, 1957; dismantled, May 1958.

Details: Five-watt prototype unit for L-77. Thermal neutrons, steady
state, burner. Fuel-moderator: 897 enriched U235 as uranyl sulfate in
light water. Cooled by natural convection; no cooling coil. Stainless-
steel spherical core vessel 12.5 in. diameter, 12.5 liter capacity.
Canopy assembly over core contains recombiner. Reflector: 1lead shield,
6 in. thick, completely surrounding reactor core and recombiner system.
Core reflector assembly supported in middle of shield tank, 8 ft diameter,
8 ft high, filled with water. Sodium octaborate later added to water to
suppress water—capture gamma radiation. Control: two vertical Cd
control rods. Experimental facilities: 5 through-tubes in the core and
2 beam tubes terminating at the lead surface. Operating power: 1 watt.
Code: 0313 13 31201 44 624 711 81;12 921 101

26

No. 16 1-77 Laboratory Reactor

Atomics International, A Division of North American Aviation

References: Proc. 2nd U.N. Int. Conf., 10, 1958, pp. 404-18;
IAEA Directory of Nuclear Reactors, 2, 1959, pp. 205-210.

Originator: Atomics International, Division of NAA.

Status: Startup May 1958; in operation, June 30, 1961.

Detajls: Same as L-47 except: fuel: 20-90% enriched U235. No cooling
coil. Reflector: Pb shot-diphenyl mixture. Core vessel: sphere,

15.75 in. diameter; same canopy assembly for recombiner. Working pressure:
25-28 in. Hg. Containment vessel: stainless steel, 8 ft diameter, 7 ft
high. Three-region shield system: inner, pelletized Pb-diphenyl

compound; intermediate, paraffin and soluble organic boron compound;

outer, borated paraffin and pelletized Pb, enclosing intermediate and
inner shields on sides and top. Safety and regulating rods; Cd-sheathed
stainless steel sealed in aluminum. Power: 10 watts.

Code: 0313 13 31201 44 624 711 81112 921 101
43 744

No. 17 5-Watt Laboratory Reactor

(Atomics International Aqueous Homogeneous Laboratory Reactor)
Atomics International, A Division of North American Aviation

References: Nuclear Eng., 1, No. 8, November 1956, pp. 344-5; 2nd Nucl,

Eng. and Sci. Conf., 2, "Advances in Nucl. Eng."

Originator: AI staff.

Status: Manufactured by AL for educational and small research laboratories.
Details: Smaller and considerably simplified version of ARR, with some
details similar to 1L-47 and L-77. Thermal neutrons, steady state, burner.
Fuel-moderator: highly enriched U235 as uranyl sulfate in H20. Coolant:
water, natural conduction and convection; no additional cooling. Core

vessel: stainless steel, 12 in. diameter. Hemispherical canopy over

" sphere houses recombination and overflow chamber. Core and canopy contained

in 6 in. thick Pb shield which acts as reflector and primary shield. These
are enclosed in Al container, 2 ft diameter and 2 ft 8 in. high. Complete
assembly slung from 3 steel tubes in a tank, 8 ft diameter, 8 ft high, of
borated H20. Control: two Cd cylinders as control and safety rods moving
vertically. Power: 5 watts.

Code: 0313 13 31201 44 624 711 81112 921 101

27
No. 18 North Carolina State College Reactor-l

(NCSCR-1, NCSR-1, RRR-Raleigh Research Reactor)

References: AECU-1986; IAEA Dir. of Nuc. Reactors, 2, 1959, pp. 189-98.
Originator: C. K. Beck.

Status: Critical September 1953; dismantled, June 1955. Modified to
NCSCR-2.

Details: Based on HYPO, SUPO, and HRE series. Thermal neutrons, steady

state, burner. Fuel: uranyl sulfate with 93% enriched U235 in aqueous

solution. Coolant: H20 circulating through 4 stainless-steel, Internal
cooling coils of hairpin shape. Reflector: 5 ft cube of graphite bars
surrounded by Pb. Core vessel: stainless-steel cylinder, 10.75 in.
diameter, 9.2 in. high. Control: 2 combined control and safety rods of
thin-walled stainless-—steel tubes filled with boron powder; 2 shim rods

of Cd-filled stainless-steel tubes ("strips") tangent to cylinder. Control
rods enter core vertically. Reactor containment vessel: octagon, 17 ft
across. Core below ground level. Power: 10 kw (+). NCSCR-1 was the
first university-owned and operated reactor in the U.S.

Code: 0313 13 31201 624 711 81111 921 105
84677

No. 19 North Carolina State College Reactor-2

(NCSCR-2, NCSR-2, RRR-Raleigh Research Reactor)

Reference: IAEA Dir. of Nuc. Reactors, 2, 1959, pp. 189-98.

Originator: C. K. Beck.

Status: Critical May 1957; dismantled, December 1958. Modified to NCSCR-4.
Details: Modification of NCSCR-1, Differences: no coolant; cooling by

natural convection. Core vessel: cylindrical with hemispherical bottom
and flat circular top. Secondary cylindrical enclosure of Al. Control:
2 combined regulating and safety rods of stainless steel tubes containing

tamped B,C, with vertical movement. Power: 500 watts (+).

4
Code: 0313 13 31201 44 624 711 81111 921 101

28

No. 20 North Carolina State College Reactor-4

(NCSCR-4, NCSR-4, RRR-~Raleigh Research Reactor)

References: NP-7041; IAEA Dir. of Nuc. Reactors, 2, 1959, pp. 189-98.

Originator: NCSC Reactor staff.

Status: In operation, June 30, 1961.

Details: NCSCR-2 moved to new location and modified. Main parts of

NCSCR~-2 used; new instrumentation, shielding, and gas recombiner installed,

Power decreased to 100 watts (+).
Code: 0313 13 31201 44 624 711 81111 921 101

No. 21 Water Boiler

University of Utah

References: AECU-2900; Nucleonics 12, No. 4, p. 11, April 1954.

Originator: ?

Status: ? (Financing incomplete, 1954; designed.)
Details: Water-boiler-type reactor. Power: greater than 100 kw.
flux: 3 or 4 x 1012 neutrons/cmz/sec.

Code: 031X 13 31201 4X 624 7XX  8XXXX  9XX 101

No. 22 Alecto I Water Boiler

Centre d'ftudes Nucléaires de Saclay, Gif-sur-Yvette, France

Reference: D. Breton, letter to the editor, Nucleonics, 19, No. 1,

p. 6, 1963,

Originator: Centre d'Etudes Nucléaires de Saclay staff.

Status: Became critical, 1961. Presumed still operating.

Neutron

Details: Special-purpose testing reactor. Thermal neutrons, steady

235

state, burner. Solution of uranyl nitrate (907 U ) in H,O0. Various

2

reflectors used. Maximum thermal flux, 108 neutrons/cmz/sec. Power:

few watts.
Code: 0313 13 31201 44 624 711 BXXXX 9XX 101

29
No. 23 Alecto I1I Water Boller

Centre d'fitudes Nucléaires de Saclay, Gif-sur-Yvette, France

Reference: D. Breton, letter to the editor, Nucleonics, 19, No. 1, p. 6,
1963.
Originator: Centre d'fitudes Nucléaires de Saclay staff.

Status: Was to become critical, 1962-63.
Details: Same as Alecto I, except that plutonium nitrate solution is fuel.
Code: 0313 13 31201 46 624 711 8XXXX 9XX 101

No. 24 Tientsin Homogeneous Research Reactor

Nan-K'al Polytechnic Institute (China)

References: AD-248402; CF-59~2-76.

Originator: Unknown; built and operated by teachers and students.

Status: Startup, 1959; in operation, 1960.

Details: Reactor is designed for research, instruction and production of
2

isotopes. Maximum neutron flux outside reactor: 4.5 x 106 neutrons/cm” /

sec. Power level: 3 watts.
Code: 0313 13 31201 4X 62X 7XX 8XXXX 9XX 101

30

No. 25 Test Tube Reactor

LASL

References: Nucleonics, 13, No. 11, pp. 72-5, November 1955; Proc. lst

U.N. Int. Conf., 2, 1955, pp. 372-91.

Originators: R. P. Hammond and H. M. Busey.

Status: Conceptual stage, 1955; no further work.

Details: "Radical modification" of water boiler type. Thermal neutrons,
steady state, burner. Fuel: highly enriched U235 as uranyl sulfate or
phosphate in water. Fuel moves by convection up hot central tube and
returns through surrounding annulus containing layers of cooling coils.
Coolant: H20° Reflector: Hfh Test—-tube-shaped structure (cylindrical)},
lower portion containing critical core and heat exchanger and upper
containing recombiner. Entire reactor contained in jacketed stainless-
steel tube; all external pipes connected through cover. Assembly
suspended vertically in a convection chimney. Controls: (1) negative
temperature coefficient; (2) vertically moving curtain of boron steel;
(3) sandwich of stainless steel or Dural containing samarium-gadolinium
oxides; or (4) other neutron-absorbing materials, Power: 100 kw,

Code: 0313 13 31201 44 624 711 81144 921 105

84677

31
No. 26 Water Boller

Babcock and Wilcox Company

Reference: TID-10117.
Originator: Babcock and Wilcox staff.

Status: Design stage, November 1955.
Details: Modified water boiler. Thermal neutrons, steady state, burner.
Fuel: enriched uranyl sulfate in HZO' Coolant: water circulating through
tubes in solution; water is cooled in external heat exchanger. Reflector:
2 ft of graphite. Shielding: & in. steel thermal shield and 5 ft high-
density concrete. Reactor housed in Zr or stainless-steel cylinder with
hemispherical bottom. Negative pressure maintained to prevent escape of
radicactive gases. Eleven Al-lined beam ports in faces of the shield for
experiments. Control: negative temperature coefficient and use of two
control rods. Power: 50 kw, with enough heat-transfer surface in core
for higher power.
Code: 0313 13 31201 44 624 711 8l1XX 921 105

84677

32
No. 27 Homogeneous Reactor Gamma Irradiation Facility

ORNL

Reference: CF-56-7-126.
Originators: R. B. Briggs and J, 0. Kolb,

Status: Preliminary design, made at request of Army Reactor Branch,
July 27, 1956.

Details: Thermal neutrens, steady state, converter. Resembles HRT.
Fuel-modera;g;-coolant: solution of u;ggyl sulfate in either HZO or

DZO; 17 g U7 /1liter in HZO or 3.7 g U " /liter in DZO would be critical

concentrations. CuSO4 and excess HZSO4 added to either solution. High-
pressure (1250 psia) circulating system in which reaction is sustained
and low-pressure system in which fuels are stored, radiolytic gases are
recombined, and fission-gases are separated for delivery to source.
Reactor vessel: stainless-steel vessel, pear-shaped, 4 ft in diameter,
with volume of 35 cu ft. Solution enters reactor at 440°F, 1500 gpm.

It leaves at 525°F. Solution leaving reactor passes through gas separator,
is cooled in heat exchangers, and pumped back to reactor. Steam produced
at 420°F and 320 psia can be used to generate 3 Mw(e). Control: negative
temperature coefficient and by alterating fuel concentration. Power:

15 Mw. Designed for irradiation of meat.

Code: (311 13 31201 44 624 711 84677 921 101

14 31202 625 83779

33
No. 28 Aqueous Homogeneous Research Reactor (HRR)

ORNL

References: ORNL-2256; TID-7540, pp. 288-301.
Qriginators: P. R. Kasten, M. I. Lundin, and C. L, Segaser.

Status: Feasibility study, 1957; project cancelled, 1961.
Details: Thermal neutrons, steady state, converter, Fuel: an 8%

enriched uranyl sulfate (+ CuSOA) solution in D,0, the coolant-moderator.

Solution enters reactor vessel through a diffusion screen at the bottom
at 225°C, rises 50°C in temperature as it moves upward, and leaves at

the top. Heat is transferred in 6 double~drum heat exchangers. Solution
pressure: 1400 psia. Reactor vessel: 10 ft ID sphere of stainless-
steel-lined carbon steel. No reflector. Control: negative temperature
coefficient. Research facilities include six horizontal and one vertical
experimental holes. A heterogeneous converter-plate design, constructed
of seven thin concentric cylinders, intended to obtain high fast-neutron
fluxes, may be used to enhance the over-—all utility of the reactor.

Power: 500 Mw(t), 125 Mw(e).
Code: 0311 14 31202 42 625 743 84677 91X 101

34
No. 29 Homogeneous Advanced Engineering Test Reactor (AETR)

Aeronutronic- Systems, Inc.

Reference: AECU-3478 (U-047), pp. 42-60; 85-103.

Originators: Staff of Aeronutronic Systems, Inc., subsidiary of Ford

Motor Co., Glendale, California,.

Status: Submitted in response to AEC request for recommendations of a
concept for an AETR, March 29, 1957; no further work.

Details: Thermal neutrons, steady state, converter. Fuel-moderator-
coolant is 107 enriched U235 as uranyl sulfate in DZO' It is a bare,
single-region reactor. Fuel solution is introduced into the core vessel
at 208°F (98°C) at 4 points symmetrically distributed about the vertical
axis in the lower hemisphere. Heated solution (308°F or 153°C) is
withdrawn into the primary coolant circuit at four similarly distributed
points in the upper hemisphere. Solution leaves the core through gas
separators and passes through heat exchangers, from which it is returned
to the core by the main circulating pumps. Core vessel: 8§ ft ID
stainless-steel sphere with external thermal shields 5 in. thick, around
which is a cylindrical pressure vessel, 10 ft diameter. HZO flowing
through this vessel acts as a secondary coolant for the thermal shields.
Seven experimental loop systems traverse the sphere vertically through
the central region. Control: the negative temperature coefficient and
by alteration of fuel concentration. Emergency shutdown could be
accomplished by the injection of a boron solution into tubes in the core.
Power: 500 Mw(t).

Code: 0311 14 31202 42 625 743 84677 921 101

35

No. 30 Homqggneoug Nuclear Reactor with Subcritical Exterior Fuel

Circulation Zone

Reference: Canadian Patent 614,567 (British No. 834,371).

Originators: F. T. Barr and J. F. Black.

Status: Patent granted, Feb. 14, 1961,

Details: Thermal (could be intermediate), steady state, burner. Fuel-
coolant-moderator: 907 enriched U235 as uranium phosphate in HZO'
Horizontal core is surrounded by a graphite reflector, which may or may
not be ccoled. The solution is circulated through a flow path, through

a reaction zone, and on to a cooling zone. Inlet temperature: 350°F;
outlet: 650°F. So tfiat the critical condition is not maintained through-
out the conduit, sufficient neutron-absorbing material (Li, Cd, or B) is
disposed circumferentially in the conduit leading into and out of the
reaction zone to establish a subcritical exterior circulation zone. This
neutron absorber may be horizontal rods or plates, clad with Al or stain-
less steel and located within the flowing solution. Reaction zone is
5-15 inches inside the closest absorbent. System pressure: 2000 psi;
flow rate: 400 gal/sec. Control: 3 rods of Cd, B, or Li, moving

vértically. This design is suitable for high-flux irradiation processing.

Code: 0313 13 31201 44 624 711 81111 921 101

0212 81112
81116
Variants: No. 1 == Fuel may be a gas or vapor, such as UF6 at 500-1000 psi.
Code: 0313 1X 31710 44 662 711 81111 921 101
81112
81116
No. 2 = Fuel may be a compound-——UF4 or UIQ—-that is liquid

under the reactor conditioms.
Code: 0311 1X 31XXX 44 613 711 81111 921 101
81112
81116
No. 3 —= Fuel could also be a solution of uranium in liquid Bi,
uranium sulfates or nitrates, or uranium oxy-salts. Uranium or plutonium
oxides could also be used.
Code: 0313 1X 31XXX 44 624 711 81111 921 101
626 81112
634 81116

36

No. 30 Homogeneous Nuclear Reactor with Subcritical Exterior Fuel

Circulation Zone (Cont.)

No. 4 == A fluidized solids system adapted to petroleum
processing is another variation. Solid UO2 particles, 75-200 y, are
conveyed by a gaseous coolant-moderator--hydrogen (or DZO vapor). The
fluid UOZ_HZ is circulated similarly to the first design, but solids
are separated in a cyclone and are later repressurized before they are
returned to the system. Neutron-absorbing material at both ends of the
reaction zone is B or Cd. The system is primarily for conversion, but
can be adapted for breeding. UF6 with solid particles of carbon suspended
in it as moderator can also be used.

Code: 0311 17 31915 4X 683 76X  81XX1 921 103
14 31902 685 81XX2

12 31710 662 109

37
No. 31 Homogeneous Liquid Fuel Reactor

Aqueous Homogeneous Design

Reference: Canadian Patent 613,636.

Originator: H. J. Hibshman.

Status: Patent issued, Jan. 31, 1961.

Details: Several embodiments of the invention are described in the patent,
which is aimed chiefly at methods by which to obtain higher neutron fluxes
from homogeneous reactors by feeding fuel to and withdrawing it from the
core through a plurality of inlet and outlet fuel streams.

Thermal neutrons, steady state, burner. Circulating fuel-moderator-
coolant: solution of 93% enriched U02804 (could be nitrate) in HZO'

Inlet temperature at the bottom of the core: 300°F; outlet at the top:
400°F. Operating pressure: 1000 psig. The core has 4 conduits which
converge to form a quartered sphere, 1 ft in diameter. Shape may be
cylindrical or rectangular. Each of the conduits, constructed of carbon
steel, is circular (6 in. diameter) shaped outside the core and is
altered to form one of the quarters inside the core. An alternate design
describes cooling means for the conduits. Velocity of fuel: 6 ft/sec.
Each conduit is provided with external heat exchanger. Reflector: Be and
if desired, graphite. Control: negative temperature coefficient and fine
or coarse Cd or B steel safety rods. Power: 28 Mw.

Two alternate arrangements are described. In the first, a plurality
of pipes containing the fuel mixture converge in one critical zone and
diverge outside of this zone. The fuel is fed in continuously through the
3 pipes and is passed on to the heat exchangers. The area of criticality
may be defined by a solid moderator and control elements, to which
structure this alternate arrangement is particularly applicable; or a
liquid moderator (HZO’ D2

pipe circuits, similar to those for fuel, through the critical zone,

0, or hydrocarbons) may be circulated within

each circuit with cooling means.

A second arrangement comprises a fixed source of fuel within which a
plurality of pipes pass liquid fuel continuously through the zone, to heat
exchangers, and back to the core. The coolant (HZO’ D20, or hydrocarbons),
which may also be the moderator, passes through the fixed fuel in pipe
circuits. The coolant is also recirculated. Core material is stainless

steel, and D,0 may be used with or in place of reflector materials listed

2

38

No. 31 Homogeneous Liquid Fuel Reactor (Cont.)

for the main concept.
All of the above variants may be further altered by using a suspension
of fuel instead of a solution.

Code: 0313 13 31201 44 624 711 84677 921 101

81X11
81X12
Variants: No. 1 —— The UOZSO4 could be dissolved in D20. Other features

are the same.

Code: 0313 14 31202 44 625 711 84677 921 101

81X11
81X12
No. 2 — A salt of U233 (the nitrate) could be dissolved in
HZO or DZO'
Code: 0313 13 31201 45 624 711 84677 921 101
14 31202 625 81X11
81X12
No. 3 —— Plutonium=239 sulfate could be used as fuel in H20
or DZO'
Code: 0313 13 31201 46 624 711 84677 921 101
14 31202 625 81X11
81X12

39
No. 32 Homogeneous Advanced Engineering Test Reactor (AETR)

Adv. Scientific Techniques Research Assoc. (ASTRA)

Reference: NYO0-4849.
Originators: R. G. Mallon, J. Saldick (ASTRA) and R. E. Gibbons (Catalytic

Construction Co.).

Status: Submitted in response to AEC request for recommendations of a
concept for an AETR, March 1, 1957.

Details: Thermal neutrons, steady state, burner. Fuel: hot, pressurized,
acid solution of uranyl sulfate in D20, highly enriched in U235. Moderator-
reflector: 4 in. graphite blocks (sticks) with slots cut on each of the
four faces into which graphite keys are inserted. The vertical cruciform
openings between sticks are passages for the air cooclant. The blocks are
arranged in an octagonal shape, (7 ft high and 8 ft in diameter) supported
on a steel plate, all of which is surrounded by a cylindrical vessel
containing flowing demineralized water. It approximates a right circular
cylinder with a vertical axis of graphite. Twenty-four Zircaloy-2 vertical
fuel tubes, 5 in, diameter, are inserted into graphite structure from the
bottom. Fuel flows upward in the fuel tubes and downward through a
concentric flow separator in the center through gas separators to steam
generators, and then back to the core, pumped by canned rotor pumps. Inlet
temperature: 428°F; outlet: 479 to 503°F, Pressure: 1200 psia.
Reflectors: top and side reflectors, 2 ft thick. Internal neutron shield
of borated graphite is located just below the core. Control: negative
temperature coefficient. Nine vertical through-holes for testing reactor
fuel elements and assemblies at high temperatures and high thermal neutron
fluxes. Power: 220 Mw(t).

Code: 0313 12 31714 44 625 711 84677 921 109

40

No. 33 High-Flux Research Reactor for Large-Volume Irradiations

Bendix Aviation Corporation

Reference: TID-2507 Del., pp. 329-334.

Originators: R. G. Mallon and W. R. Pearce.

Status: Conceptual design, June 3, 1955,
Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:
fully enriched solution of uranyl sulfate in HZO’ 30 grams per liter,
There is also a graphite moderator, which is fabricated in 4 in. x 4 in.
blocks. Reflector: 20-in.-thick borated graphite. Solution enters
core vessel at 100°F and a velocity of 25 ft/sec and circulates through
the graphite moderator in 24 parallel zirconium tubes. It leaves at an
average temperature of 145°C to external heat exchangers. Because a
conventional tube header would become critical, four manifolds at either
end of the reactor and four separate heat exchangers are used. Core
vessel: right-circular cylinder, 8 ft x 8 ft, enclosed in a concrete
shield. Experimental facilities: fifteen removable graphite
stringers——one of which permits irradiation of samples up to 16 in. x
16 in.--and up to forty-eight 4 in. x 4 in. beam tubes. Control:
through the negative temperature coefficient, alteration of fuel
concentration, and four vertical control rods. Design power: 120 Mw,
average flux in the graphite moderator: 4.4 x 1014°
Code: 0313 110 31201 44 624 71 84677 921 101

83779

8111X

41
10.

11.

12.

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AEC press release, March 23, 1962.

D. Fieno, "Consistent Pl Analysis of Aqueous Uranium-235 Critical
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A. D. Callihan, J. W. Morfitt, and J. T. Thomas, '"'Small Thermal

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C. K. Beck, A. D. Callihan, J. W. Morfitt, and R. L. Murray, "Critical
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J. R. Brown, B. H. Noordhoff, and W. 0. Bateson, "Critical Experiments
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J. Tachon, "Dispositifs de Transvasement de Solutions de Plutonium
Pour une Expérience Critique," CEA-808, Centre d'ftudes Nucléaires

de Saclay, 1958.

J. Bertrand, P. Bonnaure, C. Clouet d'Orval, J. Corpel, J. de Lamare,
P. Lecoustey, I. Prevot, R. Roche, M. Sauve, J. Tachon, and G.
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Plutonium,'" Proc. 2nd U. N. Int. Conf. on Peaceful Uses of Atomic

Energy, 12, pp. 539-562, United Nations, New York, 1958.

W. S. Hogan, '"Homogeneous Reactors,”" in U. S. Research Reactors,

Operation and Use, J. W. Chastain, Jr., ed., Addison-Wesley Publ.
Co., Reading, Mass., 1958, pp. 1lll-46.

A. W, Kramer, Boiling Water Reactors, Addison-Wesley Publ. Co.,
Reading, Mass., 1958, p. 31.

42

13. '"Water Boiler, Los Alamos Scientific Laboratory,'" LADC-442,
(MDDC-1424), LASL, declassified Oct. 27, 1947.

14, R. F. Christy, "Theoretical Discussion of a Small Homogeneous Enriched
Reactor,' MDDC-72, Institute of Nuclear Studies, University of
Chicago, Decl., June 18, 1946.

15. R. F. Christy, '"Neutronic Reactor," U. S. Patent 2,843,543, July
15, 1958. Filed Oct. 19, 1945.

l6. R. F. Wilson, J. O. Henrie, W. N. McElroy, W. E. Parkins, and J. W.

Flora, "Aqueous Homogeneous Type Research Reactors,'" Proc. 2nd U. N.

Int. Conf. on Peaceful Uses of Atomic Energy, 10, United Nations,
New York, 1958. pp. 404-18.

17. J. A. Lane, H. G. McPherson, and Frank Maslan, Fluid Fuel Reactors,
Addison-Wesley Publ. Co., Reading, Mass., 1958,

18. L. D. P. King, "Design and Description of Water Boiler Reactors,"

Proc. 1st U. N. Int. Conf. on the Peaceful Uses of Atomic Energy, 2,

United Nations, New York, 1956, pp. 372-391.

19. Ref. 17, p. 347.

20. "U. S. Manufacturers Offer these Reactors for Export," Nucleoniecs, 14,
Pp. 70-75, November 1956.

21. "An Enriched Homogeneous Reactor," Rev. Sci. Instr., 22, No. 7, pp. 489-99,
July 1951, (same as AECD-3059, LADC-887).

22. C. P. Baker, H. K. Daghlian, G. Friedlander, M. G. Holloway, D. W.
Kerst, and R. E. Schreiber, '"Water Boiler," AECD-3063 (LADC-819),
LASL, Sept. 4, 1944, decl. Feb. 21, 1951.

23. Ref. 17, pp. 340-347.

24. F. L. Bentzen, R. E. Carter, J. Hinton, L. D. P. King, J. C. Nevenzel,

R. E. Schreiber, J. W. Starner, and P. H. Watkins, "High-Power Water
Boiler," AECD-3065 (LADC-822), LASL, Sept. 19, 1945, decl. Feb. 27,
1951.

25. L. D, P, King, "A Brief Description of the Los Alamos Homogeneous
Reactor, SUPO Model of the Water Boiler," AECD-3287 (LADC-1081),
LASL, October 1951, decl. Dec. 27, 1951.

26, L. D. P. King, "The Los Alamos Homogeneous Reactor, SUPO Model,"
LA-1301, LASL, Feb. 7, 1952.

27. L. D. P. King, '"Water Boiler Reactor," U. S. Patent 2,961,391, Nov,
22, 1960.

43
28.

29.

30.

31.

32.

33.

34.

35.

36.

37.

38.

39.

40.

41.

42.

E. A, Moore, Jr., "Operating Manual for the AE~6 Reactor," NAA-SR-
MEMO-5395, Atomics International, June 23, 1960.
R. Chalker, '"Design and Construction of Water Boiler Neutron Source,"

Science, 119, pp. 9-15, Jan. 1, 1954,

M. E. Remley, "Operation of a Water Boiler Neutron Source,” NAA-SR-839,

Atomics International, Nov. 20, 1953.

]

H. Pearlman, '"Reactors for General Research Use," in Proc. Univ.

Research Reactor Conf. Held at Oak Ridge, Tenn., Feb. 17-18, 1954,

W. W. Grigorieff, ed., AECU-2900, Oak Ridge.

IAFA Direction of Nuclear Reactors, Volume II, Research Test and

Experimental Reactors, International Atomic Energy Agency, Vienna, 1959,

A, T. Biehl, T. Fahrner, and S. Kash, "A Water-Boiler Reactor as a

Source of Neutrons for Exponential Experiments," Nuclear Science and

Technology (Extracts from Reactor Science and Technology, 2,

Issues 1-4, April-December 1952), pp. 69-76, TID-2503, Del.

M. E. Remley, "Preliminary Outline of the Critical Assembly and
Initial Operational Testing of L-3 Reactor," NAA-SR-MEMO-784, Atomics
International, Sept. 25, 1953.

"Description of a Homogeneous Solution Type Research Reactor,"
NAA-AER-1023, North American Aviation, Inc., June 23, 1954,

Ref., 17, pp. 347-348.

"Starting Nuclear Sixties; Who's Building the Civilian Reactors,"

Nucleonics, 18, No. 1, p. 19, January 1960.

L. Reiffel , "The First Industrial Research Reactor Facility Design,

Operational Experience, and Research Programs," Proc, 2nd U. N. Int. Conf.

on Peaceful Uses of Atomic Energy, 10, pp. 393-403, United Nations, N.Y., 1958,

"Nuclear Reactor for Medical Research," North American Aviation, Inc.,
NAA-AFER-1180, Jamuary 1955.

J. E. Gilligan, Jr., R. S. Hart, and O. D. Seawell, "Hazards Summary
Report, L-77 Laboratory Reactor," AL Memo-2387, Atomics International,
Jan. 20, 1958.

J. W. Flora, "Educational Uses of the Small 5-Watt Laboratory Reactor,"
Proc. 2nd Nuclear Eng. and Sci. Conf., 1957, Philadelphia, Advances in

Nuclear Engineering, 2, p. 83, Amer. Soc. Mech. Engrs., Pergamon Press,
1957.
"Reactor Included in Beirut Exhibit," ANS Nuclear News, 4, No., 11, p. 26,

November 1961.

44

43.

4b,

45.

46.

470

48.

49,

50.
51.

52.

53,

54,
55,

56,

57.

"A 5-W Laboratory Reactor," Nuclear Eng., 1, No. 8, pp. 344-5,

November 1956,

J. H. Lampe, 'Nuclear Engineering Program at North Carolina State
College," TID-7527, pp. 134-142, 0TS, 1956.

C. K. Beck, A. C. Menius, Jr., R. L. Murray, Newton Underwood,

A. W. Waltner, and George Webb, "Further Design Features of the
Nuclear Reactor at North Carolina State College," AECU-1986,
(NCSC-46), North Carolina State College, January 1952,

c. M, Baldwin, Jr., D. B. Beman, H. B. Carter, H. A. Lamonds,

J. W. Meeks, A. C., Menius, Jr., R. L. Murray, E. J. Story, and

A, W. Waltner, "Summary Hazards Report for the North Carolina State
College Reactor NCSCR-4," NP-7041, North Carolina State College,
Nov. 5, 1958.

A. C. Hughes, "University Reactor Conference,'" Nuclear Power, 5,
No. 54, pp. 99-101, October 1960.

"The Role of Nuclear Reactors in University Research Programs,”

NSF-60-39, National Science Foundation, July 1960.

L. B. Borst and B. A. Mong, "The University of Utah Reactor Project,"
in Reference 31.

Nucleonics, 12, No. 4, p. 11, April 1954.

D. Breton, '"More French Reactors," (letter to the editor), Nucleonics,

19, No. 1, p. 6, January 1963.

J. W. Ullmann, "Foreign Research and Power Reactor Preliminary List,"
CF-59-2-76, ORNL, Feb. 26, 1959.

"Reactors Developed in the USSR and Its Bloc Countries,'" AD-248402,
ASTIA, Oct. 31, 1960, p. 43.

J. British Nuclear Energy Conference, 2, pp. 395-407, October 1957.

H. M. Busey and R. P. Hammond, '"Test Tube Research Reactor,'" Nucleonics,

13, No. 11, November 1955, pp. 72-75; Ref. 18.

"Reactor Studies--Final Report. Part I," TID-10117, Babcock and
Wilcox Co., November 1955, pp. 106-113.

E. D. Arnold and A. T. Gresky, "Exploratory Study: Homogeneous
Reactors as Gamma Irradiation Sources,'" CF-56-6-107, ORNL, July 1956.

45
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59.

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62.

63.

64.

65.

66.

67.

68.

69.

R. B, Briggs and J. 0. Kolb, "A Homogeneous Reactor Gamma Irradiation
Facility," CF-56-7-126, ORNL, July 27, 1956.

P. R. Kasten, M. I. Lundin, C. L. Segaser, R. E. Aven, D. R. Gilfillan,
R. F. Hughes, M. C. Lawrence, H. A. McLain, R. A. McNees, C. Michelson,
and C. W. Nestor, "Aqueous Homogeneous Research Reactor——Feasibility
Study," ORNL-2256, ORNL, May 9, 1957.

W. A, Gall and M. I. Lundin, "Homogeneous Reactors. Design

Section Progress Report for June 1956,'" CF-56-7-63, ORNL, July 13, 1956,
Pp. 9-13.

M. I. Lundin, "Homogeneous Research Reactor," in "HRP Civilian Power
Reactor Conference Held at Oak PRidge National Laboratory, May 1-2,
1957," TID-7540, OPNL, July 1957, pp. 288-301.

"A Selection Study for an Advanced Engineering Test Reactor,'" AECU-3478
(U-047), Aeronutronic Systems, Inc., March 29, 1957.

F. T. Barr and J. F. Black, "Homogeneous Nuclear Reactor with Sub-
critical Exterior Fuel Circulation Zone,'" Canadian Patent 614,567,

Feb. 14, 1961.

H. J. Hibshman, "Homogeneous Nuclear Reactors,'" Canadian Patent No.
613,636, Jan. 31, 1961.

R. G. Mallon, J. Saldick, and R. E. Gibbons, '"Conceptual Design of an
Advanced Engineering Test Reactor,” NYO-4849, (ASTRA-200-E-1.1),
Advanced Scientific Techniques Research Associates, March 1, 1957.

R. G. Mallon and W. R. Pearce, "A High-Flux Research Reactor for
Large-Volume Irradiation," Nuclear Science and Technology, lA,

Issue 2, pp. 329-334, August 1955, (TID-2507 Del.).

W. M. Breazeale, '""Research Reactor Program at the Pennsylvania State

University,'" Nuclear Engineering, Part II, Chem. Eng. Progress

Symposium Series, 50, No. 12, pp. 6-10, Amer. Inst. Chem. Engrs.,

1954,
J. A. Lane, "Where Reactor Development Stands Today,' Nucleonics, 14,
pp. 30-37, August 1956.

J. A. Lane, "Aqueous Fuel Reactors," Chapter 19 in Reactor Handbook,

Vol. IV. Engineering, 2nd ed., Stuart McLain and J. H. Martens, eds.,

Interscience Division, John Wiley and Sons, New York, 1964,

46

70. H. H. Cappel, "Hazards Summary Report for the Walter Reed Army Medical
Center Nuclear Research Reactor," AI-3739, Atomics International,
May 6, 1959.

71. "Solution-Type Research Reactor Has Unique Feature," Atomics, p. 15,
March-April 1963.

72. Atomics, p. 4, March-April 1963.

Additional References

M. Monet, "Homogeneous Pile Water Activity," Final Report, MonE-64,
ORNL, Feb. 8, 1946, decl. Jan. 30, 1956, p. 13.

K. A. Hub, "Discussion of Homogeneous Reactor Possibilities,'" AECU-3336
(IC-KAH-56-5), Internuclear Corp., May 16, 1956.

""Research Reactors," Nuclear Engineering, 5, No. 46, pp. 99-104, March 1960.

"World Reactor Chart, Third Edition," Nuclear Power, 1, No. 69, January 1962.

F. E. Croxton, "List of References on Homogeneous Reactors,'" TID-299,

USAEC, Oak Ridge, Jan. 31, 1955.

Musa Halev, "Survey of Reactor Facilities for Radiation Biology Research,"
NADC-MA-5813, U. S. Naval Air Development Center, Oct. 22, 1958.

47
Chapter 3 Non-Boiling Reactors for Power and Breeding

The reactor concepts in this chapter include those in which a
homogeneous aqueous solution or slurry (suspension) is utilized as fuel.
These fuels were among the first to be suggested, but reactors employing
them have not been developed beyond the prototype stage. An excellent
description of aqueous solution and slurry reactors is given by Lane,
MacPherson, and Maslan.! Also valuable are reviews by Briggs and
Swartout? and by Brown and Morris.S3

These reactors can be classified in many ways-—--purpose, method of
heat removal, origin, materials used, etc. Table I (Chapter 1) 1lists the
different types and their applications.“ A useful classification is by
fuel utilization--burners, converters, or breeders. Breeders are either
one-region--consisting of a homogeneous solution or slurry; or two-region--
a core plus a blanket of fertile material. In the course of the Homo-
geneous Reactor Experiment series, the Homogeneous Reactor Project, and
the Los Alamos Power Reactor Experiments, groups of closely related
concepts were developed. Thus, they are discussed together, rather than
under their respective classifications. Other homogeneous aqueous
solution reactors are then grouped as burners, converters, or breeders.
The closely related concepts of non-homogeneous aqueous reactors and
colloidal fuels are briefly discussed, as there is little in the
literature concerning them. Slurry-fueled reactors are treated separately
in this chapter because of their special features. For boiling homo-

geneous aqueous reactors, see Chapter 4.

Early History

It is hard to determine who proposed the first homogeneous aqueous
reactor for power and breeding. Early in the Manhattan Project, aqueous
homogeneous solutions and slurries were considered as fuels for producing
plutonium but, as with research reactors, the need for enriched fuel and
large quantities of heavy water made heterogeneous reactors preferable
at the time. When enough of heavy water and enriched fuel became available,
aqueous solutions and slurries were studied for breeding and power
production. A group of chemists at the Clinton Laboratories, now the Oak
Ridge National Laboratory, were active in planning homogeneous reactors

for research and breeding. In 1944, Nordheim reported on the potentialities

49
of a breeder reactor.® In an extensive experimental program, research
workers studied the choice of fuel, moderator, and cooling mechanism;
problems of corrosion, poisoning, and formation of radiolytic gases; and
most of the other aspects of reactor development.

Investigative work continued through 1945. At that time, such
difficulties as bubble formation, corrosion, solution instability, and
external holdup of fissionable material led to the decision to consider
heterogeneous reactors instead.® The early homogeneous aqueous reactor
designs are discussed under their respective classifications, and details
are given on the data sheets. Many of them utilized a heavy water solution
of sodium uranyl carbonate or uranyl sulfate as fuel. Thorium metal or
oxide was a common fertile material. They were cooled by circulating the
fuel solution to external heat exchangers.7’8

Major experimental efforts on power reactors apparently began in 1950
at Oak Ridge National Laboratory. At that time, members of the staff
began the design and building of a pilot plant for a fluid-fuel reactor,
Homogeneous Reactor Experiment No. 1, (HRE-1). Most of the later interest
in homogenecous aqueous sclution reactors arose after the successful
operation of HRE-1, which removed most of the previous concern that
difficulties in operation and limitations on power production would make
such reactors impossible. In spite of the many new reactor concepts based
on the HRE experiments that developed, HRE-1 and HRE-2 were the only

reactors built in the series.

HRE-1 and Related Concepts

Before the first Homogeneous Reactor Experiment (HRE-1), preliminary
pilot designs, were proposed and, after the success of HRE-1, a series of
related concepts followed.

HRE-1 came into being partly owing to the doubt that a high-powered
aqueous homogeneous reactor could be operated at all because of the high
gas production and consequent bubbling. The purpose was to investigate
circulating uranium solution at high enough power and temperature to
produce electricity from the heat released. Some of the nuclear and
chemical problems studied included not only the bubbling problem, but
also the corrosion of the reactor-core materials and the tendency of the

fuel solution to become unstable, causing either precipitation of

50

hydrolysis products from the fuel or formation of two fuel phases, one
being an extremely corrosive heavy liquid. Varying core arrangements
and supercharging with oxygen were considered as means of reducing
corrosion, and adding sulfuric acid made the fuel stable at higher

tem.peratures.9

Preliminary Work

Weinberg has described 20-kw(e) pilot model for designing larger
homogeneous reactors.l!® It was suggested that the long-range homogeneous
reactors would probably be moderated with heavy water but, since ordinary
water has a much smaller slowing-down length than heavy water, substitution
of ordinary water for heavy water would produce a scaled-down reactor.
This scaled-down model would reproduce all the essential features of the
full-scale plant; hence it would be ideal as a pilot plant. The pilot
model was to have 2 kilograms of 93.4 percent enriched uranyl sulfate in
ordinary water solution, and the reacting solution was to be pumped
through the reactor vessel and heat exchanger. High temperature (400°F)
and pressure (1000 psi) would be sufficlent to keep the bubble volume
manageably small and to give a 10 percent over-all power efficiency. A

heavy-water blanket as a reflector under 1000 psi surround the reactor.

The HRE Project

A full-scale reactor was not used at the start but rather a reactor
experiment, which is defined as:!!

..""a reactor in the research and development program, usually
producing less than 10 Mwt, designed for the limited purpose of
testing the technical feasibility of a reactor concept or some
unique reactor feature or piece of equipment. A reactor
experiment 1s built with the intention of making changes in fueling
structure, assoclated components, or other components of the
reactors."

According to another definition:!2

"A reactor experiment makes no attempt to incorporate features
which would be a necessary part of any fullwscale reactor.

Reactor experiments have, as their purpose, the testing of an idea
or a unique piece of equipment. As a matter of fact, a reactor

experiment is just one phase of the research and development

51

program and should be considered as such. The size is immaterial,

but, generally, reactor experiments will be small with generation

capability of less than 10 Mw. Indeed, the actual generation of
electricity is unnecessary unless it is a vital part of the
experiment."”

It was decided to build, at Oak Ridge, HRE-1, a 1000-kw, uranyl sulfate
reactor. The primary purpose was to determine whether stable operation at
the 20-kw/liter level was possible at all; all other considerations, like
ease of maintenance, corrosion, and chemical reprocessing, were almost
ignored because it was believed that the primary question could be
answered quickly, after which the experiment could be dismantled. The
core of HRE-1 was a Type-347-stainless-steel-sphere, 18 inches in diameter,
through which circulated a 0.17-molar highly enriched solution of uranyl
sulfate in ordinary water at 1000 psi, with an outlet temperature of
250°C. The core was surrounded by a 10-inch-thick blanket of heavy water.13-15

HRE-1 achieved its full power rating of 1000 kw on February 23, 1953,
becoming the second reactor to produce electricity. It reached a maximum
power of 1600 kw on January 29, 1954, It was dismantled two years after
it was built, largely because it had served its primary purpose and there
were no special provisions for maintaining it. It demonstrated that a
high-powered aqueous homogeneous reactor could be run stably, and it
resolved the original question whether gas production would make aqueous
reactors a priori unworkable.l® The HRE-1 concept, with heavy water
possibly substituted for ordinary water, was considered for submarine
propulsion.!’ For details on the design of HRE-1, see reference 18,

This experience led to a more ambitious experiment, HRE-2, also known
as HRT, the Homogeneous Reactor Test.19 It first achieved high-power
operation on February 8, 1958.13 HRE-2 had a Zircaloy core vessel and
the intention was to try thoria slurries as well as heavy water in the
reflector.?? During March and April 1958, while the reactor was being
operated at high power (maximum power of 6.3 Mw), a hole was melted in
the core tank after 30 minutes of operation at 6.3 Mw. The hole resulted
from an unsatisfactory hydrodynamic design of the core and the high-
temperature instability of the fuel. 1In addition to the hole, other
damage included severe local overheating of some metal surfaces. Hence,

more emphasis was placed on core design and hydrodynamics studies in the

52

development programs.?! See also references 22 and 23.

The core was inspected and, because the hole was inaccessible, nuclear
operation was resumed; core and blanket were operated together as a one-
region reactor. The core was repaired by patching in late 1960; by this
time there were two holes in the core vessel. HRE-2 attained a power
level of 5 Mw on January 4, 1961 with no indication of the chemical
instability previously noted. However, work on HRE~2 ended when the AEC
terminated the entire homogeneous program on June 30, 1961, After the
reactor was shut down, inspection revealed that one of the core patches
had failed.2%:23

After HRE-2 was shut down and before the Homogeneous Reactor Project
was terminated, several preliminary or conceptual design studies were made
of a 5-10 Mw experimental replacement reactor suitable for installation
in the existing HRE-2 facility with a minimum of system changes. This
design included a cylindrical core with a beryllium reflector surrounded
by a blanket of thorium pellets enclosed in a removable steel pressure
vessel. The fuel was a solution of 6.56 grams of uranium-233 per liter.26728

HRE-1 and HRE-2/HRT were the only reactors to reach the hardware stage,
unless the HRT mockup be also considered a reactor. Seven other reactors,
however, were conceived in the pattern of HRE~-1. In order of increasing
size, the nine are: HRE-1, HRE-2 as conceived, HTR (the Homogeneous
Thorium Reactor), HRE-2/HRT as built, HRT as conceived, HRE-3, HRE-4,

ISHR (the Intermediate~Scale Reactor), and TBR (the Thorium Breeder
Reactor). These concepts and closely related ones have been discussed in
several reports,29733

HRE-2 and HRT as conceived are related to the HRE-2/HRT actually built
(hereafter referred to as HRT): The development program for homogeneous
reactors as set forth in the original "Five Year Plan'" by the AEC called
for two intermediate development steps after the HRE-1: the HRE-2Z and
the HRT. It was decided, however, to extrapolate the HRE-2 to an
intermediate size, with a more-flexible experimental plant, compromising
between its original size and that for the larger HRT. 3"

The HRT design has been reviewed.3® The distinction between HRE, HRT,
and TBR (the Thorium Breeder Reactor) can be clarified by analogy with the
fast-breeder program. The Homogeneous Reactor Program resembled the

fast-breeder program in its general aspect of building a succession of

53
larger reactors ending in the "“ultimate" full-scale reactor. Analogous to
the EBR, EBR-~2, PBR sequence are the HRE, HRT, TBR. The final stage of the
Homogeneous Reactor Program would involve a full-scale breeder reactor.3°®

The HTR, or Homogeneous Thorium Reactor, preceded the HRT as a concept.
It was judged unnecessary to carry this intermediate-scale reactor to the
hardware stage, and it was therefore superseded by the Homogeneous Reactor
Test, HRT. The Homogeneous Thorium Reactor was planned to produce 65,000
kw(t) of which 16,000 kw would be converted to electricity. A blanket of
thorium, from which uranium-233 would be produced, was planned. The core
would not be as large as that for a full-scale reactor but the thickness
of thorium blanket and concentration of fertile material would be the same
as for a central power station of this type.37

The third phase of the homogeneous program—-presumably after HRE-1
and HRE-2 were conceived-~was the medium-sized Homogeneous Thorium Reactor
(HTR), a thorium breeder fueled with uranium-233 to produce 65,000 kw(t).37238
The HRE-2 concept was later modified to be more useful and more powerful
than HRE-1, with power levels of 10 megawatts rather than 3 megawatts
planned. It was more versatile, and by 1957 included a blanket system
with chemical-reprocessing equipment as part of the reactor. Thus it was
judged unnecessary to construct an intermediate-scale homogeneous reactor,
because the modified HRE-2 would provide most of the technological
information originally expected from the intermediate-size reactor.3?

At an earlier date, a stage intermediate between HRT and TBR was
considered. Research and design leading to construction of an intermediate-
scale homogeneous reactor was undertaken. The reactor was conceived of as
operating at a maximum of 40 Mw and to be of a size approaching that of a
full-scale reactor, i.e., with a core tank 8 to 10 feet in diameter,
compared to a diameter of 15 to 20 feet for a full-scale reactor."?

This Intermediate Scale Homogeneous Reactor (ISHR)“1:%2 went through
several variations. Its design was started in 1951-1952 to provide the
information necessary for reasonable extrapolation from the HRE to a full-
scale production reactor. Several types of blankets were considered for
the ISHR, especially fluidized thorium oxide."3 Further work was deferred
late in 1953, however, when it became evident from HRE-1 and the associated
development program that construction of a second homogeneous reactor

experiment would be more suitable.“"

54

Thus the ISHR, like HTR, was replaced by the HRT program, and the
sequence ran from HRE to HRT and TBR.

ISHR was intended to demonstrate nuclear, chemical, engineering, and
economic feasibility of an aqueous homogeneous solution reactor by actual
testing on a scale smaller than that of the final full-scale reactor. >
Some of the problems inherent in the power reactor but not apparent in the
HRE-1 or HRT might be expected to appear at this point.

The problem of core construction, for example, was considered with the
view of assuring feasibility of and economy in construction, stability of
flow, adequate heat removal from all points, low pressure drop (preferably
less than 15 psi), low gas holdup, and velocities near the walls consistent
with corrosion problems. Various core configurations were considered,
including straight-through and spherical shapes, but no specific choice
was made prior to 1953, the end of the effort on ISHR.*®

Designs based on the ISHR include the Cooling Tower Reactor,L+7 a

two-region converter,“8s%% a one-region breeder,SO a two-region breeder,51

and a three-region lithium converter.>?

The Thorium Breeder Reactor (TBR), was intended to develop the concept
of a thorium-uranium-233 breeder power station in sufficient detail to
provide the basis for detailed design by an architect—engineer.2

One design reported53 was for a power station, consisting of three
reactors, each delivering 100 Mw. They are spherical, two-region reactors,
with enriched uranyl sulfate solution in heavy water as fuel and a slurry
of thorium oxide in heavy water as fertile material.

In 1955, it was decided that, because of insufficient data, intelligent
study of a specific reactor was not possible and the original program was
modified to include a preliminary investigation of four variants of the
thorium breeder reactor, in the hope that advances in technology would
permit selection of one of them for further study later.>% The TBR design,
however, was set aside because of urgent work needed for the HRT.?3

HRE-3 was designed to provide operational and technical data and to
demonstrate technical feasibility of an intermediate-scale, two-region
aqueous homogeneous power breeder, but it was interrupted to apply
additional effort to problems encountered in slurry-fueled reactors, and,

to await resclution of problems that had arisen in operation of the HRT. 2226

HRE=-3 had been intended to study the two-region power concept on a small

55

scale, starting in 1961 with a 50-Mw prototype plant, with further development
depending on the success of the prototype.>’

In HRE-3, the fuel was uranyl sulfate solution in heavy water, with a
blanket slurry of thorium oxide in heavy water, a 50-Mw core, a 10-Mw blanket,
and an electrical output of 19 Mw,.>°

HRE-4 was a slurry-fueled reactor, which will be described under that

category.,

The Homogeneous Reactor Project

The ORNL Homogeneous Reactor Experiment Project (HRE) was reconstituted
in about 1951 as the Homogeneous Reactor Project (HRP) to study the
feasibility of a full-scale homogeneous reactor to produce plutonium°58
Many alternative designs were studied, and attention was given to such
problems as the behavior of solutions and slurries in reactors and corrosion
difficulties. For example, in 1956, members of the project reported on HRT
and five other possible designs: a U235 burner fueled with uranyl sulfate
in heavy water; two one-region U238-Pu converters, one fueled with uranyl
sulfate solution, the other with a slurry of uranium trioxide and plutonium
dioxide in heavy water; and two Th-—U233 breeders, a two-region and a one-region,
one fueled with solution, the other with slurry.®? Also a breeder with a
blanket of thorium oxide pellets was later suggested.3? Designs for two
other solution-fueled reactors were for power reactors producing 380 Mw(e).00,61
The project, along with the entire subject of fluid-fuel reactors, was later
evaluated, as described under "“Evaluation and Status."

A typical aqueous homogeneous reactor concept developed for evaluation
purposes is the Aqueous Homogeneous Breeder Reactor.®2 This concept resembled
many of the thorium breeders, with its aqueous uranyl sulfate solution as
fuel and a slurry blanket of ThO2 pellets. It especially resembles the
reactor described in reference 33, Special provisions were made in the
design of the blanket structure. A variation, in which thorium tetrafluoride
pellets replace the thorium oxide in the blanket, was suggested to obtain
higher fuel yields and lower fuel-cycle costs. Much of the claimed lower
cost would be due to less=expensive fuel reprocessing.53

Boiling was considered as a means to remove power from the thorium oxide
blanket of large two-region breeders like those studied by HRP, but it was

rejected in favor of circulating the slurry.®"

56

The LAPRE and Related Reactors

Two power=reactor experiments, somewhat analogous to HRE-1 and HRE=2,
were carried out at the Los Alamos Scientific Laboratory. 1In Los Alamos
Power Reactor Experiments I and II (LAPRE-I and LAPRE-IT), the fuel
solution consisted of uranium oxides dissolved in aqueous phosphoric acid,®3-68

Phosphoric acid solutions of uranium have several advantages that make
them attractive as reactor fuels. Their low vapor pressure and high thermal
stability permit operations at higher temperatures than are possible with
other aqueous solutions. Still other advantages are the high hydrogen
density and low cross-section for neutron absorption, which permit the use
of small, compact reactors and low fuel inventories. There is, however, a
grave disadvantage., The extreme corrosivity of the acid solutions makes the
choice of containment materials very difficult and requires unusual pre-
cautions. LAPRE~T and LAPRE-II were designed to determine whether such
solutions were practical reactor fuels and to design reactors that could best
use them,

The cylindrical form of the LAPRE reactor vessels has led to the name

"Test=Tube Reactors" being applied to them,
LAPRE=T

LAPRE~T (originally designated DIR-P) was intended to be a compact,
high-temperature nuclear steam generator of simple and compact design, in
which the reactor and heat exchanger would be in a single vessel. Thus,
external circulation of highly radiocactive solution is avoided. The fuel
solution was a solution of highly enriched uranium trioxide (111 grams U235
per liter) in aqueous phosphoric acid containing 53 wt.%Z acid. The cylindrical
reactor had a critical zone above the cooling coils over which solution was
circulated by a pump. Oxygen overpressure was used. Below the coils was a
reservoir region, which contained a boron rod to prevent criticality and
permitted storage of excess solution. To minimize corrosion, the vessel and
all components exposed to the acid were gold-plated, except for certain items
made of or coated with platinum.66,%7

The reactor went critical in February 1956. A maximum operating
temperature of 390°C at about 150 kw was achieved before a failure in the
heat-exchange system ended the test, Inspection showed that the gold cladding
on one of the exchanger tubes had ruptured. No more tests were made and the

reactor was dismantled in January 1957, The experience, however, provided

valuable information for the design of LAPRE-II,66,67

57
LAPRE-II

LAPRE-II differed in several ways from LAPRE~I. A more concentrated
solution of phosphoric acid was used--95 wt.Z Instead of 53. The vapor
pressure of the resulting solution, even at 450°C, was only about 800 psi,
compared with 4500 psi for the LAPRE-I solution. Because of its reducing
properties, this solution attacks only slowly any metal below hydrogen in
the electromotive series. The lower vapor pressure of the LAPRE~II solution
permits a reactor vessel of thinner walls. Instead of pumping the solution
for cooling, circulation over the cooling coils was by natural convection.

An external annular reflector of graphite and beryllium was provided. Gold
cladding was again applied to exposed surfaces.®8,62

Operation of this experiment began in February 1959 and continued into
May of the same year. The maximum fuel temperature measured was 826°F. Full-
power operation was reached on April 22. The maximum power was 745 kw, at
745°F and 670 psi. The operation was satisfactory at widely varying power
demands, and the power was limited only by the capacity of the primary cooling
loop. The gold cladding was shown to be a possible answer to the corrosion
difficulty, but improvements in materials and fabrication methods are needed.
As an advanced concept, the use of impervious graphite as container material
was suggested, and in one reactor concept, such graphite would be material for
the common container and primary heat exchanger.68,63
A version of LAPRE-II, with a spherical core, increased power, and

other modifications were proposed for application to the DEW line.”’®

Sandia Phosphoric Acid Reactor

A model employing the LAPRE concept has been designed by Sandia
Corporation for remote locations or mobile units.’! It would generate from

100 to 1500 kw(e).

Other Homogeneous Aqueous Solution Reactors

Many concepts for aqueous reactors were developed after the success of
HRE~1, and some were developed even before the completion on this experiment.
A few were one-region converters or breeders, but most were one-region

burners or two-region breeders.

One-Region Burners

One~-region burners have received considerable attention as sources of

industrial power. Published concepts are similar in several ways. In nearly

58
all, for example, highly enriched uranyl sulfate solution in water is the
fuel, and a system pressure of 1000-2000 psi is usual.

Much of the earlier work was on mobile reactors.

A reactor to produce 250 Mw(t) was described in a 1950 report by workers
at ORNL.’2 It utilized enriched uranyl sulfate in light water and super-
heating steam by burning radiolytic gas, and it is otherwise similar to most
other burners., Two variations were proposed. Each differed from the
original only in such details as fuel enrichment and concentration.

Solution power burners have been considered for locomotion. A study
at the University of Utah found that a locomotive powered by a nuclear
reactor would be technically feasible.’?® The reactor would produce 30,000 kw
as steam at 250 psig to operate a 7200-hp locomotive., The reactor contains
a core of fully enriched uranyl sulfate solution within two hexagonal slabs,
Adjacent to each slab is a water reflector connected to heat=transfer tubes,
which penetrate the core. Water circulates through tubes and reflectors for
cooling. Control is by means of rods in the core. Likewise, a Russian
design for a locomotive, powered by a reactor fueled with a water solution
of a uranium salt, has been described.’" The economic feasibility of nuclear-
powered locomotives has, however, been questioned.’® A Wright-Patterson study
of 1959, concluded that, for aircraft and remote applications, in general the
problems of aqueous homogeneous reactors were not immediately surmountable, /®

Students at the 0Oak Ridge School of Reactor Technology reported in a
term paper a design for a solution burner for ship propulsionc77 In the
generally conventional design, a spherical core with concentric inlet and
outlet was included because its flow pattern would be stable under the
rolling conditions aboard ship.

Three other reactor concepts, for producing power in remote locations,
developed by OSORT students were: a 10-Mw reactor;’8 "STUPQO"=-so named
because of an unsuccessful attempt to extrapolate the data of LOPO and SUPO; 7°
and an 80-Mw reactor.®% The first reactor is claimed to be unique in that
construction of a prototype is feasible without additional development. The
STUPO project considered a reactor with a maximum power of 10 Mw and
reliability for use in remote locations. The study concluded that feasibility
depended on whether 10 Mw of heat could be removed by natural convection,
Design power output of the 80-Mw reactor is 20 Mw(e), with an 80 percent plant
factor.

One-region burners include such utility types as the Wolverine Electric

Cooperative Reactor and those proposed by General Electric and Westinghouse,

59
The Foster Wheeler Company proposed in 1955 to construct an aqueous solution
reactor for Wolverine,®! but the project was cancelled in May 1958 because
new projections showed an excessive increase in cost. In December 1957
engineers from Oak Ridge, Sargent and Lundy, and Foster-Wheeler redesigned
the reactor, using highly enriched uranyl sulfate in heavy water, under
pressure to prevent boiling.82 Uranium-235 would have to be added periodically
to the reactor because of depletion of the original fuel by fission,83

The General Electric 100 kw(e) power plant had a circulating fuel
solution of highly enriched uranyl sulfate in light water contained in a
titanium~lined carbon-steel pressure vessel, which also served as a
reflector.8* The vessel was cylindrical, with a hemispherical bottom.
Control was by the temperature coefficient of reactivity of the core. The
proposal included a standard straight tube-and-shell heat exchanger. Many
of the details of this design were based on data from the Oak Ridge Homogeneous
Reactor Project. Westinghouse described a one-region 80,000 kw uranium-235
reactor, in which the fuel was a dilute uranyl sulfate solution.®> Corrosion
was the major consideration in selecting most of the design parameters.
Another Westinghouse concept of this type is described in reference 86,

Babcock and Wilcox considered both a non-boiling and a boiling solution
burner.8? The low-power non-boiling system consists of a uranyl sulfate
solution in a cylindrical vessel under an oxygen pressure of 1200 psi.

Heat exchange is through bayonet cooling tubes immersed in the solution.
Kasten and Claiborne®® studied fuel costs for spherical one-region
power reactors containing 90 percent enriched uranyl sulfate in either heavy

or light water and producing 25 Mw(e). The feed is also 90 percent
enriched, They concluded that the technology of this type was fairly well
developed as a result of the homogeneous-reactor program, and in fact

that the technology seemed sufficiently developed to warrant serious

consideration of the construction of these burner reactors.

One-Region Converters and Breeders

Some of the earliest concepts developed under the Manhattan Project
were of this type, usually with uranyl sulfate solution as fuel, with heavy
water as the moderator.®® A one-region breeder similar to HRE-1 was the
Homogeneous Power Reactor.?? It was fueled by a solution of plutonium-239

in heavy water, but few other details are available.

60

In 1950, the ORNL Long-Range Planning Group proposed a reactor for

power and plutonium production.?!

A solution of enriched uranyl sulfate
in heavy water flows through a spherical reactor into high-pressure
tubing, through the reactor shield, and into heat exchangers.

In 1952, the Large Scale Homogeneous Reactor concept was published at
Oak Ridge.%? 1In this reactor, designed for plutonium production, the
fuel was slightly enriched uranyl sulfate in heavy water, under a pressure
of 1000 psia. A similar reactor has also been proposed.93 Corrosion
was the chief difficulty with both.

A study of the economics of one-region breeders compared with two-
region breeders?“ showed that, for a three-reactor power station producing
375 Mw(e), the cost of power from a one-region reactor station would be
0.9 mills per kwh higher than that for the two-region station--7.1 mills
per kwh vs. 6.2 mills per kwh. In another study,>° one-region solution
reactors were compared with slurry reactors. A one-region spherical
solution reactor operating at 330°C and delivering 300 Mw(e) was found
to be competitive with slurry-fueled reactors.

A Russian reactor that is apparently a one-region converter has
been described in surveys of Russian reactors.2s97 In this homogeneous
reactor, designed as a prototype of a power station with high output,
uranium enriched to 0.9 percent U235, apparently as an oxide suspension
in heavy water, is the fuel. The pressure of the coolant is from 20 to
50 atmospheres, and the maximum temperature is 260°C. The power is from
1,150 to 2,000 Mw(t) or 280-500 Mw(e). In the design stage in 1960,
the reactor project was expected to require seven to eight years. For
this period, the consumption of natural uranium was estimated to be

306 tons, and of 0.95 percent enriched uranium, 43 tons.

Two-Region Converters and Breeders

Two-region solution reactors, some of which generally resemble the
Thorium Breeder Reactor of the HRE series, have received much attention
and many concepts have been published.

Several concepts were developed by workers in the Manhattan Project,%’99
In three of them, the fuel is enriched sodium uranyl carbonate dissolved in
heavy water. The fertile material is thorium metal, either in a slurry or
as rods. One, the Heavy Water Moderated U233 Pile, was designed to produce

100 Mw. A few others, generally similar in basic concepts, have also been

61
described.100,101

Wigner has described an unusual early concept.102 The core tank is a
U-tube, with cooling tubes that extend into it from both sides, with the
bottom portion free of these tubes. Pistons in each leg of the tube move
the solution from left to right alternately so that it comes in contact
with the cooling tubes. The U~tube contains a solution of a U233 salt in
heavy water, and a slurry blanket of thorium in heavy water surrounds the
bottom of the tube.

After World War II, and even before the completion of the HRE series,
work at Oak Ridge resulted in concepts designed for both breeding and for
electrical power.

An early power-breeder concept was developed at ORSORT. Imn the reactor,
a solution of U233 as uranyl sulfate in light water is the fuel and a
slurry of thorium tetrafluoride is the fertile material.l®3 Both core and
blanket are under 1000 psi. Students at ORSORT also developed a concept
for a reactor to produce both power and special isotopes.lO“

Two more ORSORT concepts were the Ultimate Homogeneous Reactor! 0> and
the High Temperature Homogeneous Reactor.19® The first was perhaps so
named because its purpose was to investigate the ultimate limitations of
aqueous homogeneous reactor technology and to assess, by means of a
design example, how far the homogeneous line of development could be
followed, After considering other types of reactors, the authors decided
on uranyl sulfate fuel; heavy water as coolant-moderator, with natural
circulation; and a blanket of a slurry of thorium dioxide in heavy water.
These conditions were chosen as best for higher core-power density,
higher thermal efficiency, and constant turbine steam conditions from no
load to full load in a 200-Mw breeder. The High-Temperature Homogeneous
Reactor differed especially in the fuel solution, in which uranium-233
trioxide and chromium trioxide form a solution in light water. The
pressure is also higher than usual--3500 psi. This high pressure permits
a higher outlet temperature of the solution. The design was for a plant
using two such reactors to produce a total of 290 Mw(e).

Work at ORNL produced concepts for power and breeding that had some
features that differed considerably from the earliest ideas. Visner, for

example proposed a two-region reactor without a core tank.!97 The uranyl

sulfate solution fuel is in the same contalner as the thorium oxide slurry

62

blanket. They are kept separate by operating the reactor as a centrifugal
separator, so that the solution remains in the center and the slurry at
the periphery. In the K-49 reactor,108 a solution of uranyl sulfate in
heavy water, rather than thorium metal or slurry, is the fertile material
in the blanket. A long-range concept developed at ORNL was the Thermal
Power Breeder.!9? It is a two-region breeder, in which uranium~233 as
uranyl sulfate is the fuel and a suspension of thorium oxide in heavy
water is the fertile material. This reactor is designed for steam
production in a power plant, with the additional feature of breeding more
uranium—-233.

Two other ORNL concepts aimed at power for commercial use were the
Project Dynamo concept and a reactor for central station power. Engineering
details for the first were developed by the Project Dynamo staff from an
ORNL concept.!1? Three such reactors, producing 450 Mw(t) each, were
designed for a power station. The fuel is a solution of uranyl (uranium-233)
sulfate in heavy water and the blanket a slurry of thorium oxide in heavy
water. One design for the central station power reactor utilized a
solution of uranyl sulfate in heavy water as fuel and a slurry of thorium
dioxide and uranium trioxide, also in heavy water, as fertile material,ll!
Power produced would be 100 Mw(e).

A two-region reactor developed at Oak Ridge was aimed at plutonium
production.!12 The blanket is a concentrated solution of natural uranyl
sulfate in heavy water. Plutonium of low plutonium-240 content could be
obtained by continuously processing the blanket solution. This reactor
experiment was planned to simulate reactor operations.

The Nuclear Power Group, composed of several private companies,
developed concepts for two-region power-breeder reactors. The group was
formed in October 1953 for broadening the scope of the previous nuclear
power studies by the individual companies--American Gas and Electric
Service Corporation, Bechtel Corporation, Commonwealth Edison Company,
Pacific Gas and Electric Company, and Union Electric Company of Missouri.
The initial objective was to select a reactor design that might be
constructed in the reasonably near future primarily to produce electric
power.113

Although the pressurized-water reactor seemed to meet thelr requirements

most closely, the Group conceived at least two homogeneous aqueous high-power,

63
blanket breeder concepts. One design consisted of a spherical core--6 feet
in diameter--of uranyl sulfate dissolved in heavy water. The core is
surrounded by a spherical blanket——2 feet thick-—of thorium oxide and uranium
slurry in heavy water.!1l* Uranium-235 could be used to start the reactor

and it could also be adapted to the uranium-238 - plutonium cycle. The
concentration, breeding gain, and reactor calculations were based on Oak
Ridge data,l1°7118 gpgq they were extrapolated to determine concentration

and power gain for this NPG reactor. Staff members at Oak Ridge also

alded NPG in completing this design, which is much like Oak Ridge designs.

By July 1957, the Nuclear Power Group, which now also included Central
Illinois Light Company, Illinois Power Company, and Kansas City Power and
Light Company, formed a joint study team with Babcock and Wilcox and
presented a conceptual design of a single-fluid, 150,000 kw(e) reactor
power plant developed by Babcock and Wilcox.112® To aveid the slurry
problems usually associated with homogeneous breeder reactors, a single-
fluid, two-region reactor with a blanket consisting of assemblies of beds
of thorium oxide pellets was selected. The assemblies would be periodically
rotated to equalize the breeding of uranium-233 in the pellet beds.

In developing the Homogeneous Aqueous Reactor (HAR) concept, British
workers studied many types before settling on the two-reglon power breeder,
similar in several ways to the Thorium Breeder Reactor proposed by ORNL.
More than one version of this two~region pressurized reactor have been
described, 120—123

A Japanese group considered an Aqueous Homogeneous Breeder Reactor
similar to that of ORNL.!2% The fertile material would be thorium oxide
slurry in heavy water so as to take advantage of Japan's abundant thorium
supplies.

A Russian non-boiling two-region reactor has been reported, but few
details have been given.l25 Plutonium is bred from a solution of a natural
uranium salt in heavy water enriched with plutonium produced in the same
reactor.

A concept for a Swedish two-region breeder is described in a French
patent.!?® It 1s claimed to have the advantages of both one- and two-region
breeders. Alternative fuels are a solution of uranyl sulfate, suspension of
uranium oxide, or solution of plutonium sulfate, all in heavy water. The

fertile material is either a slurry of thorium oxide, a slurry of uranium

64
oxide, or a thorium nitrate solution, also in heavy water. Means are
provided to exchange material between core and blanket and to separate

slurry into solids and liquid.

Non~-Homogeneous Aqueous Solution Reactors

Several reactor concepts that have been proposed include aqueous
solutions in a structured core in the presence of a massive moderator or
in a container.

Most of these will be included under "Semi-Homogeneous Reactors"
in Part II. One concept that will be covered here is the reactor in
which a uranium solution is contained in tubes. In a reactor designed
for power and for breeding uranium-233, a solution of uranium=233 in
heavy water is contained in tubes, around which a fertile slurry of
thorium=232 in (preferably) heavy water flows. The slurry extracts
heat from the fuel,!?7

Colloidal Fuels

Colloidal suspensions (sols) have been suggested as nuclear fuels,
No complete reactor concept has been developed, but workers at Ionics,
Inc., have proposed use of urania-thoria sols in water,!28 Particle
size is approximately 25 mu. They prepared such sols that were stable
for several hundred hours at 250-300°C., Because of their neutral
environment, very slow settling, and non-caking characteristics, such
sols are claimed to ameliorate the conditions of corrosion and phase
instability of solution fuels, and the settling, caking, and erosion
difficulties of slurries. No work, however, appears to have been done
on radiation stability or other properties that would be significant in

a reactor,

Slurry-Fueled Reactors

In general, aqueous-slurry (suspension) reactors are nearly defined
by their name. Many definitions have been given, but most differ only
in details. In this chapter, the definition of Lane, McPherson, and
Maslanl?9 will be used. Aqueous slurry fuels are suspensions of loose
and relatively independent clouds of joined particles (flocs) that are
large enough to settle. The particle size varies widely, depending on

the material, but 50 to 500 mp covers them gemnerally. They thus differ

65
from colloids, in which the particles are so finely divided that their
surface attraction forces exert a strong influence on the mechanical
properties of the material as a whole. Normally colloidal particles do
not settle out on standing.

Slurries attracted attention early as possible reactor fuels. In fact,
as mentioned in Chapter 2, Von Halban and Kowarski used a suspension of
uranium oxide in heavy water to demonstrate the possibility of a self-
sustaining nuclear reaction. Their several advantages include the
possibility of obtaining higher concentrations of fissionable material
and freedom from the corrosivity of acid solutions. Their use, however,
also has disadvantages, which include caking and settling of the solid
particles, erosion, and pumping difficulties. Solving these problems has
been a large part of the research and development on these reactors.

As with solution reactors, no commercial reactors using slurries have been
built, although several have been designed, and the feasibility of such
reactors has not yet been conclusively established.

The early history of slurry reactors generally parallels that of
solution reactors. There was early interest and extensive research, which
later diminished until the renewal of interest with the HRE project.

Lane, MacPherson, and Maslan have reviewed the early development.'

Workers at the University of Chicago, Columbia University, the Clinton
Laboratories and other laboratories studied the behavior of slurries,
difficulties of using them, and reactor concepts that could be based on
them, This research continued until near the end of 1944, when most of

it ended. Some continued, however, especially at the Clinton Laboratories.

The early work resulted in some reactor concepts, and provided
valuable information for later work under the Homogeneous Reactor
Project. One of the earlier concepts was for a homogeneous production
and power pile using a slurry of uranium oxide in heavy water.!30 The
slurry is pumped through pipes to heat exchangers or flash evaporators, then
back to the pile.

At Columbia, a general concept for a slurry reactor was reported in
1944,131 A slurry of slightly enriched (uranium=233) uranium trioxide is
suspended in heavy water and, presumably, cooled by circulation to
external heat exchangers. The power was estimated at 600 Mw. Other

concepts from the Manhattan Project included the Homogeneous Slurry Pileld2

66

and the Heterogeneous Slurry Pile.l33 Most of the other early slurry
concepts either closely resemble these concepts or have been incorporated

into later designs.

Slurry Development in the HRE Project

Before and during the Homogeneous Reactor Project, extensive
development was carried out at Oak Ridge on slurries and slurry reactors.
Such problems as the preparation of suitable slurries, prevention of
caking and settling, and determination of optimum particle sizes were
investigated, and designs for slurry reactors were formulated. Much of
this work is reported in the Quarterly Progress Reports of the Homogeneous
Reactor Project, particularly some of those for 1950-1952,134=-136

A reactor originally proposed in 1943, the Natural-Uranium-Plutonium
Producer, was considered by the Homogeneous Reactor Project in 1949.°20
In this one-region converter, the fuel is a slurry of natural uranium
oxide in heavy water. The alternative fuel considered was a solution of
uranyl fluoride in heavy water. For this fuel, the reactor would be
operated at high pressure., This concept was considered to offer the
possibility of being the most economical plutonium producer.

A slurry reactor was conceived as an alternative to the solution
breeder for central-station power described under '"Two-Region Converters
and Breeders."!l! It is a single-region breeder, with a slurry of
thorium dioxide and uranium trioxide in heavy water serving as fuel,
coolant, moderator, and fertile material. The power is rated at 100 Mw(e).

A one-region uranium-238 converter and power reactor was one of the
reactors evaluated in 1956 for possible large-scale development.®? 1In
this reactor, a slurry of uranium trioxide and plutonium dioxide in heavy
water is the fuel. It is designed for a conversion ratio of 0.88 and
production of 440 Mw(t), or 100 Mw(e) from the generating plant.

In 1956, nuclear characteristics were computed for more than 400
two-region spherical reactors having a slurry of thorium dioxide and
uranium-233 dioxide in the core and blanket.l37

After the solution-fueled reactors in the HRE sequency, HRE-4 was
designed as a slurry-fueled reactor.l3%:13% The purpose was to study
such aspects as criticality, effects of radiation on slurries, and
operational problems of a slurry-fueled reactor. It is a one-region

reactor fueled with a slurry of thorium dioxide and enriched uranium

67
dioxide in heavy water. The system is pressurized by boiling slurry in
the dump tank. The original design power level was 5 Mw(t). It was
planned to operate HRE-4 later with higher concentrations of slurry in
the blanket, concentrations approaching those in the blanket of a two-
region breeder. A later report!“0 refers to HRE-4 as a 100~kw critical
system, with a moderator of 80 percent heavy and 20 percent light water,
A reactor concept developed late in the Homogeneous Reactor Project
was a power breeder with a slurry core and blanket.®? The fuel is a
slurry of thorium dioxide and enriched uranium oxide in heavy water. The
fertile blanket is a slurry of thorium oxide in heavy water. This reactor was

designed for high power--1140 Mw(t) or 333 Mw(e).

ORSORT Concepts

Students at the Oak Ridge School of Reactor Technology designed slurry
reactors for generating electrical power. In the reactor for electrical
power in a stationary plant, the '"Homogeneous Power Producer,"!%1l the core
is a stainless-steel sphere, with concentric shells. It is cooled by
water circulating between the inner and outer shells. The fuel-moderator-
coolant-fertile material is a slurry of natural uranium in heavy
water. The plutonium produced remains as supplementary fuel. In another
slurry reactor for power, natural uranium trioxide, as a slurry in heavy
water, is the fuel.l“*2? Two reactors are combined to produce a total power

of 1000 Mw(t).

The PAR Reactor

The Dow Chemical Company and Detroit Edison worked on developing a
slurry reactor for power.l"3 It would be fueled with a slurry of uranium
oxide or oxyfluoride in heavy or light water, and it would have a graphite
core structure. It would be workable, but the need for expensive low-
temperature, low-pressure steam installations resulted in the abandonment
of this reactor.

The Pennsylvania Advanced Reactor (PAR) has been referred to as the
second, and last attempt in the U. S. to use the concept of homogeneous
aqueous fuels industrially.l“® The first was considered to be the Wolverine
solution reactor. Thus the authors disregarded the Dow-Detroit Edison
concept.

The PAR, also known as the Pennsylvania State Advanced Reactor, was a

68
one-region aqueous slurry type using uranium oxide fuel and thorium oxide
fertile material, with very fine particles—-under 5 u--suspended in heavy
water, which acts as both moderator and coolant. The slurry is pumped
through the spherical reactor vessel. It then goes to heat exchangers
for steam generation. The power for a reference design is 550 thermal

megawatts. ">

The PAR Project originated in November 1954, when Pennsylvania
Power and Light Company and Westinghouse Electric Corporation joined
to survey feasibility of various reactor types for power generation.

The PAR Project, formally set up in August 1955, was part of the
Westinghouse commercial atomic power activities in Pittsburgh.!“® Shortly
after the start of the Project, Union Carbide Nuclear Company joined the
group to study the chemical-reprocessing problems involved.!"’

After two and a half years of the program, the AEC indicated
willingness to support the research and development leading to the design
of a plantyi"*8 in 1958, Pennsylvania Power and Light decided to abandon
the project, but decided to develop a final reference design. A small
prototype had been judged necessary before the final industrial design,
but it was abandoned because it was considered too expensive,l“?

A major effort on the project was directed toward finding solutions
to problems associated with transferring and circulating aqueous
slurries of thorium oxide at the concentrations and conditions of a
single-region homogeneous reactor.1%0 The world's largest test loop
for homogeneous reactors, capable of circulating 4000 gpm at the same
temperatures and pressures planned for the power plant, was used in

pumping slurry for 35,000 hours.

KEMA Reactors

The research laboratory of N. V. KEMA, the Netherlands, has instituted
a program to develop a suspension-fueled power reactor. The three steps
planned are: a subcritical assembly, a small-scale process reactor, and
a power-demonstration reactor.!>! To date, only the first two steps have
reached the concept stage.

The suspension-type reactor is known, together with its subcritical

69
assemblies, as the Aqueous Homogeneous Suspension Reactor, Arnhem Suspension
Reactor, BABYPOP, KSRT or KEMA Suspension Test Reactor, Subcritical
Homogeneous Suspension Reactor, Suspension Test Reactor, SUSPOP, or 250
kw Homogeneous Suspension Reactor.

KEMA chose a suspension reactor, although it had been abandoned in the
U. S., for several reasons.l®2»153 The homogeneous reactor with fissile
material distributed in heavy water needs no structured core and also has
a high power extraction. Fission products can be removed from the reactor
medium in a decontamination plant. No safety or control rods are required.
Also a high conversion rate of fertile to fissile material with low fuel-
fabrication and processing costs is possible.

The initial reactors were BABYPOP, and SUSPOP. BABYPOP was a zero-

power reactor in which uranium dioxide or U_O_ is suspended in light water.

Little has been published concerning it,1213b3t it apparently has most of
the features of SUSPOP, a subcritical assembly that would become critical
later if possible. SUSPOP is fueled with an aqueous homogeneous suspension
containing 20 percent enriched uranium as uranium dioxide.151,153,154 1¢
is moderated with light water and reflected by beryllium oxide and graphite.
R. C. N. (Reactor Centrum Nederland) owns SUSPOP, which KEMA has operated
since 1955 to investigate the flow behavior of a circulating suspension
reactor.

The second stage, or process reactor, is also known as the KEMA
Suspension Test Reactor. It has been designed as a one-region reactor
fueled with a suspension of 20 percent enriched uranium dioxide suspended
in light water.!®%"156 It is pressurized with hydrogen to reduce oxidation,
water decomposition, and stress corrosion. The design power is 250 kw.

This reactor is intended for investigating, under power-reactor conditions,
chemical and mechanical properties of the suspension, water decomposition,
and special purification systems that had been suggested. The power
density, temperatures, and suspension concentration are meant to equal
those in a normal power reactor. This reactor was originally scheduled

for construction after the zero-energy reactor had operated satisfactorily.

The third stage is intended to be a simplified power demonstration
reactor in which important reactor components--vessels, valves, pumps,

etc.—-would be investigated, as would the possibility of remote maintenance

of such an installation.

70

Other reactors were considered briefly.157 A heterogeneous suspension
reactor might use natural uranium if it were heterogeneous both geometrically
and thermally. A very concentrated suspension would flow through tubes
inside the reactor tank filled with heavy water. Just after leaving the
reactor the concentrated suspension would be diluted slightly so it could
easily flow through the external heat exchanger. A homogeneous reactor
comparable to SUSPOP, was also considered. It would use light water (in
dilute suspension only) as both moderator and suspension liquid for uranium
trioxide or U308°

In 1962 it was reported that work in SUSPOP was continuing, in
collaboration with teams at Harwell, Mol (Belgium), and Saclay (France) .18

Construction of the test reactor was under discussion.

Other Slurry Concepts

In addition to those concepts already discussed, other slurry-fueled
reactors have been considered in England, Canada, Czechoslovakia, and
Russia, with varying degrees of development taking place.

In developing the Homogeneous Aqueous Reactor (HAR), the British
Atomic Energy Research Establishment studied slurries and considered them
as possible fuels. In 1954, the two-region pressurized reactor, with a
solution of uranium-233 salt in heavy water and a blanket of thorium
oxide slurry, was chosen over the single-zone slurry reactor. The choice
was made because the single~region slurry and other reactors either
were less economic, required new developments, or had stability or
safety problems,!%9

Later, however, when the system again came under review, the physics
of a single-region slurry system were reassessed. The reactor is assumed
to be an unreflected sphere of 7 to 15 feet in diameter, containing a
slurry of thoria with uranium-233 in heavy water.1©0

The Atomic Energy Commission of Canada in 1953 made calculations on a
two-region, heavy-water-slurry fuel reactor to assess the benefits of a
moderate amount of plutonium enrichment.!®! They concluded that the matter
deserved further study. In one concept, the reactor is fueled with a
heavy-water slurry of natural uranium oxides enriched with plutonium, with

a fertile slurry of natural uranium oxide as a blanket. The reactor

consists of two concentric spheres, with the uranium oxide slurry between

71
the spheres, and the inner sphere containing the plutonium.

HR~1 and HR-2 are Czechoslovak reactors that are slurry-fueled. HR-1
is a 10-Mw one-region homogeneous reactor, in which the fuel is a
suspension of 15 percent enriched U308 in light water. HR-1 is an
experimental pressurized homogeneous reactor suitable for research work and
for obtaining practical manufacturing and operational experience in this
field.1%2 Experiments on the hydrodynamics of the suspension led to the
HR-2 concept.!®3 The basic concept of HR~2 is an arrangement in which the
slurry moves only in the vertical plane to eliminate precipitation of the
solid phase. The active zone, which is in the lower part of the reactor,
consists of two coaxial components. The suspension flows downward in the

outer part and upward in the inner. Work on this reactor was discontinued.l6"

Evaluation and Status

The extensive developmental programs on aqueous homogeneous reactors as
power and breeder reactors have been followed by several evaluations of
these breeders for industrial promise. Some evaluations of individual
types have been discussed previously in this chapter.

One of the earlier evaluations of these reactors was that made by
the General Electric Company in 1955. They found that these reactors
must be considered as a promising approach to expansion and development of
atomic power. 16>

In 1956, Foster Wheeler Corporation and the Pacific Northwest Power
Group reported they were considering the HRT or TBR type; Foster-~Wheeler
proposed building a two-region homogeneous thorium breeder for delivery in
1961 or 1962; the exact type was unspecified.3"

The program of the Homdgeneous Reactor Project was one of those
evaluated by the Fluid Fuel Reactors Task Force. In January 1959, the
Task Force decided to reduce work on fluid fuels and consolidate it into
one program at the Oak Ridge National Laboratory, with basic research to
be carried out for an undetermined period.!6®

This decision was made despite the more~favorable report by members
of a review and evaluation group at Argonne National Laboratory and the
opinion of Lane. The Argonne group concluded that aqueous homogeneous
reactors offer many advantages as plutonium producers, but they must be

considered as long-range possibilities because of present difficulties.1®”

72

In March 1959, Lane commented:
..""the consensus of those associated with the Homogeneous
Reactor Program is that, in spite of technical problems which
have been encountered since the inception of the program, the
potential of aqueous homogeneous systems is as great as it ever
has been. This conclusion is based on the most recent
information.168
In April 1961, HRT was permanently shut down and the fuel was processed
for recovery of uranium-235. Work on the aqueous homogeneous reactor
concept was terminated on June 30, 1961, although some research on aqueous
systems was continued under the Thorium Utilization Program, initiated at
Oak Ridge during the latter half of 1961.2% The thorium program is
related to the previous program to develop a thorium-uranium-233 breeder.!69
Another evaluation of thorium breeder reactors, made in 1961, concluded
that the Aqueous Homogeneous Breeder Reactor ranked first over the molten-
salt, liquid-bismuth, and gas-cooled types in regard to nuclear capability,
fuel-cycle potential, and status of development.®?
To date, the current status of developmental programs and the actions
by the Atomic Energy Commission and by industry indicate that the future

commercial development of aqueous homogeneous reactors—-either solution or

slurry--remains uncertain.

73
DATA SHEETS

NON-BOILING REACTORS FOR POWER AND BREEDING
No. 1 Pilot Model for HRE Series

ORNL

Reference: CF-49=7-135 Del.

Originator: A. M. Weinberg et al.

Status: Conceptual design, July 1949; predecessor of HRE series.
Details: Thermal neutrons, steady state, single-region converter, Fuel-
moderator: solution of enriched uranyl sulfate in HZO: 2 kg UOZSO4 per
50 liters; 40 g U235 per liter. Moderator-reflector: DZO in 15=-cm
blanket around reactor. TFuel solution, at 482°F and 1000 psi, is pumped
downward from the reactor through two vertical heat exchangers and cooled
to 437°F by heat exchange with boiling water at 400°F and 250 psig.

Steam is produced at 600 1lb per hour. The D,0 moderator is cooled by

2
coils immersed in retention tank for DZO' Reactor wvessel: cylindrical
tank, 40 cm diameter, 40 cm high. 50 liters volume. Controls: fine
control by vertical regulating rods in enclosed thimbles around reactor

vessel; raising and lowering level of D O blanket is control (without

blanket, reactor is non-critical). Forzrapid shutdown, blanket and fuel
can be emptied into dump tanks. Power: 20 kw(e).

Code: 0311 14 31201 43 624 744 8111X 921 101

82188

83189

77
No. 2 HRE-1, Homogeneous Reactor Experiment

ORNL

References: ORNL-527; TID-10082; ORNL-730; J. Brit. Nuc. Eng. Soc., 1,
p. 35, January 1962; IAEA Directory of Nuclear Reactors, 2, p. 153; Lane,

et al, Fluid Fuel Reactors, pp. 348-59.

Originator: ORNL staff.

Status: Critical April 1952; operating at full power spring of 1954.
Dismantled.
Details: Thermal neutrons, steady state, burner. Fuel-moderator: solution
of 93.47% enriched U235 as uranyl sulfate in light water circulating through
closed circuit, spiral path in core, at 100 gpm. Coolant: circulating
fuel. Other cooling: water-spray on reflector pressure shell, natural
convection water system on solution and reflector dump tanks. Operating
temperature in core: 250°C; pressure of liquids in both core and
reflector: 1000 psi. Reflector: DZO pressurized by helium, contained
in 39 in. ID forged steel pressure vessel surrounding core, and cooled by
recirculation through a heat exchanger. Stainless-~steel spherical core,
18 in, ID, 50 liters capacity. Control: changing concentration of
solution: 2 shim and safety rods and 1 regulating rod (boral plates clad
with stainless-steel and bent to form segments of vertical cylinder around
core) in reflector zone tangent to core; and changing reflector level.
MTR-type operating mechanism. Pilot-plant power level: 200-1000 kw.
HRE-1 demonstrated the stability of a circulating fuel reactor operating
at high power densities. It was planned to substitute certain parts,
i.e., reactor vessel of Zr instead of stainless steel, later in the
experiment. Consequently, unmortared concrete blocks were used for
shielding. The substitutions later were incorporated in HRE-2,
Lode: 0311 13 31201 44 624 711 83779 921 101

81141

82188

78

No. 3 HRE-2, Homogeneous Reactor Experiment-2

(HRT, Homogeneous Reactor Test) (HRE~-2/HRT)
ORNL

References: ORNL-2096; ORNL-2148; ORNL-2222; ORNL-2272; ORNL-2379; ORNL-3167;
Proe. 2nd U.N. Int. Conf., 9, pp. 509-527; Lane, et al, Fluid Fuel Reactors,

pp. 359-97; Proc. 1st U.N. Int. Conf., 3, pp. 263-82.

Originator: ORNL staff.

Status: Critical December 1957; permanently shut down, 1961.

Details: Occupied same building that housed HRE-1l. Thermal neutrons,
steady state, burner. Fuel-moderator: 937 enriched U235 as UOZSOA-CuSO4—
HZSO4 in D20. CuSO4 added to supress radiolytic decomposition of water.
Coolant: circulating fuel solution. Reflector: D20, 13,69 in. thick
between core tank and stainless-steel clad carbon steel pressure (2000
psi) vessel (60 in. ID); cooled by circulation through heat exchanger.
Average temperature for both fuel solution and moderator: 280°C;
pressure: 1700 psi. Core tank: inverted-pear shape (top section 32 in.
ID, spherical) of Zircaloy-2, 290 liters volume. Designed as two-region
reactor, but hole in core tank let fuel into reflector. Consequently run
as one-region. Control: no control rods; ratio of heat withdrawal from
reactor (power demand at turbine) automatically met by negative temperature
coefficient (~0.3% Ak/°C at 280°C). Normal control by dilution of fuel
with DZO' Power in core: 5 Mw(t). HRE-2 was built to answer questions
of reliability, performance, and maintenance in long-term operation.
Construction completed in April 1956. From then until December 1957,
non-nuclear tests were performed.

Code: 0311 14 31202 44 625 711 84677 921 101

83779

79
No. 4 HRE-2 Replacement Reactor

ORNL

Reference: CF-61-7-54; CF-59-8-110.

Originator: R. H. Chapman.

Status: Reference design report as a result of termination by AEC of HRP,
Details: Thermal neutrons, steady state, breeder (when blanket is used as
described here). Fuel-moderator-coolant: solution of U235 (presumably
highly enriched uranyl sulfate) in D20. Reflector: 4-in., thick Be
surrounding Zircaloy-2 cylindrical core tank (21 in, ID x 42 in. long).
Fuel enters tangentially through two "slotted-entry" headers on the
cylindrical portion of the tank with sufficient velocity to provide

proper cooling of the wall. At each end of the headers, the slots

direct some flow into the hemispherical ends; the major portion leaves
through an outlet on the top of the tank. An 8-in.-thick ThO2 pellet
blanket surrounds the Be. Fuel and blanket temperature: 280°C. Breeding
ratio: estimated at 1.07 for 6 in. blanket. Control: (probably) by
negative temperature coefficient. Power: 5-10 Mw. This design is for

an experimental reactor suitable for installation in the existing HRE-2
facility with a minimum of system changes.

Code: 0312 14 31202 44 625 786 84677 941 101

80
No. 5 Intermediate-Scale Homogeneous Reactor (ISHR),

Single Region Design

ORNL

References: ORNL-1280, pp. 113-33; ORNL-1318, pp. 95-111; ORNL-1424,
pp. 43=-53; Lane, et al, Fluid Fuel Reactors, pp. 8, 504-5; CF-52-8-31,

Originators: Homogeneous Reactor Project staff.

Status: Design studies commenced October 1952; further work deferred

late in 1953 because "construction...[of HRE-2]...seemed a more suitable
course of action." HRP cancelled 1961.

Details: Thermal neutrons, steady state, converter. Fuel-moderator:

2.5% enriched U235 as uranyl sulfate in D20. Solution temperature: 250°C.,
Both vortex and axial flow studied as possibilities. Core vessel:
spherical stainless steel, 6 ft in diameter. Maximum pressure: 1000 psia.
No reflector other than the steel core shell. Cooling water circulates
through three 3/4-in.-wide passages between three 3/4—in. thick steel
shells, all of which are enclosed by a 2-1/2-in.-thick steel outer shell.
Main circulating system: reactor vessel, external contrifugal gas
separator, shell-and-tube heat exchanger (vertical, forced-circulation
boiler), and circulating pump. Control: assumed through varying
concentration of solution and negative temperature coefficient. Reactor
designed as a power and plutonium producer. Power: 48 Mw.

Code: 0311 14 31202 42 625 743 84677 921 101
83779

81
No. 6 Cooling Tower Reactor

ORNL

Reference: CF-53-10-185.

Originators: J. P. Sanders and P. N. Haubenreich.

Status: Preliminary design, 1953; no further work.
Details: Thermal neutrons, steady state, burner, converter, or breeder.
Based on the ISHR Single-Region design, the Cooling Tower Reactor differs
mainly in the method of removing heat from the solution. Fuel-moderator:
U02804 solution in D20 at 700 psia. Solution enters the packed or slotted
cooling tower at 482°F (250°C as in ISHR) and leaves at 416°F. A portion
of the D20 is evaporated into a recirculating gas stream, which is then
cooled. Part of the vapor is condensed in a heat exchanger producing
steam. The tower bed and the core are located in a horizontal pressure
vessel 10 ft in diameter and 30 ft long. The reactor may be used as a
breeder, or, by substituting HZO for D20, as a burner to produce power,
Code: 0311 14 31202 42 625 743 84677 921 101

0312 13 31201 44 624 711

0313

82

No. 7 Intermediate Scale Homogeneous Reactor (ISHR)

Two-Region Converter

ORNL

References: ORNL-1424, pp. 53-7; ORNL-1478, pp. 23-40; ORNL-1605,
pp. 31-46.
Originators: HRP staff.

Status: Design studies, 1952-53; emphasis shifted in October 1952 from

one— to two-region design. ISHR deferred in £fall 1953 for work on HRE-2,

HRP cancelled, 1961.

Details: Thermal neutrons, steady state, converter with blanket conversion.
235

Fuel-moderator: 93.5% enriched U in dilute DZO solution of uranyl

sulfate, Fertile material: breeding blanket, 2 ft thick, of ThOZ-DZO
slurry. Other blanket compositions considered were: thorium for a
breeder, lithium compound to produce tritium, and uranium to produce
plutonium. Both solution and slurry are under 1000 psia pressure at
250°C. A modification of the ISHR Single-Region design, the core vessel
of the ISHR Two-Region Converter is a sphere, 4 ft diameter, of stainless
steel. Fuel enters at the bottom, where it is heated to 482°F, and
leaves at the top to return to the two-pass, horizontal, shell-and-tube
heat exchanger for cooling. An arrangement of screens in the core tank
and the small-diameter holes through the pressure shell from the breeder
material inlet and outlet headers provide efficient flow of fluids
through the wvessel. Cooling by thermal convection in the high-pressure
recirculating system makes fuel dumping unnecessary. Control: control
rod designs were considered but discarded on basis of self-regulation by

temperature coefficient. Power: core-48 Mw; blanket-9.6 Mw.

Code: 0311 14 31202 44 625 756 84677 941 101

83
No, 8 One—-Region Breeder ISHR Design

ORNL

Reference: ORNL-1605, pp. 46-7.
Originator: HRP staff.

Status: Calculations completed, 1953; discarded.

Details: Thermal neutrons, steady state, breeder. Fuel: U233.
Moderator-coolant: D,0. Fertile material: Th232. They are mixed

2
together in single core 10, 15, or 20 ft in diameter. No reflector.

Constants are the same as for two-region reactors. Power: between
500-2000 Mw.
Code: 0312 14 31X02 45 6X5 7X6 84677 91X 101

No. 9 Two-Region Breeder, ISHR Design

ORNL

Reference: ORNL-1605, pp. 44-5.
Originator: HRP staff.

Status: Calculations completed for various core and blanket sizes,
1953; discarded.
Details: Thermal neutrons, steady state, breeder. Fuel-moderator:
same as for the ISHR Two-Region Converter. Fertile material: blanket
containing ThO2 pellets or slurry. Core diameter: 6 ft. Slurry blanket
would be 2 ft thick and the pellet blanket 1 ft thick, although the latter
may have to be larger. Other details the same as for the converter.
Power (core): 320 Mw.
Code: 0312 14 31202 44 625 756 84677 941 101

786

No. 10 Intermediate Scale Lithium Converter Three-Region Reactor for

Production of Tritium

ORNL

Reference: ORNL-1478, p. 40-1.

Originator: HRP staff.

Status: Conversion calculations completed, 1953; discarded.

Details: Same as ISHR Two-Region Converter Design except that core tank is
surrounded by a 3-cm layer of molten Li as a blanket which, in turn, is
surrounded by a BeO reflector.

Code: 0311 14 31202 44 625 738 84677 941 101

84

No. 11 Two-Region Thorium Breeder Reactor (TBR) (K-23 Reactor)

ORNL

References: TID-7524, pp. 65-77; Lane et al, Fluid Fuel Reactors, pp. 507-9;
Proc. 1st U.N. Int. Conf., 3, pp. 175-187; ORNL-1761.
Originator: HRP staff.

Status: Conceptual design completed, 1955 (unpublished); set aside because
of urgent work needed for HRT; no further work.

Details: Thermal neutrons, steady state, breeder. Fuel-moderator:
solution of U0,50, (mainly 0233 wien U234, 0?32, and 1?3 in D,0. Fuel
flows into core through center pipe (at 250°C) and leaves through the
annulus at the top (at 300°C). Fertile material: blanket, or breeding
region (27 in. thick) containing slurry of ThO2 in D20, which enters at the
bottom of blanket vessel and exits at the top. Breeding ratio is 1.11
(gross). Pressure (core solution and blanket slurry): 2000 psi. Core
vessel: sphere, 5 ft diameter (6-ft-diameter core was first considered),
of Zircaloy-2, and blanket pressure vessel, 10.5 ft diameter, of
stainless-steel-clad carbon steel. Vertical steel envelope encloses all
parts of reactor plant that handle radiocactive fluids under pressure.

HZO flows through heat exchangers built into core and blanket circulating
pumps to remove excess heat. Main heat exchangers: core and blanket
steam generators. Control: high negative temperature coefficient. Three
reactors comprise proposed plant; each produces 100 Mw(e)--440 Mw(t)--for
plant total of 300 Mw(e) net.

Code: 0312 14 31202 45 625 756 84677 941 101

85
No. 12 HRE-3 Homogeneous Reactor Experiment-3

ORNL

References: CF-58-11-112; Lane, et al, Fluid Fuel Reactors, pp. 509-11,

Originator: HRP staff.

Status: Design of HRE-3 as a two-region power breeder reactor was suspended
pending resolution of problems arising in operation of the HRT. Preliminary
design was summarized, November 1958, Design studies discontinued, 1958,

to investigate slurry-handling problems. Homogeneous Reactor Project
cancelled, 1961.

Details: Thermal neutrons, steady state, two-region, thorium breeder.

U233

Fuel-moderator-coolant: as uranyl sulfate solution in D,0. Fertile

2
material: 2-ft blanket of slurry, ThO, in D,.0. Pressure of core and

blanket: 1500 psi; average temperaturi of cire: 280°C, Oxygen is favored
as pressurizing medium for core and blanket. Core tank: Zr sphere,

4-ft diameter inside pressure vessel, 9-ft diameter, although a cylindrical
vessel 3-ft diameter and 6-ft high considered. Sphere volume: about
950 liters. Pressure vessel: carbon steel clad with stainless-steel.
Control: not described, assumed to be self-regulation by reactivity
temperature coefficient. Power: 50 Mw(t) in core; 10 Mw(t) in blanket.

Code: 0312 14 31202 45 625 756 84677 94 101

86

No. 13 U235 Burner

ORNL

Reference: CF-56-1-26, Del., p. 9.

Originator: HRP staff.

Status: One of five reactors evaluated by HRP for possible large-scale
development in January 1956; considered most feasible. Project terminated
1961,

Details: Thermal neutrons, steady state, one-region burner. Fuel-

moderator—-coolant: wuranyl sulfate solution in D,0, with CuSO, added;

1.8 g uranium per liter; 0.54 g fissile materialzper liter. ;aximum
fuel temperature: 300°C. Reactor pressure: 1800 psia. Total system
volume: 32,000 liters. Reactor diameter: 10 ft. Power: 440 Mw(t);
100 Mw(e) from generating plant.

Code: 0313 14 31202 43 625 711 84677 91 101

SXXXX

No. 14 One-Region U238—Pu Converter and Power Reactor

ORNL

Reference: CF-56-1-26, Del., p. 9.

Originator: HRP staff.

Status: Ome of five reactors evaluated for possible large-scale
development, January 1956. Project terminated 1961.

Details: Thermal neutrons, steady state, one-region converter. Fuel-
moderator-coolant: wuranyl sulfate solution in D20 with CuSO4 added;
300 g uranium per liter; 4.1 g fissile material per liter. Maximum
fuel temperature: 250°C. Reactor pressure: 1000 psia. Total system
volume: 46,000 liters. Reactor diameter: 12 ft. Conversion ratio:
0.81. Power: 480 Mw(t); 100 Mw(e) from generating plant.

Code: 0311 14 31202 42 625 743 84677 91X 101

87

33

No. 15 One-Region Th.-U2 Breeder

ORNL

Reference: CF-56-1-26, Del., p. 9.
Originator: HRP staff.

Status: One of five reactors evaluated for peossible large-scale
development, January 1956. Project terminated, 1961l.

Details: Thermal neutrons, steady state, one-region breeder. Fuel-
moderator-fertile material: slurry of ThO2 and UO3 in DZO; 300 g uranium
per liter; 6.2 g fissile material per liter. Maximum fuel temperature:
300°C. Reactor pressure: 1800 psia. Total system volume: 46,000
liters. Reactor diameter: 12 ft. Conversion ratio: 1.0+. Power:

440 Mw(t); 100 Mw(e) from generating plant.

Code: 0312 14 31302 45 635 753  B4677 91X 101

756

No. 16 Two-Region Th~U233 Breeder

ORNL

Reference: CF-56-1-26, Del., p. 9.
Originator: HRP staff.

Status: One of five reactors evaluated for possible large-scale
development, January 1956. Project terminated, 1961.

Details: Thermal neutrons, steady state, two-region breeder. Fuel-
0, with CuSO

moderator-coolant: wuranyl sulfate solution in D in core;

6.3 g uranium per liter; 2.4 g fissile material fier liter in iore, plus

3 g per liter in blanket. Fertile material: slurry blanket of ThO2 and
UO3 in D,0; 1000 g per liter. Maximum fuel temperature: 300°C. Reactor
pressure: 1800 psia. Total system volume: 27,900 liters. Reactor
diameter: 9 ft, with 5 ft diameter core. Conversion ratio: 1.1, Power:
440 Mw(t); 100 Mw(e) from generating plant.

Code: 0312 14 31202 45 625 754 84677 941 101

756

88

No. 17 400-Mw Thorium Breeder with Pellet Blanket

ORNL

Reference: Unpublished report, ORNL, Jan. 4, 1960,

Originators: I. Spiewak and F. N. Peebles.

Status: Conceptual design for evaluation, Jan. 4, 196Q; project terminated,
1961.
Details: Thermal neutrons, steady state, two-region breeder. Fuel-

233

moderator: UO,SO, scelution in D,0; 5 g total U per liter; 1.55 g U

plus U2 per iitir. Fertile maierial: 2 ft blanket of 1/8-in. thorium
oxide pellets, 1450 g Th per liter. Fuel solution enters at top of core,
at 240°C and 26,600 gpm, through swirl-producing vanes. Fluid blankets
the whole inner surface of the reactor tank. Fuel leaves at 29Q0°C through
top of reactor to go to external heat exchangers. Th pellets are in three
annular rings, which are cooled by flowing DZO' Core pressure is 20 psi
above blanket pressure. Core vessel: Zircaloy-2 cylinder, 3/8 in. wall
thickness, 4 ft diameter, 12 ft long. Control: negative temperature
coefficient of solution and emergency dumping. Power: 464 Mw(t),

135 Mw(e). Similar design for 40 Mw test reactor.

Code: 0312 14 31202 45 625 786 84677 941 101

83189

89
No. 18 Solution-Core, Slurry-Blanket Power Breeder

ORNL

Reference: CF-59-7-129, pp. 11-18.

Originators: HRP staff.

Status: Conceptual design for a large breeder plant, 1959; HRP cancelled,
1961.

Details: Thermal neutrons, steady state, two-region breeder. Fuel-
moderator: enriched uranyl sulfate solution in DZO’ 5 g uranium per liter
(1.9 g U233, 1.4 g u?3%, 0.2 g %30, 1.4 g U239y, Cus0, , NiSO,, and D,SO,
added. Fertile material: 2 ft thick blanket of ThO2 in DZO slurry, with
U233 and Pa?33. Slurry: 1000 g Th per liter (11 vol. % solids); 2.5-4.5 g
U233 plus Pa233 per kg thorium. Fuel solution enters core tank at 250°C
and 33,000 gpm. It leaves at 290°C and divides into two parallel

circuits, goes to steam generators, and treatment equipment, and returns

to core, Core pressurized by boiling D,0 in surge chamber. Slurry

circulates at 15,000 gpm, enters blankei at 240°C, and leaves at 250°C.
Pumped to heat exchangers to heat feedwater for generators. Reactor
vessel: zirconium-alloy cylinder, 4 ft diameter, 12 ft long. Control:
regulating release of steam to generator and negative temperature
coefficient. Power: 380 Mw(t). Three reactors combined to furnish
steam to one 333 Mw(e) generator.

Code: Q312 14 31202 47 625 756 84677 941 101

31302

90
No. 19 Two-Region Homogeneous Solution Breeder

ORNL

Reference: ORNL-2920, pp. 27-36.

Originators: HRP staff.

Status: Preliminary mechanical design, January 1960; project cancelled,
1961.

Details: Thermal neutrons, steady state, breeder. Fuel-moderator:
designated as a sclution but composition not given. Fertile material:
slurry blanket; composition not given. Core: annular-inlet, polar-
outlet type with swirling flow. Incoming fuel is evenly distributed in
the annular inlet by a double-volute distributor, Blanket, a modified
polar~inlet, annular-outlet type, provided with a single-volute discharge
collector to distribute the flow evenly in the outlet. Core vessel:

a4 ft x 12 ft cylinder; core walls cooled to keep the inner surface

below v 260°C by circulating a cooled stream of DO through the annular

gap between double core walls. A 6 iIn. shroud suirounding the core provides
passage for upflowing slurry. Total blanket width: 2 ft. Coolant flow:
300 gpm; inlet temperature: 183°C; coolant outlet temperature: the same

as inlet fuel temperature, 250°C. Power: 380 Mw(t).

Code: 0312 14 31202 4X 62X 75X 84677 941 101

91
No. 20 Aqueous Homogeneous Breeder Reactor (AHBR)

ORNL

Reference: CF-61-3-9, pp. 51-67a.
Originators: L. G. Alexander, W. L. Carter, R. H. Chapman, B. W. Kinyon,
J. W, Miller, and R. Van Winkle.

Status: Typical concept used for evaluating purposes, 1961. Authors

judged AABR first in regard to nuclear capability, fuel cycle potential, and
status of development; however, homogeneous program cancelled by AEC in 1961.
Details: Thermal neutrons, steady state, breeder. Fuel-moderator-coolant:
solution of UOZSO4 in D20 (no enrichment or isotope given). Fertile
material: blanket 2 ft thick of 3 equal-volume annuli of ThO2 pellets and
4 annuli of D20. Fuel solution enters core at 482°F through annular

nozzle fitted with vanes to give it a swirling motion, which stabilizes the
flow and prevents boundary-layer separation as the solution flows down
along the core-vessel wall. Fuel returns along the axis of the core and
exits at 554°F through a central nozzle leading to the pump, which
discharges into two heat exchangers in parallel. Fuel returns to the

core. Core vessel: 4.4 ft diameter x 14.6 ft long vertical cylinder of
Zircaloy plate strengthened by ribs and stiffeners. Blanket pellets
retained between Zircaloy-2 cylinders, concentric with the core vessel

and perforated by small holes to a porosity of 50%. DZO coolant flows

up from the bottom of the vessel, along the core vessel wall, and then
"percolates" through the 3 layers of pellets. Each layer is divided into
33 sections ("baskets") by Zircaloy plates. Blanket D,0 flows through
pumps and then heat exchangers, from which it returns to the blanket

inlet. Maximum allowable velocity for both fluids: 20 fps. Operating
pressure: 2000 psi. Reflector: D20, 5.5 em thick, surrounding blanket
and core. Control: assumed to be by negative temperature coefficient.

A station of four reactors would produce 3640 Mw(t) or 1000 Mw(e)=--910
Mw(t) each.

Code: 0312 14 31202 4X 625 786 84677 941 101

92

No. 21 LAPRE-I, Los Alamos Power Reactor Experiment I

LASL

References: LAMS-1611; Lane, et al, Fluid Fuel Reactors, pp. 397-405;
Proc. lst U. N. Int. Conf., 3, pp. 283-6; IAEA Directory of Nuclear

Reactors, 2, pp. 159-64.
Originators: Originally designated DIR-P, project under Director's

Office made up of two groups headed by R. P. Hammond and L. D. P. King.
Status: Critical February 1956; dismantled, January 1957.

Details: Thermal neutrons, steady state, burner. Fuel-moderator:
solution of UO3 (93.4% enriched in U235) in H3PO4 (acid concentration 50%
by weight in HZO)' Coolant: HZO pumped through coiled, gold-clad,
stainless-steel tubing. Reflector: 3 in. thick steel wall of vessel
and 4 ft water shield surrounding reactor. Fuel solution temperature:
430°C; pressure 3600 psi. Critical zone of core: cylinder 15 in.
diameter and 16 in. high of gold-clad stainless steel. Core: test

tube shaped; hollow boron cylinder divides non-critical lower portion of
vessel into inner and outer section. Lower portion a storage for excess
fuel, upper the critical region. Heat exchanger: between critical zone
and storage region. Blanket gas: oxygen. Control: 1 shim safety rod
and 4 regulating rods of platinum-clad stainless-steel thimbles filled
with boron disks; vertical movement. Shim rod extends to bottom of
vessel, regulating rods extend through reactor core region only. Power:
2 Mw(t). Reactor housed in cell at LASL which had been built to handle
highly radiocactive materials.

Code: 0313 13 31201 44 624 711 81111 921 105

110

93
No., 22 LAPRE-II, Los Alamos Power Reactor Experiment II

LASL

References: LA-2465; Lane, et al, Fluid Fuel Reactors, pp. 398-405.

Originators: Evidently same group that worked on LAPRE-I,

Status: Critical experiment scheduled February 1959;

dismantled 1959.

Details: Simplified convection version of LAPRE-I. Thermal neutrons,
235

steady state, burner. Fuel-moderator: 95% enriched U O2 dissolved in

95% phosphoric acid—HZO. Coolant: HZO circulated through coils above

reactor core. Core: cylinder 23.6 in. high and 14.75 in. ID. Core and

coils constructed same as those of LAPRE-I., Solution circulates by

natural convection over coils. Solution temperature: 800°F; pressure: about
800 psi. Hydrogen over-pressure: 200 psi, Reflector: 2 graphite

concentric cylinders (sleeves). Shim control: BeO and graphite sleeve;

no other controls. Power: 0.8-1 Mw(t). Reactor located in underground

steel tank, 20 ft deep, 42 in. diameter.

Code: 0313 13 31201 44 624 711 82148 921 105

110 84677

No. 23 LAPRE~II Type Reactor for DEW Line Application

LASL

Reference: Unpublished internal report, LASL.

Originators: L. D. P. King, R. P. Hammond, G. I. Bell, P. J. Bendt,
R. A. Clark, B. J. Melton, R, E. Peterson, and E. O. Swickard.

Status: Proposal, June 1955.
Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:
homogeneous solution of 93.5% enriched UO2 in phosphoric acid with some HZO;
operated with 200 psi hydrogen overpressure. Secondary coolant: H20o
Core: sphere 24 in, high, of stainless steel, silver, and gold. Reflector:
graphite, 11 in., thick, in the form of a vertically-movable sleeve for shim
control; is thicker than that for LAPRE-II. Other modifications of
LAPRE-II include increased power (to 0.9-1.34 Mw) and insulated graphite
sleeve with 6 in. foam glass. No control rods.
Code: 0313 13 31201 44 624 711 82148 921 101

84677

94
No. 24 Homogeneous Phosphoric Acid Reactor (HPAR)

SANDIA Corp.

Reference: SC-4459 (RR).
Originatc~st Sandia Corp., Div., 5433 staff.

Status: Design study, August 1960,
Details: Thermal neutrons, steady state, burner. Fuel-moderator:
solution of fully enriched UO2 in H3P04. Reactor vessel: gold-lined
stainless-steel cylindrical pressure vessel, divided by cylindrical
graphite flow-divider into reactor core portion (central) and heat
exchanger portion (outer annulus). Pressure vessel: 20.25 in. OD
(18 in. ID) and 45 in. high, contains the core, fuel circulator, and heat
exchanger. Maximum pressure: 800 psia; fuel velocity: 20 ft/sec.
Temperature at heat exchange inlet: 800°F; outlet: 795°F. Fuel is
circulated by pump at top of vessel; graphite impeller and scroll are top
parts of flow divider. Fuel solution enters at bottom of core, rises
upward through core and is forced by pump down through outer heat-
exchanger annulus. Secondary demineralized water-steam circuit of heat
exchanger consists of spiral-wound heat-transfer tubes embedded (brazed
behind gold) in walls of pressure-vessel liner. A separate superheat
circuit is provided (top eight tubes). TForty tubes comprise boiler
portion of heat exchanger. Reflector: shim reflector of BeO (3 in. thick)
and graphite (3 in. thick) completely surrounding vessel in the form of
two movable concentric cylinders; stationary graphite reflector outside
the movable one., Control: mainly through negative temperature
coefficient; shim reflector provides some control. Power level: 100-1500
kw(e). The reactor, designed for remote locations or mobile units,
weighs 4200 1bs including reflector.
Code: 0313 13 31201 44 624 711 84677 921 101

82148

95
No. 25 500 Mw Mobile Reactor Design No. 1, 93.47Z Enriched, HZO-Moderated

ORNL

Reference: CF-50-10-114 Rev,
Originatng: ORNL staff.

Status: Conceptual design, 1950. No further work.

Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:
solution of 93.4% enriched U235 as uranyl sulfate in HZO' Core: 6 ft
diameter stainless-steel lined, carbon-steel sphere. There are 6
horizontal shell-and-tube heat exchangers. Fuel solution is circulated
by canned rotor centrifugal pumps at 25,000 gpm and is introduced through
tangential inlets to the reactor core. The circular motion of the liquid
in the core drives the gas bubbles toward the vortex and reduces the
volume of vapor holdup in the core. A recombiner burns the gas to provide
heat for superheating the steam. Solution inlet temperature: 208°C;
outlet: 250°C. Reactor pressure: 2000 psia. Control: reactor is
expected to be self-stabilizing; consequently, no control rods included,
Power: dual units considered, each producing 250 Mw(t).

Code: 0313 13 31201 44 624 711 84677 921 101

No. 26 500 Mw Mobile Reactor Design No. 2, 3.5% Enriched, HZO—Moderated

ORNL

Reference: CF-50-10-114 Rev.,
Originators: ORNL staff.

Status: Conceptual design, 1950. No further work.
Details: Same general operating characteristics as Design No. 1. All

235

other details are the same except for enrichment, which is 3.5%Z U , and

the corresponding alterations in concentration of fuel, thermal flux, etc,

Code: 0311 13 31201 42 624 84677 921 101

96
No. 27 500 Mw Mobile Reactor Design No. 3, 1.5% Enriched, DZO—Moderated

ORNL

Reference: CF-50-10~114 Rev.
Originators:; ORNL staff.

Status: Conceptual design, 1950. No further work.,

Details: Same operating characteristics as Design No. 1 with the following
changes: fuel enrichment: 1.57; moderator: D20; core: 9 ft diameter
sphere. There are corresponding changes in fuel concentration, thermal flux,
power density, etc. This design is considered by the authors as the most
promising from the standpoint of initial fuel inventory and low fuel

consumption.

Code: 0311 14 31202 42 625 743 84677 921 101

No. 28 Atomic-Powered Locomotive

University of Utah

References: Report, AIF Meeting March 15-16, 1954, New York, pp. c-2 to
c-15; Nucleonies, 12, No. 3, March 1954, pp. 78-80,

Originator: L. B. Borst, with graduate students and fellow staff members.

Status: Proposal, 1954; no further work.
Details: Water-boiler type. Thermal neutrons, steady state, burner,
Fuel-moderator: solution of fully enriched uranyl sulfate in H,O.

2
Coolant: H,O0 circulated to remove heat. Reflector: 6 in. thick water.,

Core: 2 heiagonal slabs 3 ft square and 1 ft thick of stainless steel.
Adjacent to each slab is a reflector, both connected by 10,000-kw heat-
transfer tubes that penetrate core vessel. Core temperature: 455°F;
reflector: 405°F. Control: 6 control rods that enter the core at a 45°
angle; no other data given. Power: 30,000 kw(t) as steam at 250 psi to
operate 7200~hp locomotive.

Code: 0313 13 31101 44 624 711 8111X 921 105

97
No. 29 Homogeneous Reactor for Ship Propulsion

ORSORT

Reference: CF-54-8-236 Del.
Originators: P. R. Clark, W. O. Chatfield, A. E. Cox, J. M. Detwyler,
J. A. Murphy, C. W. Nestor, R. R, Roof, and D. H. Walker.

Status: Conceptual design and feasibility study; term paper, August 1954,
Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:
highly enriched uranyl sulfate in DZO' Core: 5 ft diameter sphere; with
concentric inlet and outlet; design chosen because its flow pattern is
stable under rolling shipboard conditions. Operating temperature: 300°C;
pressure: 2000 psi. All parts in contact with solution are stainless

steel. Thermal shield cooled by D,0 flowing in 1 in. annulus between

2
shield and pressure vessel. Complete internal recombination; no contrel
rods or fission-product removal. Power: 75 Mw--10% more power than needed
by the cargo ship, C-4 Mariner Class, for which it was designed.

Code: 0313 14 31202 44 625 84677 921 101

No. 30 Aqueous Homogeneous Circulating Solution Reactor

ORSORT

Reference: CF-53-10-22,
Originators: D. W, Montgomery, W. J. Dodson, F. F. Kaiser, W. K, Luckow,
and T. J. Pashos.

Status: Conceptual design; term paper, August 1953,
Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:

circulating solution in D,0 of uranyl sulfate, approximately 93.5%

enrichment of U235. Presiure: 1000 psia; temperature: 250°C. Core:

4 ft diameter sphere. Pressure vessel: cylinder. Control: regulation
of steam pressure, hence core temperature. Power: 10 Mw. '"Package"
power unit is removable, 17 ft 6 in. high and 6 ft 8 in. diameter,

Code: 0313 14 31202 44 625 711 84677 921 101

98

No. 31 STUPO, Homogeneous, Non-Boiling, Natural Convection Power Reactor

ORSORT

Reference: CF-53-8-225.

Originators: W. E. Kinney, R. Brodsky, D. Hillis, J. T. Wagner, and
T. J. Ward.

Status: Conceptual design; term paper, August 1953.

Details: Thermal neutrons, steady state burner. Name is “extrapolation"
of LOPO, HYPO, SUPO series as a result of an abortive attempt to extra-
polate Los Alamos data. Fuel-moderator: 93.4Z enriched U235 in DZO' Zr
baffle in center of core. Core diameter: 4-1/2 ft. Stainless-steel
pressure vessel (8 ft 4 in. 0OD) provides some reflection. Solution heats
in active core, rises through central pipe, enters upper sphere, passes
through heat—-exchanger tubes, and returns to core. Operating pressure:
1000 psi; temperature: 482°F. Natural convection circulation. Control:
self-controlling; negative temperature coefficient expected to maintain

level of operation. Power: 10 Mw max., 7 Mw average.

Code: 0313 14 31202 44 625 711 84677 921 101

No. 32 80 Mw Aqueous Homogeneous Burner Reactor

ORSORT

Reference: CF=-57-8-~6.

Originators: R. H. Chapman, H. L. Collins, W. J. Dollard, D. Fieno,

J. Hernandez-Fragoso, J. W. Miller, H. von Hollen, and C. V. Wheeler,
Status: Feasibility study; term paper, August 1957.

Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:
solution of 93% enriched U235 as uranyl sulfate in HZOD Core: a 5 ft
diameter sphere of stainless—-clad carbon steel with a concentric inlet
and outlet to distribute the flow. The fuel solution enters the core at
260°C and, 1750 psia, is heated to 290°C before leaving. Steam is
produced in 4 vertical heat exchangers coupled to a common horizontal-
type steam drum. Control: through negative temperature coefficient and
"automatic demand control." Design power: 20,000 kw(e) with an 80% plant
factor.

Code: 0313 13 31201 44 624 711 84677 921 101

99
No. 33 Aqueous Homogeneous Power Plant (Wolverine)

ORNL

References: CF-57-12-8 Rev., Lane, et al, Fluid Fuel Reactors, pp. 473-9.

Originators: M. I. Lundin, R. Van Winkle, and staff, Foster Wheeler

Corporation, Wolverine Electric Cooperative.

Status: Wolverine Electric Cooperative proposed joint participation with
AEC, Feb. 1, 1956. The AEC cancelled plans to negotiate contract with
Foster--Wheeler Corp. for development and construction of the reactor,

Oct. 3, 1957, The AEC then requested ORNL to review the proposed design
and to prepare a detailed cost analysis, Oct. 22, 1957. The above was
issued by ORNL to fulfill this committment. A conceptual design study was
included Dec. 11, 1957, Project was completely cancelled in May 1958 as

a result of rise in estimated costs.

Details: Thermal neutrons, steady state, single-region burner. Fuel-
moderator-coolant: solution of 90%Z enriched U235 as uranyl sulfate in

D20° Additional U235 added periodically to sustain criticality. Fuel
solution circulation, maintained by canned-motor pump, goes from core to
U-shaped, shell-and-tube heat exchanger and back to core. Outlet temperature:
300°C; inlet: 260°C. Entire primary system is pressurized to 1900 psia by
oxygen. Concentric-inlet and outlet design promotes a very high degree of
mixing in the core. Pressure vessel: 6 ft diameter stainless-steel

within which are thermal shields coocled by H20. Control: no control

rods; normal control is presumably through negative temperature coefficient.
Facilities for emergency dumping. Design power: 10,000 kw(e); 5,000 kw(e)
initially. Second superheater-turbogenerator could be added later to
increase output to design power.

Code: 0313 14 31202 44 625 711 84677 921 101

100
No. 34 Homogeneous Fuel Natural Convection Power Reactor

General Electric Company

Reference: GEAP-2 Del.

Originator: Staff, General Electric Co., Atomic Power Equipment Department.

Status: Conceptual design, May 31, 19553 no further work.

Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:
solution of U02804 in H2O under 1300 psi pressure with 93.57 enriched U235
(20-23 gm/1). Core: 26 in. ID hemisphere 13 in. high with a cylinder,
11.75 in. high, attached to the top. .Solution temperature increases from
250°C to 280°C in the core; fuel rises by natural convection in the central
riser until it reaches the top header. From there it flows down in
titanium tubes of the straight tube—and-shell-type heat exchanger that
forms the body of the reactor vessel. Heat is removed by natural
convection to boil water on the secondary side of the vessel. Pressure
vessel: Ti-lined carbon steel that is also a reflector. Control: reactor
is self-regulating. Power level: 400 kw(t); 100 kw(e).

Code: 0313 13 31201 44 624 711 84677 921 101

No. 35 80,000 Kwe Homogeneous Reactor Plant

Westinghouse Electric Corp.

Reference: WIAP-9,.

Originator: Industrial Atomic Power Group, Westinghouse (J. M. Stein,

nuclear; A. J. Surowiec, chemical processing; staff of Vitro Corp. of
America, Engng. Div.).

Status: Proposal, February 1955; no further work.

Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolant:
solution of 93.4% enriched U235 as uranyl sulfate in D20. Fuel flows from
the core to gas separators to recombiner dump tanks and back to the primary
loop of the reactor. Inlet temperature: 482°F; outlet: 572°F. System
pressure: 2000 psia. Pressure vessel: 10 ft OD, of 6 in. thick carbon
steel; contains thermal shields lined with titanium to make the core region
7 ft ID. Control: negative temperature coefficient and changing fuel
concentration. Power rating: 80,000 kw(e).

Code: 0313 14 31202 44 625 711 84677 921 101
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101
No. 36 Enriched Pile

Clinton Laboratories

Reference: Unpublished report.
Originator: L. W. Nordheim.

Status: Proposal; calculations, Nov. 16, 1944.

Details: Thermal neutrons, steady-state, breeder. Fuel-moderator: 800 g

U235 salt dissolved in D,0 to make 12.5%Z solution, or 500 gm plutonium salt

2 232 233
dissolved in 100-200 liters D.0O., Fertile material: Th with U and

U238 with plutonium. Reflectir: graphite. Solution circulated to
external heat exchanger. Temperature difference of 50°C should yield
1 Mw or more. At least 0.59 g U233 produced per day.

LCode: 0312 14 31202 45 625 7X6 84677 921 101

46 7X2 BXXXX

No. 37 Pu239 Homogeneous Power Reactor

ORNL

Reference: ORNL-527, p. 10.

Originator: C. B. Graham et al.

Status: Considered a long term objective, 1949.
Details: Thermal neutrons, steady state, breeder (could be converter),
Few details are given regarding this proposal, but they would undoubtedly
be very similar to the HRE~1l. Fuel-moderator: solution of Pu239 salt in D20°
Lode: 0312 14 31202 46 625 742 84677 921 101
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102
No. 38 Plutonium—-Producer Power Reactor

ORNL

Reference: Unpublished report.

Originator: ORNL, Long—Range Planning Group.

Status: Conceptual design, 1950.
Details: Thermal neutrons, steady-state, one-region breeder. Fuel-
O, 91 grams per

2
liter of solution. Fuel enrichment: 0,8 isotopic percent U235u Core

moderator: slightly enriched uranyl sulfate solution in D

contains 175 metric tons of DZO and 186 tons of uranium. Solution enters
reactor through tangential inlets at equator and leaves through a pipe at
the top and bottom of the shell. It then flows in high-pressure (1000 psi)
steel pipes through an 8 ft shield surrounding the reactor and into heat
exchangers in cells surrounding the shield. Inlet temperature of the
solution: 208°C; outlet: 250°C. Reactor vessel: stainless-steel sphere,
24 ft diameter, with 5-1/4 in. thick walls. Control: adjusting concentration
to make the solution just critical, negative temperature coefficient of the
solution, and emergency dumping. Power: 1000 Mw(t); 230 Mw(e).
Code: 0312 14 31202 42 625 743 83779 921 101

84677

83179

No. 39 Large Scale Homogeneous Reactor (LSHR)

ORNL

Reference: CF=52-8-7,
Originators: R, H. Ball, A, M, Hallene, and R. J. March.

Status: Economic and conceptual design study, 1952,

Details: Thermal neutrons, steady state, one-region converter, Fuel:
homogeneous solution of 1.,075% enriched uranyl sulfate, 250 g/1,

dissolved in DZOD Core contained in 16 ft ID pressure shell, 4 in., thick.
Thermal shield: three inner shells, 1-1/4 in., 1 in., and 3/4 in. thick,
respectively for thermal shield. All shells of carbon steel except inner-
most, which is Type 347 stainless steel. Core operates at 250°C and 1000 psia.
Reactor core surrounded by 5 cells, each containing 2 heat exchangers (steam

generators) and 1 turbogenerator. Working fluid in secondary system: H200
Code: 0311 14 31202 42 625 743 84677 921 101

103
No. 40 Heavy Water Moderated U233 Pile

Clinton Laboratories

Reference: CF-3199.

Originator: Laboratory staff.

Status: Preliminary design, 1945,

Details: Thermal neutrons, steady state, two-region breeder. Fuel:
solution of sodium uranyl carbonate, 4 g U233 per liter, Fertile material:
thorium metal in D,0 slurry. Solution circulates to and from heat

2
exchanger. 3 tomns D,0 and 12 kg U233 needed. Beryllium tank. Power:

2 233
100 Mw. Breeding gain per day: 13 gm U .
Code: 0312 14 31302 44 635 756 84677 92 101

15

No. 41 U235 Pilot Plant for 23 Thermal Breeder

Clinton Laboratories

Reference: CF-3199,

Originator: Laboratory staff.

Status: Preliminary design, 1945.
Details: Thermal neutrons, steady-state, two-region breeder. Fuel:

235

solution of enriched sodium uranyl carbonate in D,0, 1 g U per liter.

2
Fertile material: thorium metal. Solution circulated to heat exchanger
at 200 liters per second. Spherical core tank, 12,000-15,000 liters.
Neutron flux: 1014 neutrons/cmzlsec. Power: 10 Mw.

Code: 0312 14 31202 44 625 726 84677 931 101

104
No. 42 Heavy-Water Moderated U235 Solution Pile

Clinton Laboratories

Reference: CF-3352, pp. 2-3.

Originator: Laboratory staff,

Status: Conceptual design, October 1945.

Details: Thermal neutrons, steady state, two-region converter. Fuel:

solution of Uz35 salt (probably sodium uranyl carbonate) in DZOo
Moderator: DZO in solution and surrounding core as reflector. Fertile
material: thorium rods in row around DZO reflector. Reflector:

cylindrical graphite surrounding thorium. Solution at 50°C when it
enters reactor; at 67°C when it leaves. Control: control rod, to shut
and open pile, in thimble in center of reactor; Cd curtain in DZO
reflector also considered; fine control by rods in reflector. Conversion
yield: about 0.65.
Code: 0311 14 31202 44 624 726 84677 941 101

8111X

81142

No. 43 Thermal Breeder for Uranium—233

Clinton Laboratories

Reference: Unpublished report.

Originator: E. P. Wigner.

Status: Proposal, May 1945.

Details: Thermal neutrons, steady state, two-region breeder. Fuel-

moderator: U233 as sodium uranyl carbonate dissolved in DZO’ 8 gm
U233 per liter. Excess N32C03 hinders decomposition of salt. Fertile
material: blanket of Th02 slurry, 100 g Th per liter, in DZO' Blanket,

50 cm thick, in Pb-lined tank, is surrounded by 20 cm thick graphite
reflector. Solution and slurry pumped to and from separate heat
exchangers and through gas separators and other auxiliary equipment.,
Core vessel: 1lead-lined Al cylinder, 90 cm diameter and 90 cm high,
U233 in reactor: 4 kg; total in system: 13-16 kg. DZO in reactor:
2 tons; total in system: 13-18 tons. Breeding ratio: 1.23,

Code: 0312 14 31202 45 624 756 84677 941 101

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105
No. 44 Breeder with Circulating Uranium

Clinton Laboratories

Reference: Unpublished report.

Originators: E. P. Wigner, A. Weinberg, and G. Young.

Status: Proposal, December 1944,

Detajls: Thermal neutrons, steady state, two-region breeder. Fuel-
233

moderator: solution of U dissolved in DéO as nitrate, sulfate, or
fluoride. Using U233 solution helps eliminate Xe135. Nitrate is preferred
because of less decomposition and corrosion. About 3 kg U233 needed in
reactor. Fertile material: ThO 1 g/cm3 D,.0. Reflector: thorium and

2? 2
D20 in equal parts, 50 cm thick layer. Solution pumped into the 400-1liter

spherical reactor at the bottom, leaves at the top, passes through heat
exchangers and gas-removal systems, and returns to the reactor. Outlet
temperature: 50°C higher than inlet. Solution flow: 500 liters per
second. Power: estimated at 100-200 Mw(t).
Code: 0312 14 31202 45 625 756 84677 941 101

8XXXX

No. 45 Homogeneous Thermal Carbonate Breeder

Clinton Laboratories

Reference: Unpublished report.

Originator: E. P. Wigner.

Status: Proposal, 1945.

Details: Thermal neutrons, steady state, two-region breeder. Fuel:
sodium uranyl carbonate, about 4 g U235 per liter., Fertile material:
thorium blanket. Solution required: 300-350 liters for 5000-kw reactor.
Fuel circulates to and from heat exchanger. Spherical reactor.

Code: 0312 14 31202 44 625 756 84677 931 101

8XXXX

106

No. 46 Homogeneous Thermal Breeder

Clinton Laboratories

Reference: Nuc. Sci. & Eng., 6, No. 5, pp. 421-2, November 1959.

Originators: E. P. Wigner, W. H. Zinn, A. M. Weinberg, G. Young,

L. Szilard, and others not mentioned.

Status: Proposal developed during World War II.

Details: Thermal neutrons, steady state, breeder. Fuel-moderator:
solution of U233 as a salt, probably sulfate, in DZO, Core tank:
U-shaped tube, 100 cm in diameter, with cooling tubes that extend into
the U-tube from both sides, leaving the bottom portion of the U free for
chain reaction to take place., Fertile material: Th~D20 blanket
surrounding U-tube at the space free of cooling tubes. Pistons in both
legs of the tube move solution alternately from left to right and back so
that liquid comes in contact with cooling tubes. Solution does not fill
up the tube to the pistons; this space is filled with a gas (presumably
helium) to prevent contact of the solution and pistons. Control: design
does not include a control system but temperature control can be assumed.

Power level: none indicated,
Code: 0312 14 31202 45 625 7X6 84677 941 101

107

No. 47 U233 Power Breeder

ORSORT

Reference: CF-51-8-213.
Originators: J. Bick, J. M. LaRue, P. N. Haubenreich, D. D. Rauch, T.

Williams, and J. Putnam.

Status: Conceptual design and feasibility study; term paper Aug. 6, 1951.

Details: Thermal neutrons, steady state, breeder. Fuel-moderator-coolant:
233

solution of U as U02504 in HZO' Breeding blanket: slurry of ThF4 in

D,0, 78.23 cm thick. Core temperature: 230°C, blanket, 200°C, Pressure
of both: 1000 psi. Core vessel: =zirconium tank, 139.5 cm radius, with
two inlets and two outlets, each with several annuli. A novel feature of
this concept is that no slurry passes through the pump in the main slurry

flow line. The slurry is separated from the D,0 in a cyclone separator;

the DZO goes through a heat exchanger and thenzthrough the pump, then

DZO is recombined with the slurry and both return to the blanket.
Reflector: blanket. Control: mainly by negative temperature coefficient;
center pipe of Zr running completely through reactor is thimble in which

1 regulating and 1 safety/shim rod travel; neither is described further.
Total U233 production: 715.6 gms per day. Power: 500 Mw(t); 467 Mw(t)

in core and 33 Mw(t) in blanket.

Lode: 0312 13 31201 45 624 756 84679 941 101

8111X

108

No. 48 1Isotope and Power Producer

ORSORT

Reference: Unpublished report, ORSORT, August 1951,

Originators: W. E. Edwards, G. E. Garker, F. D. Orazio, P. 0. Nadler,

L. H. Thacker, P. N. Wood, W. E. Unbehaun, and D. T. Bray.

Status: Conceptual design; term paper, August 1951.

Details: Thermal neutrons, steady state, two-region converter. Fuel-
moderator: wuranyl sulfate solution, enriched to 1.35% U235 in DZO'
Fertile material: slightly enriched uranium solution; Al-Li alloy shot in
6=in. thick layer. Reactor solution flows through a gas-recombination
system and a main heat exchanger, to produce steam that is superheated

by heat from the recombination system. Blanket cooling: helium at an
inlet temperature of 150°F and outlet temperature of 650°F. Helium is

cooled by H,O in a heat exchanger. The primary heat—exchange system has

eight separite loops. Control: vertical concentric rods--shim and
regulating—-boron-steel sleeve; control of solution concentration;
negative temperature coefficient; and emergency dumping. Biological
shield: barytes concrete. Power: 1360 Mw(t); power density: 40,5
kw per liter.
Code: 0311 14 31202 42 625 743 8111X 931 101
788 81111

83779

84677

83179

109
No. 49 Ultimate Homogeneous Reactor

ORSORT

Reference: CF-54-8-239.
Originators: R. A. Thomas, J. F. Brunings, L. S. Hall, C. Michelson,
J. F. Parrette, L. R. Pletke, R. L. Whitelaw, and W. A. Wittkopf.

Status: Conceptual design; term paper, August 1954.

Details: Thermal neutrons, steady state, two-region breeder. Fuel-
moderator—-coolant: wuranyl sulfate in DZO at U233 concentration of about
10 g/1. Natural circulation. Fertile material: boiling DZO—Th.O2 slurry
as 75 cm—thick blanket. Core inlet temperature: 505°F. Pressure: 1850
psia. Blanket temperature: 625°F. Core heat generates steam; blanket
heat provides superheat and reheat. Core in Zircaloy shell, blanket in
stainless-steel part-cylindrical, part-spherical shell. Control:
automatic safety controls and secondary regulating controls effect shut-
down by decreasing uranium concentration. Total power: 200 Mw(t);

44 Mw(t) supplied by blanket. Core 90 cm diameter.

Alternate blanket: HZO pellet bed blanket may be substituted for
boiling DZO—ThO2 slurry. Blanket of ThOz, HZO’ steam; fluidized pebble
bed. Pellet spheres 1/16 in. containing U233 and ThOZ' Controls same
as for slurry; increased U and Th inventory to obtain same blanket power
as for slurry.

Code: 0312 14 31202 45 625 756 83779 941 101
786 84677

110

No. 50 High Temperature Homogeneous Reactor

ORSORT

Reference: CF-55-8-191.

Originators: A. Hauspurg, J. M. Finan, J. W. Geiser, B. H. Hamling,

J. J. Happell, K. Moore, and G. R. Thomas.

Status: Design and feasibility study; term paper, August 1955,

Details: Thermal neutrons, steady state, breeder. TFuel-moderator-
233

coolant: solution of U O3 and CrO3 in H20 at 3500 psia. Breeding

blanket: slurry of ThO2 in D20. Spherical zirconium core: about

5 ft ID. Blanket container: steel, 10 ft ID; blanket slurry is 2 ft
thick. Solution inlet temperature: 750°F; outlet: 1200°F. System is
essentially reflector-moderated. Control: negative temperature
coefficient. Breeding ratio: 1.003. Power: plant using two such
reactors to produce a total of 690 Mw(t), 290 Mw(e).

Code: 0312 13 31201 45 624 756 84677 931 101

111
No, 51 Two—-Region Homogeneous Without a Core Tank

ORNL

Reference: CF-54-6-180.

Originator: S. Visner.

Status: Proposal, 1954; no further work.
Details: Thermal, steady state, breeder., Fuel: uranyl sulfate solution;
moderator-coolant: not given, but probably D,O0 or H,O0. Slurry blanket:

233 2 2
U is produced. Both fuel and fertile

ThOz, also probably in water.
material are located in the same 6-ft—-diameter vessel. By use of a
rotational flow pattern, the reactor vessel is operated as a centrifugal-
type separator to locate the solution in the center and the slurry at the
periphery. However, some fuel may be with the slurry, and the density
distribution of the slurry may not be distinct. Two designs are given for
the core and flow pattern: 1) A fine slurry is recirculated from the
outlet in the wall of the cylindrical vessel. After heat is removed, it

is injected tangentically at the opposite end in a direction so as to
reinforce the rotational flow pattern. Direction could also be reversed
to top to bottom. Solution is circulated separately. 2) A coarse slurry
may be employed, for which the entire separation can be achieved near the
inlet on each pass in one second. One main circulating system is used

in either of two ways: slurry and solution could mix at the outlet,

pass through the heat exchanger together, and separate at the inlet; or

the sclution alone could be circulated and the slurry kept in the reactor.
Flow velocity for both designs (1 and 2): 20 ft/sec. No other data given.
Code: 0312 1X  312XX 4X 62X 756 84677 931 101

112
No. 52 K-=49 Reactor

ORNL

Reference: Unpublished report, March 11, 1954.

Originator: HRP staff.

Status: Calculations, March 1954; project terminated, 1961,

Details: Thermal neutrons, steady-state, two-region breeder. Fuel-

moderator: uranyl sulfate solution in DZO; 93.5% U235 in feed.
Critical concentration: 2.810 g U per kg D,0O; 1.405 g U235 per kg D, 0.

2 2
Cooling: presumably by circulation to external heat exchangers. Fertile

material: 3-ft blanket of uranyl sulfate in D_0; 342 g U per kg D, O;

2
99.757% U238. Average core temperature: 275°C. Average blanket

2

temperature: 250°C. Core vessel: zirconium sphere, 6 ft diameter,
1/2 in. thick, 3200 liters volume. Control: negative temperature
coefficient of solution. Total (core plus blanket) Pu produced per day:
433.5 g. Total power: 450 Mw(t).

Code: 0312 14 31202 44 625 745 84677 931 101

113
No. 53 Thermal Power Breeder

ORNL

Reference: Unpublished report, ORNL, 1950.

Originator: ORNL Long-Range Planning Group.

Status: Conceptual design, 1950.

Details: Thermal neutrons, steady-state, two-region breeder. Fuel-
moderator: U23302804 dissolved in D20. Fertile material: ThO2 suspended

in D20° Core solution and blanket circulated through heat exchangers,

the solution to a boiler of a steam power plant and the blanket to the

feedwater heater of a power plant. Operating temperature: core, 250°C;

blanket, 200°C. Spherical core. Three values of ratio: (cm3 Dzo/g U233)
give different reactor characteristics:
cfiSDéOQ;UZBB Spherical core Core volume, liters Power, kw/kg U233
radius, cm
500 73.4 1650 13,450
1000 91.0 3160 24,800
1500 105.0 4860 34,000

Code: 0312 14 31202 45 625 756 84677 931 101
8XXXX

114

No. 54 Project Dynamo Aqueous Homogeneous Reactor Design

Reference: MIT-5003 Del., pp. 263-74.

Originators: Project Dynamo staff (MIT) used engineering details for this

concept, which had been worked out by ORNL, and modified power level and

estimated costs.

Status: Engineering and economic analysis, 19533 no further work.

Details: Thermal neutrons, steady state, breeder. Fuel-coolant-moderator:
233

a solution of U OZSO4 in D20 o Blanket: dispersion of ThO2 in

DZO' Inlet temperature of both fluids: 216°C; outlet: 250°C. Both
are under 1000 psia pressure. Fuel flows upward through diffuser
screens in the core to a gas separator. From there it is pumped to heat
exchangers, to pumps, and back to the core. Blanket enters at the bottom,
flows upward in a 2-ft-thick annulus between the Zr core liner and the
steel pressure vessel, and then goes to heat exchangers, pumps, and back
to blanket. Core vessel: roughly spherical, 6 ft in diameter. Heat
exchangers: single-pass, horizontal, shell-and-tube type with water and
steam in a floating shell. Control: shim control is by varying fuel
concentration; negative temperature coefficient provides self-regulation
of reactor, Breeding gain: 0.167 of U233. Three such reactors are
designated for a power station. Power of each reactor: 450 Mw(t)--

320 Mw(t) in core, 130 Mw(t) in blanket.

Code: 0312 14 31202 45 625 83779 931 101

84677

115
No. 55 Aqueous Homogeneous Central Station Power Reactor Design No, 1

ORNL

Reference: CF-55-11-35,

Originator: R. B. Briggs.

Status: Conceptual design, 1955,

Details: Thermal neutrons, steady state, breeder. Fuel-coolant-moderator:
solution of UOZSO4 in D20° (No enrichment given.,) Fertile material:
2,25=-ft=thick blanket, surrounding the 5-ft-diameter core, of ThOZ-UO3-D20
slurry. Both fluids circulate, either in the "straight-through" (from
bottom to top) flow arrangement or by means of a "concentric inlet and
outlet."” Operating temperature at the inlet: 482°F; at outlet: 572°F,
Pressure: 200 lbs/sq in. The core tank constructed of Zircaloy plate,
and the pressure vessel of carbon steel clad with stainless steel,
Control: negative temperature coefficient. Total power: 100 Mw(e)--from
core, 360 Mw(t); from blanket, 80 Mw(t). (See Data Sheet No. 66 for
Design No. 2.)

Code: 0312 14 31202 4X 625 756 84677 931 101

116

No. 56 Aqueous Homogeneous Power Reactor Utilizing
Thorium=Uranium-233

Nuclear Power Group

Reference: NPG-112.

Originators: Nuclear Power Group (American Gas & Electric Service

Corp., Bechtel Corp., Central Illinois Light Co. , Commonwealth Edison
Co., Illinois Power Co., Kansas City Power and Light Co., Pacific Gas
& Electric Co., and Union Electric Co.).
Status: Conceptual design and feasibility study, Feb. 1, 1955; no further
work.,
Details: Thermal neutrons, steady state, two-region breeder. Fuel-
moderator—-coolant: sclution of U233 as uranyl sulfate in D20° The
2-ft-thick blanket is a slurry of ThO2 in DZO; the breeding gain is 0.1l1.
Blanket in spherical annulus between the 6-ft-diameter Zr core vessel
and the 11 ft, 4~in. diameter pressure vessel of stainless-steel-clad
carbon steel. A 175-ft-diameter, pressure-~tight sphere contains the
reactor plant. Fuel solution enters at the top of the core through the
24=in., inner pipe at 258°C and exits at 300°C through the annulus between
the two concentric pipes forming the inlet and outlet connections to the
vessel. After leaving the core, the solution enters a centrifugal gas
separator, from which it passes to 6 heat exchangers. It is then returned
to the core by 3 canned-rotor circulating pumps. Pressure is 1800 psia.
Control: by fuel concentration and negative temperature coefficient;
no control rods. Power: three steam generators to produce 180 Mw(e)
net, 360 Mw(t) from core and 280 Mw(t) from the blanket for total of
640 Mw(t).
Code: 0312 14 31202 45 625 756 83779 941 101

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117
No. 57 Single-Fluid Two-Region (SFTR) Aqueous Homogeneous
Power Plant

NPG and Babcock & Wilcox

Reference: NPG-171 (BAW=7).

Originators: Nuclear Power Group (see Data Sheet No. 56 for member

companys) and Babcock & Wilcox Co., Atomic Energy Division.

Status: Conceptual design and feasibility study, July 1957; no further work.
Details: Thermal neutrons, steady state, breeder. Fuel-moderator-coolant:
solution of U233 as uranyl sulfate in D,0. TFertile material: blanket of

2

14 cylindrical assemblies, filled with ThO, pellets and constructed of

Zircaloy-2 with internal flow baffles, allzarranged around the periphery

of the core. The fuel solution enters from a ring header below the

reactor vessel and circulates up through the ThO2 pellet blanket. Fuel
solution leaves blanket assemblies at the top, flows down through the core
region to exit headers from which it passes to four groups of six boiler
heat exchangers each. The design calls for a standby cooler, because the
blanket must be cooled at all times. System pressure: 1500 psia; inlet
temperature: 514°F; outlet: 572°F. No core vessel as such; one

vertical, cylindrical pressure vessel, 9 ft 11 in. ID and 11 ft 9 in. long,
with ellipsoidal heads contains the active region 5 ft 4 in. diameter,

the blanket assemblies (18 in. diameter) and reflector consisting of
alternate layers of stainless steel and the fuel solution (9.5 in. thick).
Blanket assemblies are rotated incrementally by a hydraulically-operated
mechanism to equalize breeding of U233 in the pellet beds and to minimize
absorption of neutrons by protactinium. Control: varying the fuel
concentration. Power: calculations for various power levels are included;
NPG and B & W concluded that 150,000 kw(e), 520 Mw(t) would best serve
their purposes.,

Lode: 0312 14 31202 45 625 756 83779 941 101

118

No. 58 HAR Model

AERE

Reference: Unpublished information.

Originator: C. L. Brown.

Status: Proposal, based on calculations, suggested as basis for further
study, August 1955.

Details: Thermal neutrons, steady state, two-region breeder. Fuel:

enriched uranyl sulfate dissolved in D,0, 1 gm U233 per liter. Fertile

2

material: slurry of ThO2 in D20, 1 kg ThO2

either a right cylinder, 5 ft diameter and 5 ft high, or a sphere,

per liter. Core vessel:

5 ft, 9 in. in diameter. TFuel solution (at a flow of 33,000 gpm) and

the fertile slurry are circulated into the reactor, out of it to separate
heat-exchange systems, and back to the reactor. System pressure: 1000
psi. Total D20 required: 34 tons. Total uranium required: 17 kg to
start up reactor plus 20 kg to achieve 0.1%7 uranium concentration in
blanket. Total power: 350 Mw,

Code: 0312 14 31202 44 625 756 8XXXX 931 101

84677

No. 59 Homogeneous Aqueous Reactor

AERE

Reference: Unpublished information,

Originators: AERE staff.

Status: Design in process, March 14, 1956; later suspended.

Details: Thermal neutrons, steady-state, two~region breeder. Fuel-
moderator-coolant: uranyl sulfate dissolved in DZO: 1.3 g U233 per liter,
Fertile material: blanket (70 cm thick) of ThO2 suspended in DZO’ 1l kg

per liter, Core vessel: spherical zirconium pressure vessel, wall
thickness of 1/2 in, and an ID of 10 ft. Core solution and blanket slurry
are cooled by circulating, in separate circuits, to outside heat exchangers,
Inlet temperature: estimated at 240°C; outlet: 290°C; system pressure:
1500 psi. Total power: 600 Mw.

Code: 0312 14 31202 45 625 756 84677 931 101

119
No. 60 Swedish Homogeneous Solution or Suspension Reactor

Allmanna Svenska Elektriska Aktiebolaget

Reference: French Patent 808,774.

Originator: Staff members.

Status: French patent granted, July 11, 1960.
Details: Thermal neutrons, steady state, two-region breeder. Alternative
fuels: solution of uranyl sulfate, suspension of uranium oxide, or
solution of plutonium sulfate, all in D20. Alternative fertile materials:
ThO2 slurry, thorium nitrate solution, or uranium oxide slurry, all in
DZO' Thorium oxide slurry used in example. Critical concentration in
core: 1 g U235 per liter. Concentration ThO2 in blanket: 1000 g ThO2
per liter. Core vessel: cylinder, 150 cm diameter, 300 cm high, with
volume of 7 m3. Core solution and fertile blanket circulated to separate
heat exchangers. Means provided to exchange material between core and
blanket and to separate slurry into solids and liquid. Arrangement
claimed to have advantages of both one- and two-region breeders. System
pressure: 105 kg per cm2. Total power (core plus blanket): 550 Mw(t),
Code: 0212 14 31202 46 625 746 8XXXX 931 101

41 635 756 84677

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120

No. 61 U233 Breeder—U235 Converter Reactor

ANL

Reference: Nuclear Sci. and Eng., 1, pp. 343-54, October 1956.

Originators: L. I, Katzin and B. I. Spinrad; authors give credit to M,

Treshow for originating the design concept,

Status: Conceptual stage, 1954; no further work,

Details: Thermal neutrons, steady state, two-region breeder (with little
modification of operating conditions and none of structure, reactor can be

235 _ U233 233

operated as a U converter). Fuel-moderator: solution of U

in D20 (enrichment not specified for converter), contained in 72
zirconium fuel tubes (1 cm diameter) per fuel unit or subassembly (15 cm
diameter), of which there are 80. Tubes are hung from the top grid so
that the average height of tubes from the tank bottom is 1 ft. Around
the fuel unit is the coolant guide (or sleeve) of thin-wall Zr or Al.

The coolant-breeding blanket: slurry of Th232 (preferably) in DZO’
flowing into the plenum or header system at the bottom of the tank and

up past the fuel tubes extracting heat from the fuel system. At a height
of 11-1/4 ft sleeves are slotted so that the slurry coolant flows over
into the moderator. It is extracted at the tank bottom through coolant
exit points. Core vessel: steel tank, 12 ft in diameter and 13 ft high.
Sleeves are located in a 10 ft diameter area to allow a 2 ft reflector
area, presumably the DZO' The outlet and inlet sides of the tubes have a
common header box, the bottom coated with neutron-absorbing material

(Cd, B, or Au) located 12 ft above the tank bottom. Fuel goes from there
to a splash chamber, above which is a xenon trap and DZO recombiner,

then to a pump before returning to the tubes. Control: not described.
Power: 220 Mw(t) from the core and 358 Mw(t) from the blanket.

Code: @311 14 31302 45 625 756 84677 92 106

0312 4X 8XXXX

121
No. 62 600-Mw Homogeneous Slurry Pile

Manhattan District, Corps of Engineers, U.S5. Army

Reference: CC-1383.
Originator: Columbia University Group.

Status: Calculations, 1944,

Details: Thermal neutrons, steady state, one-~region converter, Fuel:
slurry of enriched (0.71 mol 7% U233) U03 in DZO (200 g per liter).
50 toms DZO In system., Slurry presumably cooled by circulation to
external heat exchanger. Spherical reactor, 420 cm diameter, 37,000
liters volume. Power: 600 Mw(t).

Code: 0311 14 31302 42 635 753 84677 921 101

No. 63 Homogeneous Slurry Pile

Clinton Laboratories

Reference: Unpublished report.

Originator: Laboratory staff.

Status: Proposal, 1945.

Details: Thermal neutrons, steady state, one-region breeder. Fuel-
moderator: natural uranium oxide suspended in D20, 250 g per liter,
Suspension pumped (34,600 gpm) under pressure to stainless-steel heat
exchangers for cooling by H20. Coolant inlet temperature: 122°F;

outlet temperature: 248°F, Power: estimated at 750 Mw(t), if 50 toms of
D20 were used.

Lode: 0312 14 31302 41 635 752 84677 921 101

8XXXX

122
No. 64 Heterogeneous Slurry Pile

Clinton Laboratories

Reference: Unpublished report.

Originator: Laboratory staff.

Status: Proposal, 1945,

Details: Thermal neutrons, steady state, breeder. Fuel: slurry of
natural uranium as UO2 in D20, 3 g U per cc. Moderator: D20° Reactor
vessel: vertical cylindrical shell-and-tube exchanger. Slurry is
inside tubes immersed in D20 moderator. Separation of concentrated

slurry from D,0 claimed to give higher multiplication constant than for

homogeneous siurry. Reflector: graphite around reactor, Slurry heated
by passing through tubes, cooled by external heat exchangers, and
returned to reactor. Moderator can be cooled by separate heat~exchange
system. Power: probably exceeds 50 Mw(t).

Code: 0312 14 31302 41 635 752 84677 923 104

8XXXX

No. 65 Natural Uranium-Pu Producer Homogeneous Reactor

ORNL

Reference: ORNL-527, p. 10.
Qriginator: HRP staff.

Status: Considered a long term objective, 1949,

Details: Thermal, steady state, converter, Fuel-moderator-coolant:
slurry of natural uranium oxide in DZO—-about 200 g/l with total
volume of about 30 m3, Alternative fuel: solution of U02F2 in D20,
run at high pressure and operated at 100°C. Concept offers the
possibility of being the most economical Pu producer. Total power:
600 Mw(t), from 6 metric tons of uranium,

Code: 0311 14 31302 41 635 752 84677 921 101

625 742

123

No. 66 Aqueous Homogeneous Central Station Power Reactor Design No. 2

ORNL

Reference: CF-55-11-35.

Originator: R. B. Briggs.

Status: Conceptual design, 1955,

Details: Thermal neutrons, steady state, one~region breeder. Fuel-
coolant-moderator-fertile material: slurry of ThOZ-UO3-D20, Core vessel:
12 ft diameter, of stainless-steel-clad carbon steel. Fuel flow may be
"straight-through," from bottom to top of core, with perforated plates at
the inlet to insure smooth expansion of fuel. In this arrangement,

there is essentially no separation at the wall and no recirculation of the
fluid in large eddies. Alternatively, fuel flow may be accomplished by
means of a "concentric inlet and outlet™ system. Inlet temperature:
482°F; outlet: 572°F; operating pressure: 200 psi. A natural circulation
steam generator extracts heat from the circulating slurry. Control:
temperature coefficient of reactivity. Power: 440 Mw(t), 100 Mw(e).

(See Data Sheet No. 55 for Design No, 1l.)

Code: 0312 14 31302 4XX 635 756 85779 932 101

No. 67 One-Region U238 Converter and Power Reactor

ORNL

Reference: CF-56-1-26, p. 9.

Originator: HRP staff,

Status: January 1956; one of five reactors evaluated for possible large-
scale development. Project terminated, 1961.

Details: Thermal neutrons, steady state, one-region converter. Fuel=-
moderator-coolant: slurry of uranium trioxide and plutonium dioxide in
DZO; 300 g uranium per liter; 5.6 g fissile material per liter,

Maximum fuel temperature: 300°C, Reactor pressure: 1800 psia., Total
syétem volume, 46,000 liters. Reactor diameter: 12 ft. Conversion
ratio: 0.88., Power: 440 Mw(t); 100 Mw(e) from generating plant.

Code: 0311 14 31302 47 635 753 84677 921 101

124

No. 68 HRE-4

ORNL

References: ORNL-2654, pp. 45-523; ORNL-2743, pp. 67-9; ORNL-2561,
pp. 63-7.
Originators: HRP staff.

Status: Design studies, Oct. 31, 1958.

Details: Thermal neutrons, steady state, burner (could be breeder).
Fuel-moderator-coolant: slurry of ThO2 with enriched U23502 in DZO'
It flows upward through the spherical core and down through the heat
exchanger. A pump would be added if necessary. System is pressurized
by boiling the slurry continuously in the high~pressure dump tank.
Operating temperature: 275°C (average); pressure: 1500 psia. Core
is 4.5 ft in diameter with a volume of 1520 liters. Control: by

moderator during normal operation. Design power level: 5 Mw(t).

Code: 0312 14 31302 44 635 756 84677 921 101

No. 69 Slurry-Core, Slurry-Blanket, Power Breeder

ORNL

Reference: CF-59-7-129, pp. 19-20.
Originators: HRP staff.

Status: 1959, conceptual design; project cancelled, 1961.

Details: Thermal neutrons, steady state, two-region breeder. Fuel-
moderator: slurry of ThO2 and highly enriched uranium-235 (presumably
as oxide) in D20; 200 g Th per liter, 14.3 g U per liter. Fertile
material: 2-ft thick blanket of thorium in D20 (1000 g Th per liter).
Fuel slurry circulates through core at 90,000 gpm, entering at 256°C.
Core vessel: zirconium-alloy cylinder, 7 ft diameter, 21 ft long.

It circulates through six parallel steam—generator and pumping circuits,
and returns to the reactor vessel. Steam is produced at 400 psia and
445°F, Core system is pressurized by heating slurry in a pressurizing
vessel, Blanket slurry circulates through two parallel circuits, only
one of which contains a heat exchanger. Core and blanket pressurizers
are connected to prevent excessive pressure difference across the core
tank. Conversion ratio: 1.07-1.10. Power: 1140 Mw(t), 333 Mw(e).
Code: 0312 14 31302 44 635 756 84677 931 101

125
No. 70 Homogeneous Power Producer

ORSORT

Reference: CF-51-8-137.

Originators: B. W. O. Dickinson, K. A. Hub, G. P. Letz, W, A. Redfield,
J. N. Renaker, R. A. Wall, and R. J. Beeley.

Status: Design and feasibility study; term paper, 1951; no further work.

Details: Thermal neutrons, steady state, converter. Fuel-moderator-
coolant is a natural UOB—DZO slurry at 250°C under 1000 psi. Pu produced
in core remains as a supplementary fuel source. Reactor core: 27 ft
diameter stainless-steel sphere of the multi-concentric shell type,
cooled by 1 in. wide layer of circulating HZO at 800 psi between the first
pressure shell and the outer one. Slurry enters at bottom of the reactor
in an annulus surrounding the fluid outlet pipe, passes between first and
second inner shells of the core, and enters core through system of
orifices in the inner shell. Slurry then leaves the reactor through a
flared exit tube, which extends 9 ft into the reactor. Fuel is pumped
to 16 heat exchangers and then back to the reactor. No reflector other
than pressure shells. Control: boron steel rod clad with stainless-
steel; concentric with the liquid outlet, the vertical rod is hydraulically
operated; other control is by varying the DZO/U ratio (shim) and by the
negative temperature coefficient. Power: 2000 Mw(t), 400 Mw(e).
Code: 0311 14 31302 41 635 752 81141 91X 101

83779

84677

126
No. 71 Homogeneous Slurry Reactor

ORSORT

Reference: Unpublished report,

Originators: W. G. Atkinson, et al.

Status: Conceptual design; term paper, August 1951.
Details: Thermal, steady state, breeder. Fuel-moderator-coolant:
,0 slurry (14.3 ¢ UO3 to 100 g D20)°

Slurry circulated to 16 vertical tube heat exchangers. System

natural uranium trioxide in D

pressurized at 1015 psi with steam. Reactor temperature: 250°C.

Two 21-ft reactors combined to produce power; 21.5 tons natural uranium
required; 160 tons DZO' Cylindrical pressure vessel, 21 ft in diameter,
high-tensile carbon steel lined with stainless steel. Control: mnegative
temperature coefficient, concentration control, injection of xenon,

and Hg-filled tubes; the tubes, which pass through the center of the core,
can add Hg or other poison. Reactor produces saturated steam, at 250 psi,
which is superheated by combustion of radiolytic gases. Power per

reactor: 500 Mw(t). Plutonium produced can be either used as fuel or

separated.

Lode: 0312 14 31302 41 635 752 84677 921 101
83779
81585
81596

127
No. 72 Pennsylvania Advanced Reactor

Westinghouse Electric Corp., and Pennsylvania Power and Light Co.

References: Proc. 2nd U.N. Conf. on Peaceful Uses of Atomic Energy, 9,

pp. 202-210; Proc. Am. Power Conf., 19, pp. 640-50, 1957.

Originators: Staff members.

Status: Reference design completed December 1958; project discontinued.
Details: Thermal neutrons, steady state, one-region breeder. Fuel:

uranium oxide suspended in DZO’ with a uranium and proactinium concentration
of 9 g per kg D,0. Fertile material: ThOz, suspended with the uranium

2
oxide in D,0, with a thorium concentration of 251 g per kg D20° Moderator

and coolani: DZOo Slurry of uranium and thorium oxides i1s pumped into
reactor vessel. It leaves through four loops to heat exchangers that
supply steam to a secondary system for power generation, and is pumped
back to the reactor. Slurry enters reactor at 465°F and at a mass flow
rate of 16.5 x 106 1b per hr, and leaves at 580°F. System pressure:
2000 psi. Reactor vessel: spherical, 13 ft in diameter. Equipment
included for recombining radiolytic gases and reprocessing slurry. Control:
negative temperature coefficlent and varying slurry concentration; for
emergency shutdown, solution is dumped. Power: 550 Mw(t).
Code: 0312 14 31302 44 634 756 84677 921 101

83779

83179

128
No. 73 KEMA Reactor I (SUSPOP)

RCN-KEMA Reactor Development Group

References: Proc. 2nd U.N. Conf., 12, pp. 525~528; Philips Tech.
Review, 21, No. 4/5, pp. 109-152, March 12, 196Q.

Originators: J. J. Went, et al.

Status: March 12, 1960; in operation.
Details: Thermal neutrons, single-region subcritical assembly. Fuel-
moderator-coolant: suspension of uranium dioxide (20% U235) in HZOQ
Fuel circulates through heat exchanger. Neutron flux maintained by
external neutron source. Reactor vessel: stainless-steel cylinder
28 cm diameter, with cone-shaped ends. Reflector: blocks of Be0O around
the reactor vessel and blocks of graphite surrounding Be0O. Control:
adjustment of concentration of suspension and heat removal, plus control
rods; three safety rods of boron carbide, one of which is control rod;
rods, suspended by electromagnets, drop into the reflector; emergency
dumping of suspension. SUSPOP is stage in developing homogeneous
suspension reactor,
Code: 0333 13 3101 43 634 754 83779 931 101

81111

84677

81379

No. 74 KEMA Reactor BABYPOP

RCN-KEMA Reactor Development Group

Reference: J, British Nucl. Energy Conf., 2, pp. 395-407, October 1957,

Originators: Laboratory staff.

Status: Unknown.
Details: Thermal neutrons, steady state, zero power, converter,
Details otherwise the same as for SUSPOP.
Code: 0311 13 3101 43 634 754 83779 931 101
81111
84677
81379

129

No. 75 250 Kw Aqueous Homogeneous One—Region Suspension Reactor

RCN-~KEMA Reactor Development Group

References: Proc. lst U.N. Conf, on Peaceful Uses of Atomic Energy, 3,

pp. 116-120; Proc. 2nd U.N. Conf, on Peaceful Uses of Atomic Energy, 9,
pPp. 427-44Q; IAEA Directory, p. 191.

Originators: P. J. Kreyger, et al.

Status: 1958, design; construction of model experiments in progress, 1962.
Construction under consideration, 1962.
Details: Thermal neutrons, steady state, converter. Fuel-moderator-
coolant: suspension of 20Z enriched U02 (4Z by volume) in HZO’ particle
size 4-13 u. System pressurized with hydrogen to reduce oxidation, water
decomposition, and stress corrosion., Slurry pumped through reactor vessel,
gas-liquid contactor, gas separator, and heat exchanger. By-pass stream
continuously removes portion of slurry for removing small slurry particles,
as well as products formed by fission, corrosion, and erosion. Closed-
cycle operation. Inlet slurry temperature: 230°C. Outlet temperature:
250°C. Outlet steam pressure: 40 atm., Total system pressure: 60 atm,
Reflector: BeO, surrounded by graphite., Circulating suspension can be
cooled or heated by a flow of water into a slurry cooler through which
the circulating slurry passes. Activated charcoal added to slurry to
reduce sorption of insoluble fission compounds on fuel particles or
reactor walls. Control: three boron carbide shim safety rods in the
reflector; negative temperature coefficient. Power: 250 kw.
Code: 0311 13 31301 43 634 754 81111 921 101

31101 84677

83779

130

No. 76 Canadian Two-Zone Slurry Reactor

Reference: CRNE-558 (AECL-249).

Originator: Chalk River Project, Atomic Energy of Canada, Limited,

Status: Calculations, 1953.
Details: Thermal neutrons, steady state, two-region breeder. Fuel:
slurry of natural uranium oxide in DZO’ enriched with plutonium, Fertile
material: slurry of natural uranium oxide. Reactor consists of two
concentric spheres, inner sphere containing the fuel slurry. The space
between the spheres, which comprises the bulk of the reactor, contains
the blanket slurry of natural uranium oxide., Pressure on slurry:
1000 psi.
Code: 0212 14 31302 47 635 752 8XXXX 931 101

84677

No. 77 10 Mw Homogeneous Suspension Reactor (HR-1)

Institute of Nuclear Research, Czechoslovak Academy of Sciences

References: Proc. 2nd U.N. Conf. on Peaceful Uses of Atomic Energy, 9,

PP. 441-446; IAEA Proc., Symp. Power Reactor Exp., 1962, pp. 187-195,

Originators: Staff members.

Status: Conceptual design, July 1961; project discontinued.
Details: Thermal neutrons, steady state, converter. Fuel-moderator-
coolant: suspension of enriched (157%) U308 in H,0. Circulates into
core, through heat exchanger, and back to core, Provisions for
separating radiolytic gases. Operating temperature: 280°C; design
pressure: 120 atm (1764 psi)., Suspension flow rate: 65 liters per sec.
Cylindrical core vessel, lined with stainless-steel, with 90 cm diameter
and 135 cm height. Control: changing concentration of suspension;
vertical safety rod; emergency dumping. Power: 10 Mw(t), 2.5 Mw(e).
Code: 0311 13 31301 43 624 754  8l1XX 921 101

83779

84677

81779

131
11.

12.

13.

14,

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"Nuclear-Powered Locomotive's Economic Feasibility Questioned by

Railroad Men," Nucleonics, 12, No. 3, pp. 78-80, March 1954.

"General Design Criteria for USAF Nuclear Power Plant Applications;

a Final Engineering Report," GAI-1499 (AS-225792 or RADC-TR-59-132),
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P. R. Clark, W. O. Chatfield, A. E. Cox, J. M. Detwyler, J. A. Murphy,
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Del. , ORSORT, August 1954,

D. W. Montgomery, W. J. Dodson, F. F. Kaiser, W. K. Luckow, and T, J.
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for Producing Electrical Power in Remote Locations," CF-53-10-22,
ORSORT, August 1953,

136

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83
84

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88

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91.
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93,

W. E, Kinney, R. Brodsky, D. Hillis, J. T. Wagner, and T, J. Ward,
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R. H, Chapman, H., L. Collins, W, J., Dollard, D, Fieno, J. Hernandez-
Fragoso, J. W, Miller, H. von Hollen, and C. V. Wheeler, "An 80
Megawatt Aqueous Homogeneous Burner Reactor,'" CF-57-8-6, ORSORT,
August 1957,

C. B, Ellis, "The 10,000 Kw Aqueous Homogeneous Reactor," Proc.

Annual Conf. of Atomic Industrial Forum, Inc., Forum Report No. 12,

IIl, p. 43, September 1956,

M. I, Lundin and R, Van Winkle, '"Conceptual Design and Evaluation
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Ref. 1, pp. 473-479.

"Homogeneous Circulating Fuel Reactor Power Plant: Conceptual Design
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D, H, Fax, "Proposed 80,000 Kilowatt Homogeneous Reactor Plant;
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Ref. 34, p. 8.

B. A. Mong, J, E, Colgan, R, A, D'Elia, J. S. Mooradian, G, K, Rhode,
and P, M. Wood, "A Design Study of a Low-Power Aqueous Homogeneous
Boiling Reactor Power Plant,'" BW=~AED~-502, Babcock and Wilcox Company,
June 1955.

P, R. Kasten and H. C. Claiborne, '"Fuel Costs in Homogeneous U=~235
Burners," Nucleonics, l4, No. 11, pp. 88-91, November 1956.

L. W. Nordheim, unpublished report, Clinton Laboratories, Nov., 16, 1944,
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Long=Range Planning Group, unpublished report, ORNL, 1950,

R. H, Ball, A. M. Hallene, R, J. March, "An Economic Study of a 1000
Megawatt Homogeneous Reactor," CF-52-8«7, ORSORT, Aug. 12, 1952,

"A Third Report on the Feasibility of Power Generation Using Nuclear
Energy,' CEPS=1121, Del., CEPS Group of Commonwealth Edison, June 15,
1953,

137
94, H. C. Claiborne and M. Tobias, 'Some Economic Aspects of Thorium Breeder
Reactors," ORNL-1810, ORNL, Oct. 27, 1955.

95. H. C. Claiborne and T. B. Fowler, "Fuel Cost of Power Reactors Fueled
by U0,50,-11,80,-D,0 Solution," CF-56-1-145, ORNL, Jan. 30, 1956.

96. "Reactors Developed in the USSR and Its Bloc Countries,"™ AD-248402,
Air Information Division, Oct. 31, 1960.

97. R. Oravec, '"'Survey of Soviet Reactors to September 1958,'" AECL-1011,
Atomic Energy of Canada, Ltd., 1959.

98. Ref. 8, pp. 5-7.

99. Ref. 7, pp. 2-3.

100, E. P, Wigner, A, Weinberg, and G. Young, unpublished report, Clinton
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101. E., P. Wigner, unpublished report, Clinton Laboratories, 1945,

102, E. P, Wigner, "New Ideas for Nuclear Reactors,'" Nuc. Sci. and Eng., 6,
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103. J. Bick , J. M. LaRue, P. N, Haubenreich, D. Rauch, T. Williams, and

G. Putnam, "Reactor Design and Feasibility Problem, U-233 Power Breeder,'
CF-51-8-213, ORSORT, Aug. 6 , 1951.

104, W. E. Edwards, G. E. Garker, F. D. Orazio, P. 0., Nadler, L. H. Thacker,
P. N. Wood, W. E. Unbehaun, and D. T. Bray, unpublished report, ORSORT,
August 1951.

105, R. A, Thomas, J., F. Brunings, L. S. Hall, C. Michelson, J. F., Parrette,
L. R. Pletke, R. L. Whitelaw, and W. A. Wittkopf, "Ultimate Homogeneous
Reactor," CF-54-8-239, ORSORT, August 1954, p. 5.

106, A, Hauspurg, J. M. Finan, J. W, Geiser, B. H, Hamling, J. J. Happell,
K. Moore, and G. R. Thomas, '"High-Temperature Aqueous Homogeneous
Reactor," CF-55-8-191, ORSORT, August 1955.

107. Sidney Visner, "A Two-Region Homogeneous Reactor Without a Core Tank,"
CF-54-6-180, ORNL, June 23, 1954,

108. R. B. Briggs, "Calculations for the K-49 Reactor," unpublished report,
ORNL, March 11, 1954,

109. Unpublished report, ORNL, 1950,

110, '"Power Plants with Thermal Reactors, An Engineering and Economic
Analysis," MIT-5003 Del., Mass. Inst. Tech., Sept. 15, 1953,
pp. 263=74.

138
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112.

113,

114,

115.

117.

118.

119.

120.

121.

122,

123.

124,

R. B. Briggs, '"Aqueous Homogeneous Reactors for Producing Central
Station Power," CF-55-11-35, ORNL, Nov. 8, 1955.

W. L. Carter to H. E. Goeller and W. E, Unger, "Preliminary Design
Study of Proposed Plutonium Loop," CF-55-3-178, ORNL, March 21, 1955,
T. G. LeClair, "A Current Design for a Full-Scale Reactor Power

Plant,'" Nuclear Engineering, Part III, Chem, Engr, Prog. Symposium

Series, 50, No. 13, Am. Inst. Chem. Engrs., New York, 1954, pp. 122-8,

H. G. Carson and L. H. Landrum, eds., 'Preliminary Design and Cost
Estimate for the Production of Central Station Power from an Aqueous
Homogeneous Reactor Utilizing Thorium-Uranium-233," NPG-112, Chicago
Nuclear Power Group, Chicago; Babcock and Wilcox, Lynchburg, Feb. 1, 1955,
Sidney Visner, '"Nuclear Calculations for Homogeneous Reactors Producing
U-233," CF-51-10-110, ORNL, Oct. 22, 1951, Decl., Feb. 14, 1957,

P. N. Haubenreich, "Calculations for Thorium— and Uranium~Fueled
Reactors,'" CF-53-12-1, ORNL, Feb. 8, 1954, Decl. Feb. 13, 1957,

M. Tobias, P. N. Haubenreich, and R. E. Aven, "Conversion in a Two-
Region Reactor," CF-53-2-134, ORNL, Feb, 16, 1955, Decl. Dec., 20, 1955,
E. D. Arnold, A. T. Gresky, A, R. Irvine, and R. J. Klotzbach,
"Preliminary Cost Estimation: Chemical Processing or Fuel Costs
Associated With a Power Station of Three K-23 Reactors,'" CF=-54-7-63,
ORNL, July 14, 1954, Decl. Nov. 13, 1957.

"Single-Fluid Two-Region Aqueous Homogeneous Reactor Power Plant;
Conceptual Design and Feasibility Study. Final Report," NPG-171
(BAW-7) , Nuclear Power Group, Chicago; Babcock and Wilcox, Lynchburg,
July 1957.

C. L. Brown, "A Preliminary Assessment of Two-Zone Heavy Water
Homogeneous Systems," unpublished report, Atomic Energy Research
Establishment, Harwell, England, August 1955,

R. Hurst, I. Wells, and D. Newby, "The Homogeneous Aqueous Reactor,"

J. British Nucl. Energy Conf., 2, pp. 395-407, October 1957.

Unpublished report on a homogeneous aqueous power reactor, Atomic
Energy Research Establishment, 1956,

P. G. Jones and R. G. Snowden, "Thorium Nitrate Solution as a Breeder
Blanket in the H.A.R. Pt. 1, Thermal Stability," AERE C/M298

(HARD(B) /P-26), Atomic Energy Research Establishment, 1956,

K. Ishikawa, ''Aqueous Homogeneous Reactor," Conference on the Peaceful
Uses of Atomic Energy, May 13-17, 1957, Japan Atomic Industrial Forum
and U. S, Atomic Industrial Forum, 1957, pp. 86-95,

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126.

127,

128.

129.

130.

131.

132.
133.
134.

135,

136,

137.

138,

139.

140,

V. Vartosheck, "Utilization of Natural Uranium in a Homogeneous

Reactor," Soviet Journal of Atomic Energy, 7, No. 6, pp. 981-6, April 1961,

Allmanna Svenska Elektriska Aktiebolaget, "Homogeneous Nuclear Reactor,"
French Patent No. 808,774, July 11, 1960.

L. I, Katzin and B. I. Spinrad, "U-233 Breeder—-U-235 Converter Reactor,"
Nuclear Sci. and Eng., 1, pp. 343-354, October 1956.

W. A. McRae and T. L. 0'Connor, "Study of a New Fuel Concept for
Aqueous Homogeneous Reactors," Q-317, Ionics, Inc., Oct. 21, 1959.
Ref. 1, p. 129,

I. Kirshenbaum, G. M. Murphy, and H. C. Urey, eds., "Utilization of
Heavy Water," TID-5226, Columbia University, New York, 1951, Decl.
Feb. 27, 1957; published in National Nuclear Energy Series III-4B.

C. F. Hiskey, "Chemical Research. The Heavy-Water Homogeneous Pile:
A Review of Chemical Researches and Problems," CC-1383, Columbia
University, 1944.

Unpublished report, Clinton Laboratories, 1945.

Unpublished report, Clinton Laboratories, 1945.

C. E. Winters and C. H. Secoy, eds., ''Homogeneous Reactor Experiment
Quarterly Progress Report for Period Ending November 30, 1950,"
ORNL-925, ORNL, Jan. 30, 1951, Decl. March 2, 1957.

W. E. Thompson, comp., '"Homogeneous Reactor Experiment Quarterly
Progress Report for Period Ending February 28, 1951," ORNL-990,

ORNL, May 18, 1951, Decl. March 1, 1957.

W. E. Thompson, comp., "'Homogeneous Reactor Project Quarterly Progress
Report for Period Ending May 15, 1951," ORNL-1057, ORNL, Oct. 10, 1951,
Decl. March 12, 1957,

M. W, Rosenthal and M. Tobias, "Nuclear Characteristics of Two~Region
Slurry Reactors," CF-56-12-82, ORNL, Dec. 20, 1956.

"Homogeneous Reactor Project Quarterly Progress Report for Periods
Ending April 30 and July 31, 1958," ORNL-2561, ORNL, Feb. 4, 1959,
pp. 63-67.,

"Homogeneous Reactor Project Quarterly Progress Report for Period
Ending October 31, 1958," ORNL-2654, ORNL, Feb. 24, 1959, pp. 45-52.
"Homogeneous Reactor Project Quarterly Progress Report for Period

Ending April 30, 1959," ORNL-2743, ORNL, Aug. 8, 1959.

140
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142,

143.

144.
145,

146.

147.

148,

149,

150.

151.

152,

B. W. O. Dickinson, K. A. Hub, G. P, Letz, T. A, Redfield, J. N.
Renaker, R. A. Woll, and R. J. Beeley, "Reactor Design and Feasibility
Problem., Power Producer," CF-51-8-137, ORSORT, Aug, 6, 1951,

W. G. Atkinson, H. L, Leichter, Jr., G. N. Mahoffen, T. J. Morley,

A, M. Rubenstein, M. Shaw, C. M. Wetzel, and L. S. Mims, unpublished
report, ORSORT, August 1951,

J. J. Grebe, "Why is Dow-Detroit Edison Working on a Fast Breeder

Reactor for Power," Nucleonics, 12, No. 2, pp. 13-15, February 1954,
Nucleonics, October 1957, p. 19; June 1958, p. 25,

W. E. Johnson, S. Bartnoff, D. H. Fax, and E. U. Powell, '""Design

Considerations for the Pennsylvania Advanced Reactor Plant,'" Proc., 2nd

U.N. Int, Conf. on Peaceful Uses of Atomic Energy, 9, pp. 202-210,
United Nations, New York, 1958,
W. E. Shoupp, "The Homogeneous Reactor Project--PAR," Proc, Conf,

on Peaceful Uses of Atomic Energy, Tokyo, May 13=17, 1957, Japan

Atomic Industrial Forum and U., S. Atomic Industrial Forum, 1957,
pp. 80-86,

W. E. Johnson, D. H. Fax, and S. C. Townsend, "The PAR Homogeneous
Reactor Project--Plant Design and Operating Problems," Proc, Am,
Power Conf., 19, pp. 640-50, March 1957,

"Atomic Power Development at Westinghouse,'" Power Engineering, 62,

Pp. 44-50, December 1958,

"PAR--Big Homogeneous Dropped," Nucleonics, 17, No. 1, p. 28,

January 1959.
W. E. Johnson, D. H. Fax, H. J. Garber, G. R. Taylor, "Technology
of Aqueous Thoria Slurries for Single-Region Homogeneous Reactors,"

Proc. 2nd U.N. Int., Conf, on Peaceful Uses of Atomic Energy, 7,

Pp. 34-8, United Nations, New York, 1958.

K. J. de Jong, J., A, H. Kersten, J. J. Went, H. H. Woldringh and J. J.
van Zolingen, "A Subcritical Circulating Suspension Reactor," Proc.
2nd U.N. Int. Conf. on Peaceful Uses of Atomic Energy, 12, pp. 525-
528, United Nations, New York, 1958,

J. A. H. Kersten, J. J. Went, and J. J. van Zolingen, "Physical

Aspects of an Aqueous Homogeneous Suspension Reactor," Proc, IAEA

Symposium on Power Reactor Experiments, I, No. 21-26, pp. 197=223,

IAEA, Vienna, 1962,

141
153'

154.
155.

156.

157.

158.

159.

160,

161.

162,

163.

164.

J. J. Went, "Reasons Behind the Choice of a Homogeneous Suspension
Reactor,'" Phillips Tech. Review, 21, No. 4/5, pp. 109-21, March 12,

1960. (Part I of "Instrumentation for a Subcritical Homogeneous

Suspension Reactor," pp. 109-156.)

Ref. 15, 3, pp. 190-191,

P. J. Kreyger, Th. van der Plas, B. L. A. van der Schee, J. J. Went,
and J. J. van Zolingen, '"Development of a 250 Kw Aqueous Homogeneous

Single Region Suspension Reactor," Proc, 2nd U.,N. Int., Conf. on

Peaceful Uses of Atomic Energy, 9, pp. 427-40, United Nations, New

York, 1958.
H. de Bruyn, M. E. A. Hermans, Th., v. d. Plas, B. L., A. v, d. Schee,
and J. J. Went, "The Design of a Small Scale Prototype of a Homogeneous

Power Reactor Fueled with a Uranium Oxide Suspension,” Proc. lst

U.N. Int. Conf. on Peaceful Uses of Atomic Energy, 3, pp. 116-120,

United Nations, New York, 1955.
J. J. Went and H. de Bruyn, "Liquid-Fuel Reactors with Uranium

Oxides," Nuclear Engineering, Part II, Chem. Engr., Progress Symposium

Series, 50, No. 12, Am. Inst, Chem. Engrs., New York, 1954,

Euratom Fifth General Report, European Community Information Service,

Brussels, April 16, 1962.

C. L. Brown and J. B. Morris, Ref. 3, p. 16.

W. G. Clarke, "A Study of Single-Zone U-233 Thoria Slurry Reactors,"
AERE-R/R-2537, (HARD-(A)/P-48), AERE, Harwell, 1958,

M. F. Duret and I. L. Wilson, "A Two-Zone Slurry Reactor,'" AECL-249,
(CRNE-558), Atomic Energy of Canada, Ltd., Dec. 17, 1953, Decl,
Nov. 9, 1955.

V. Zajic, J. Pfann, and J. Gregor, "Design of an Experimental 10 Mw
Homogeneous Reactor Fueled with Circulating Uranium Oxide Suspension

in Light Water," Proc. 2nd U.N. Int. Conf. on Peaceful Uses of

Atomic Energy, 9, pp. 441-446, United Nations, New York, 1958,

"Hydrodynamical Aspects of a Homogeneous Suspension Reactor," Proc.

IAFA Symposium on Power Reactor Experiments, I, IAEA, Vienna, 1962,

pp. 187-95.
V. Zajfc, "The Development of Homogeneous Reactors and the Experience

with Czechoslovak Project," Jaderni Energie, 7, pp. 296-300, 1961,

142

165. "Terminal Report to U. S. Atomic Energy Commission for December 20, 1954,
to July 9, 1955," AECU-3197, Atomic Power Equipment Division, General
Electric Company, Dec. 2, 1955,

166, "Report of the Fluid Fuels Reactor Task Force to the Division of
Reactor Development, United States Atomic Energy Commission,' TID-8507,
USAEC, February 1959.

167. Unpublished report, ANL, Dec. 18, 1952,

168, J. A. Lane, "Objectives and Potential of Aqueous Homogeneous

Reactors,"‘QQO-Mbderated Power Reactors, A Symposium, March 3-5,
1959, TID=7575, USAEC, August 1959, pp. 84-91.

169. J. E. Jones, "HRT Mockup Runs, CS-25 and CS-26," CF-61-4-96, ORNL,
April 26, 1961,

Additional References

J. S. Leonard and I. H. Mandil, "An Analysis of a Steam Power Cycle as
Applied to a Large Homogeneous Circulating Fuel Reactor," CF-51-1-20,
ORNL, Jan, 11, 1961, Decl. Feb. 14, 1957.

S. E. Beall and C. E. Winters, "The Homogeneous Reactor Experiment,”

Chem. Eng, Progress, 50, No. 5, pp. 256-62, May 1954.

4

H., C. Ott, "The Homogeneous Natural Uranium-D,0 Pile as a Source of Power,'

2
TID-10092 (MonC-374), Clinton Laboratories, January 1948, Decl, Mar., 7,

1957.

I, Kirshenbaum, G. M., Murphy, and H. C. Urey, ed., Utilization of
Heavy Water, USAEC, Oak Ridge, 1951.

R. B. Briggs, "Developments in Reactors with Mobile Fuels," Nuclear

Congress, New York City, June 4-7, 1962, Preprint No. 30, Vol. I,

Engineers Joint Council, New York, 1962.

C. J. Borkowski to J. A. Swartout, "Design Review Committee Report on
the Homogeneous Reactor Test (HRT),'" CF-55-6-53, ORNL, June 7, 1955,
Decl, Feb., 16, 1957,

143
Chapter 4 Boiling Reactors for Power and Breeding

In this chapter boiling reactors are defined as those in which the
heat of fission boils an aqueous solution or slurry in the core, so that
the rising steam condenses on a heat exchanger and returns to the core.
Here again, confusing nomenclature makes it necessary to distinguish these

' as described

reactors from several others of similar names. ''Water Boilers,'
in Chapter 2, ordinarily do not boil: the boiling appearance is due to
evolution of radiolytic gases. These reactors are cooled by circulating the
fuel solution. Such boiling-water reactors as the Experimental Boiling

Water Reactor are not homogeneous, and do not have fuel in solution., Also
excluded are those homogeneous aqueous reactors in which boiling takes

place in the blanket and not in the core., These reactors are discussed iIn
Chapter 3.

The boiling reactors in this chapter are of three general types: one-
region solution; one-region slurry; and two-region slurry. Also included
because of its similarity is a reactor fueled with a slurry or solution
that passes through tubes in the critical region and then to the heat exchanger.

Lane has listed some of the advantages of beoiling as a method of heat
removal.! The reactor responds more quickly to sudden increases in reactivity,
so that power excursions are minimized. Pumps for circulating fuel are not
needed. For the same reactor pressure, steam is delivered to a turbine at a
higher temperature. Corrosion and radioactive contamination of external
systems are reduced because the fuel solution or slurry is not circulated to
an external heat-exchange system. Early doubts as to the nuclear stability
of boiling solutions have been removed, and experiments indicate that boiling
reactors are comparable with non=boiling reactors in power densities obtained.
Much experimental work has been carried out on the reduction and consequent
precipitation of uranyl sulfate solutions in steel reactors. Titanium-lined
equipment and oxygen have been used to solve this problem.

One of the early boiling concepts was that of E. P. Wigner, who, in 1944,
proposed an unusual boiling-solution reactor.? It consisted of a series of
trays, filled with plutonium=-salt solution, that are placed above each other,
With enough trays, the solution becomes critical and boils. Steam is
condensed and returned to the trays. The heat can be recovered as useful
power. Simplicity, no outside holdup, and elimination of fission poisons are
some of the advantages discussed. Entrainment of plutonium salts, bubbling,

and solution instability are among the possible problems.

145

Fewer concepts have been developed for boiling aqueous homogeneous
reactors than for non-boiling ones, and only one has actually been operated
in the U.S.--SUPO, a water boiler, was operated under boiling conditions at
different times to test the feasibility of boiling reactors and to study
variables. A Russian reactor, however, using boiling slurry was constructed,
and a French experimental prototype, PHOEBUS--a reactor physics experiment--is

scheduled for construction.

Feasibility of Boiling Reactors

Because of the doubts that a boiling reactor would be stable, several

preliminary studies were made.

SUPO Experiments

In 1951, after the construction of HRE-1, the Homogeneous Reactor
Project group considered removing heat from a homogeneous reactor by boiling
the fuel solution rather than by circulating it. Experiments on bulk boiling
at atmospheric pressure in a cylindrical tank 1 ft in diameter indicated that
power densities up to 5 kw might be achieved.

To determine nuclear stability, a group from Los Alamos and Oak Ridge
operated SUPO under boiling conditions in October 1951, thereby indicating
that it could be operated safely as a boiling solution reactor at its
maximum power density of 4 kw per liter.3 Experiments were continued in 1952
and 1953 to confirm the stability of operation" and to study such problems
as power removal,® chemical stability,® bubble formation,’ and different
core arrangements,®

A small feasibility program started early in 1951 with the construction
of a simulator to study air bubbles to determine the rate of heat removal
and solve certain control problems.® Later in 1951 after the SUPO experiment,
there was postulated a reactor with boiling liquid flowing up the center of
the core, liberating part of the vapor at the top surface, and flowing down
in an annular region along the wall of the containing vessel, Design of a
small boiling homogeneous reactor was begun, with power output to 100 kw
(2 kw/liter), and pressures to 150 psi.1® It was designed for temporary use
only, to demonstrate higher power densities and void volume than the SUPO
and to give experience in design and operation of a boiling reactor. A
vertical thimble design was incorporated11 and the boiling reactor research

program was continued to investigate the controllability and operating levels,

146

as well as the mechanics of actually designing, building, and operating
such a reactor.l?

Stein and Kasten concluded on the basis of developed kinetic equations
that boiling reactors appear to be inherently stable, although the power
density available may limit their applications.!3 Welton also commented
that kinetic equations indicate boiling reactors would be stable,l"

Kasten reported that stability of boiling reactors should increase
with the power level, and it should also increase as average vapor bubble
size decreases., The reactor pressure should be as high as possible to
improve the power density and yet have sufficient bubble nuclei for good

stability.l®

Teant

Although the experiments with SUPO had demonstrated the stability of
boiling reactors, more work was needed. SUPO had unfavorable geometry for
steam release,!® and, since it had cooling tubes rather than coolant in
the moderator-fuel solution, water boiled out of the cooling tubes. This
boiling removed reactivity from the reactor and made uncertain the amount
of reactivity required to produce various power levels when boiling."
Also, high-pressure density measurements would be required.l’

The preliminary design of a boiling reactor experiment (BRE) was
initiated at ORNL, and it was nearing completion by March 1962. This
apparatus, the TEAPOT, consisted of an unreflected homogeneous reactor of
variable height with a solution of uranyl sulfate (93.4% U235) in water as
fuel and moderator. It was to be run under bulk boiling conditions, with
nuclear power supplying all the energy for vapor formation, at 150 psi and
a thermal power load of 250 kw (4 to 5 kw per liter of solution),ws19
TEAPOT should not be confused with various heterogeneous boiling reactor
experiments leading to BORAX.

In January 1953, the difficulty in preventing reduction and precipitation
of the uranium in a boiling uranyl sulfate solution became apparent, and
construction of the reactor was postponed until experiments in solution-
stability were completed. These experiments, completed in October 1953,
indicated that at oxygen concentrations likely to be encountered in a boiling
reactor (6 to 7 ppm), uranium would be reduced in solution in contact with
stainless steel. Because the metallurgy of titanium or zirconium was not

sufficiently advanced to build a reactor with them, it was decided to

147
abandon the boiling reactor experiment and continue experiments on such
problems as solution stability and steam separation.20 Also, work on

HRE-1 indicated that construction of a second homogeneous reactor experiment
would be preferable,?!

A study was also initiated for a large-scale homogeneous boiling reactor
to exist only on paper (Paper Bolling Reactor, PBR), initially a power and
plutonium producer whose study would uncover problems of constructing an
actual boiling reactor. A 6-ft reactor to operate at one atmosphere was
planned, in addition to TEAPOT.l2 Late in 1952, however, intermediate design
effort on the experiment was reduced and the immediate objective changed from
constructing an operating experiment to developing a mockup of the TEAPOT
for more complete testing of alternate components, Construction of the
building was also to be delayed until after additional studies of the

components and characteristics of the mockup could be made.?2?

Other Concepts

Although there has been comparatively little actual operation of boiling-

core reactors, several concepts have been developed.

One-Region Solution Reactors

In an ORNL concept, boiling uranyl sulfate solution in heavy water is
contained in a cylindrical reactor.!3 The heat of the fission reaction
vaporizes the liquid; the generated steam rises through the liquid and vapor
spaces and is condensed by heat exchangers. Slurry and other types could be
tried also,

The 10 Mw Homogeneous Boiling Reactor, was designed by ORSORT for use in
remote locations.?3 The fuel is 93.4%-enriched uranyl sulfate dissolved in
ordinary water . It is controlled by the level of fuel in the core. Power
removal is predominantly by natural bubble rise and growth.

Another ORSORT concept postulated production of power--1200 Mw(t)--and
plutonium.?* The core is a boiling solution of uranyl sulfate or slurry of
uranium trioxide with an enrichment of about 1% U235. The steam-liquid
mixture from the reactor flows to external vapor separators and the liquid
returns to the core by gravity through external downcomers. Control might
be by change in void fraction or by holdup of the solvent moderator or solute
fuel. It was concluded that: (a) to be economical, boiling homogeneous

reactors must be very large and operate at high pressure and high volumetric

steam fraction, suitable only for large power plants; (b) they appear to be

148

feasible except for unsteady-state behavior; (¢) the cost is equal to that
of non-boiling forced-circulation reactor.

Babcock and Wilcox in 1955 made a design study of a power plant using
a solution of uranyl sulfate in light water as fuel.?® The total thermal
output is 1300 kw, and power is removed from the core by boiling the fuel
solution, The heat is transferred to the secondary steam system by
condensing primary water on the external surface of a bayonet-type boiler
and boiling secondary water within the tubes. In addition, vertical
coolant tubes extend through the solution to the bottom of the pot, The
reactor is controlled automatically by power demand.

Another single-region reactor type that might use solution or slurry
fuel is described in a German patent.2® 1In the pear-shaped reactor vessel,
the solution or slurry boils, and the vapor is condensed on a horizontal
finned cooling tube in the upper portion of the reactor.

A Canadian patent describes a reactor with a core of boiling slurry
or solution.?’ The core acts as a heat generator and also as a vapor
separator. Thus there is a small external circuit so that less fuel and
moderator is needed. The chief feature is a method for separating steam
from liquid rapidly. The solution or slurry iIs injected tangentially to
the core periphery to produce a vortex movement in the liquid. Steam

bubbles are driven toward the center and the vapor escapes at high speeds.

One~Region Boiling-~Slurry-Core Reactors

An ORSORT group published a preliminary design and feasibility study
for a plutonium power producer, in which the fuel is boiling uranium
trioxide slurry, either natural or slightly enriched, in heavy water,28
Heat is removed by the saturated vapor leaving the reactor shell and
passing directly to the turbines or through chemically fired superheaters
upstream from the turbines. Control is by increasing or decreasing the
reactive volume or changing the moderator-to-uranium ratio.

Another ORSORT reactor yields 500 Mw(t) and produces uranium=233, The
fuel is uranium dioxide and thorium dioxide suspended in heavy water.
Natural circulation, with internal liquid-vapor cyclone separators and an
internal downcomer, is used for power removal. Control may be effected
by varying the volume of slurry or heavy water in the core,2?

In the Boiling Slurry Reactor Experiment (BSRE), or SLURREX, of the

Argonne National Laboratory, slurry is circulated, and power is removed by

149

boiling. The purpose was to determine feasibility and to gain information
on operating problems.

Slurry flows up through a 24 in. pipe that defines the core region and
returns through downcomers to the core inlet. The slurry boils in the core,
and steam is separated in a sausage-shaped section at the core outlet,. 39
This reactor can be a burner—=converter using the U-HZO—Th or U—HZO-Pu fuel
cycles, or a breeder using the U—DZO—Th cycle. It can be controlled by
either adjusting the slurry concentration or varying the recirculation
velocity.3! Between April and December 1960, an initial hazards analysis
had shown that SLURREX could be operated safely near EBR-II at the National
Reactor Testing Station in Idaho. Also, because the performance of boiling
reactors depends largely on the hydrodynamic characteristics of the reactor
system, an atmospheric-pressure mockup of the vessel had been built with
fabrication and erection essentially complete.32

A French 500 to 1000 kw experimental prototype, PHOEBUS (Pilot Homogeneous
Reactor with Ebullition and Suspension), is comparable to SLURREX in that
both use a suspension containing uranium dispersed in light or heavy water
with heat extracted by boiling in the reactor. In PHOEBUS a cyclone
device is proposed for accelerating this heat extraction, whereby the
bubbles of vapor move toward the vortex where they are collected. Liquid
fuel is injected tangentially at the periphery of the core and dispersed.

The fuel could be either a solution or a suspension, although at present

the use of a suspension in water having a pH of about 7 is being considered.
Many hydraulic problems are involved, such as stability of the vortex

shape and proportions, the mechanism for forming the bubbles, and the
fluctuations in spatial concentration of the final suspensions. A 500-kw
electrical model has answered some of these questions. PHOEBUS itself is
conceived as a reactor physics experiment. Light water will be used for
moderator, and the fuel will be suspensions of fine glass spheres of

uranium, or possibly particles of uranium dioxide fused at a high temperature°33

Two-Region Slurry-Core Reactors

A two-region reactor with boiling slurry in both the core and breeding
blanket has been briefly discussed in an ORSORT report.2L+ The fuel-moderator-
coolant is a slurry of uranium trioxide in heavy water, with thorium dioxide

slurry in the blanket. The reactor produces 66 Mw(e).

150

A Russian experimental power reactor located at Ulyanovsk, near the
Volga, uses uranium oxides suspended in heavy water as fuel-moderator-coolant,
and a thorium suspension as fertile material. The power level is 25-35
Mw(t)3"% and the coolant pressure 700 psia.3%»3% 1In related Russian concepts
for boiling homogeneous reactors, a boiling solution is given as an
alternative; and a reflector filled with a boiling suspension of thorium

in heavy water, with an external circulating system, was suggested,%s?’7

Miscellaneous

A British patent describes a reactor in which heat is removed from the
critical region by using the steam bubbles formed by beoiling in the
critical region, which has many upwardly extending tubes of small diameter,
to circulate the liquid fuel through a heat exchanger.38 Such fuels as
slurries or sclutions-~-enriched uranyl sulfate in ordinary water, for
example--might be used. The length and diameter of the tubes distinguish

this invention from previous circulating systems.
Status

As with the non-boiling homogeneous aqueous reactors, the future of
boiling reactors is uncertain. Few concepts have gone beyond the design
stage, and the early interest does not appear to have led to industrial
development, in the U.S. at least. Some investigations have been made of
comparative costs. In 1953, Briggs reported that differences in costs

between boiling and circulating reactors were small,3?

151
DATA SHEETS

BOILING REACTORS
No. 1 Boiler-Converter (Tray Pile)

Manhattan Project

Reference: U. S. Patent 3,070,529.

Originator: E. P. Wigner.

Status: Proposal, 1944,
Details: Thermal neutrons, steady-state, converter. Fuel: solution of
plutonium salt in H20° Solution fills shallow trays placed 50 cm above
each other. Trays preferably not of neutron-absorbing material, Area of
trays: approximately 1 m2. If 60 trays are used, system becomes critical
and solution boils., Steam is conducted to condensers and water flows back
to trays. Reflector: thorium. Pressure: 10 atm suggested to reduce
volume of steam. Advantages: simplicity, no outside holdup, elimination
of poisoning fission products, very slight change in critical size caused
by bubbles; usability of power. Problems: tray material, plutonium
entrainment, solution foaming, possible solution instability. Power:
400 Mw(t).
Code: 0311 13 32201 46 624 726 84677 931 109

8111X

155
No. 2 Teapot
ORNL

References: CF-52-5-226; CF-52-6-78.
Originators: ORNL staff.

Status: Conceptual stage, 1952; abandoned, 1953.
Details: Thermal neutrons, steady state, burner. Fuel: solution of
enriched uranyl sulfate (93.4% U235) in HZO; 59 g U235 per liter of solution,
Pressure: wup to 150 psi. Reactor vessel: stainless-steel cylinder, 6 ft
long, 17-1/2 in. diameter, 1/4 in. thick walls, with 18-~in. dished heads.
Steam leaves reactor and is cooled by internal and external condensers;
condensate returns by gravity. Control: Cd or B10 rod. Vertical
thimble in center facilitates: inserting source for start-up; inserting
control rod; kinetic experiments in which rod would be quickly removed;
and measuring vertical flux distribution. Maximum power: 250 kw(t).
Code: 0313 13 32201 44 624 711 81111 921 109

81112

84677

No. 3 Homogeneous Boiling Reactor Package Power Plant (HBR)

ORSORT

Reference: CF-53-10-23,
Originators: R. J. Rickert, W. C. Gribble, B. A. Mong, R. G. Stater, and
J. J. Wesbecher.,

Status: Preliminary design; term paper, Aug. 21, 1953,

Details: Thermal neutrons, steady state, burner. Fuel: 93.4%7 enriched
U235 as uranyl sulfate. Moderator: HZO' Core temperature: 482°F,
Pressure: 580 psi. Reactor: 6~ft hemisphere lined with Ti sheet. No
reflector, but some reflection provided by tank. Control: adjusting fuel
level in core to meet power‘requirements. Power: 10 Mw(t).

Code: 0313 13 32201 44 627 711 83679 921 101

156
No. 4 Boiling Homogeneous Power and Plutonium Producer

ORSORT

Reference: CF-54-8-238 Del.
Originators: H. J. Kamack, P. J. Flickinger, F. C. Haag, P. B. Haga,

R. G. McGrath, B. R, Moskowitz, D. E. Murphy, and R. B. Nicholson,
Status: Preliminary design; term paper, August 1954,

Details: Thermal neutrons, 'unsteady-state behavior," converter. Fuel-

moderator-coolant: U02804 solution or UO3 slurry in boi%égg DZO with

Pu oxide as a precipitate. Fuel enrichment: about 17 U . Core vessel:
Ti-lined, carbon steel circular cylinder, 20-ft diameter, 20 ft high.
Single~flux region is operated at 1000-1500 psi and 250°C. To correct

"unsteady-state behavior,'" authors suggest a reflector, operation at

higher pressure, and use of HZO instead of D20° Fuel flow: by natural
circulation; no pumps are necessary. In the primary power-removal systen,
the steam-liquid mixture from the reactor flows to multiple external
vapor separators, from which the liquid returns to the core by gravity
through external downcomers. Steam goes to heat exchangers and the
condensate circulates through the thermal shield as it returns to the

core by gravity. Control: by moderator holdup. The reactor is designed
as a power producer--1200 Mw(t), 346 Mw(e)--with plutonium as by-product.
Code: 03X1 14 32213 42 625 743 84677 921 101

32313 635 753

157
No. 5 Aqueous Homogeneous Boiling Reactor Power Plant

Babcock and Wilcox Co.

Reference: BW-AED-502,
Originators: B. A. Mong, J. E. Colgan (Vitro Corp. of America), R. A.

D'Elia, J. S. Mooradian, G, K. Rhode, and P. M. Wood.

Status: Design study, June 1, 1955,

Details: Thermal neutrons, steady state, burner. Fuel-moderator-coolants:
solution of 93% enriched U235 as UOZSO4 in HZO' Solution contained in a
27 in, diameter Ti liner in a 30 in. ID cylindrical pressure vessel of
stainless steel with an elliptical head at one end. Full-power operating
temperature: 250°C at 600 psia. Steam generated from the boiling
solution passes through an entrainment separator to be condensed on a
bayonet tube~type heat exchanger in the top section of the reactor

vessel, Steam from the bayonet tubes passes through a turbine to generate
electricity. Control: mainly by power demand. Design is for a small
power plant for possible remote locations. Power: 1300 kw(t), 100 kw(e),
plus 400 kw space heating.

Lode: 0313 13 32201 44 624 71 84677 921 101

8XXXX

158
No. 6 Large-Scale Boiling Slurry Plutonium Power Producer

ORSORT

Reference: CF-51-8-84 Rev.
Originators: G. H. Cohen, K. W. Downes, F. R. Grisak, J. H. Hill,

V. P, Kovacik, D. G, Ott, and L. C. Widdoes.

Status: Preliminary design and feasibility study; term paper, August 1951,

Details: Thermal neutrons, steady state, breeder. Fuel-moderator-coolant:

boiling slurry of UO3 (0.81%Z enriched) in DZO' Ratio of moderator to

uranium: 50, Operating temperature: 313°C. Operating pressure:

1487 psi. Reactor vessel: stainless-steel spherical pressure vessel,

30 ft diameter. Reactive volume: lower half of sphere. Core volume:

200,159 liters. Saturated DZO vapor goes from reactor to turbines or

through chemical-fired equipment. Condensed D20 is pumped directly back

to reactor. Shielding: thermal shield--hemisphere of boron-containing

metal in lower hemisphere; shield in upper hemisphere--over-lapping

baffles to let steam pass. Control: control of slurry concentration

and volume; emergency dumping of slurry; xenon poisoning. Power: 805 Mw(t).

Code: 0312 14 32302 42 635 753 83779 921 101
84677

81596

159
No. 7 Boiling Homogeneous Reactor
ORSORT

Reference: CF-54-8-240.
Originators: H. R, Zeitlin, H. W. Bertini, W. F. Bourgeois, J. N. Calvin,

J. R. Engel, G, H. Farbman, O, R. Meyer, and D. P. Ross.

Status: Preliminary design; term paper, August 1954,

Details: Thermal neutrons, steady state, one-region breeder, Fuel-
233) in boiling D

moderator-coolant: ThO -UO2 (U 0 slurry. Operating

pressure: 1000 psi (dezign 1350 psi). Natural-iirculation type.
Temperature: 545°F, Core: cylindrical, 17.5 ft diameter, 17.5 ft high,
Pressure vessel: hemispherical-headed cylinder, 20.5 ft 0D, 50 ft high.
Steam-dumping condenser, Control: relief and control valves. Power:
500 Mw(t).

Code: 0312 14 32302 45 635 756 84677 921 101

No. 8 Boiling Slurry Reactor Experiment (SLURREX)

ANL

References: ANL-6148; ANL-6248.
Originators: J. F. Marchaterre and M. Petrick.

Status: Terminal report, December 1960, Criticality expected 20 months
after Title I design is approved.
Details: Thermal neutrons, steady state, converter. Fuel: Th02°UO205
aqueous slurry, 937 enriched U235° Coolant: demineralized H20° Natural-
circulation boiling reactor. Design pressure: 13,6 atm; temperature:
205°C (operating--10.2 atm and 181°C). Reflector: H20° Control:
cruciform boron - stainless-steel control rod moving vertically; gas
injection into downcomer; change in fuel concentration., Power: 5 Mw(t).
Fabrication and erection of the half-scale mockup were essentially
complete at the termination of the SLURREX project.
Code: 0311 13 32301 44 634 756 81111 921 101

81596

83779

160
No. 9 Aqueous Homogeneous Reactor Plant with Boiling Slurry Blanket

ORSORT

Reference: CF-54-8-238, p. 146.
Originators: H. J. Kamack, P, J. Flickinger, F. C. Haag, P. B. Haga,

R. G. McGrath, B. R, Moskowitz, D. E. Murphy, and R. B. Nicholson.
Status: 'Leading Data" given as appendix to "Boiling Homogeneous Power
and Plutonium Producer," term paper, August 1954.

Details: Thermal neutrons, steady state, evidently a breeder. Fuel-
moderator-coolant: slurry of U03 in D,0. Fertile material: ThO, in

2 2
D20 as boiling slurry blanket., Pressure: 1850 psia; inlet temperature
in core: 505°F, outlet: 580°F. Breeding ratio: 1.12. Power:
200 Mw(t), 66 Mw(e).

Code: 0312 14 3X302 45 625 756 8XXXX 9X 101

161
No. 10 Steam Stirred Homogeneous Nuclear Reactor

Reference: British Patent No. 840,740,

Originators: U, S. Atomic Energy Commission.

Status: Patent granted, July 13, 1960.

Details: Thermal neutrons, steady state, burner. Fuel: enriched uranyl
nitrate, sulfate, or other salt in HZO or DZO’ or p%g;onium. Example

2O, 90% enriched in U

900 g; operating mass: 2 kg. Reflector: 60-in,

» Moderator: H,O,

uses 0,5 M uranyl sulfate in H 2

Critical mass U235:
graphite cube, Reactor housing: three regions: <critical, heat-exchanger,
and gas-recombination. All are inside larger cylinder. Critical region
is spherical shell of stainless steel, 1/4 in. thick. It is perforated
at bottom so that solution may enter, A bundle of small tubes welded
to top of reactor shell connects it to heat exchanger. Tubes are of
such a length that most steam bubbles form in tube rather than critical
region. The fuel level is just below the top of the tubes. The nuclear
reaction boils the solution, which forms steam bubbles. Multiple tubes
minimize bubble surges. At optimum circulation and at 100°C operating
temperature, tubes should be at least as long as the critical region. For
a critical-region diameter and a tube length of 25 cm each, the optimum
tube diameter is about 1 cm. Bubbles move upward into the tubes, forcing
out the fuel solution that fills them., This solution circulates through
heat exchangers, and back to the reaction vessel, where it is again
heated, Heat exchanger: series of flat spaced stalnless-steel coils
forming a cylindrical bundle. Water vapor and radiolytic gases pass
through gas-recombination systems and cooling channels. Condensed water
returns to the reactor. Control: negative temperature coefficient;
control of fuel concentration, B or Cd sleeve; vertical control rod.
Code: 0313 13 32201 44 624 711 84677 921 101
14 32202 625 83679

81141

81142

8111X

162
11.

12,

13.

14,

15.

References

J. A, Lane, H. G, MacPherson, and Frank Maslan, eds.,, Fluid Fuel

Reactors, Addison-Wesley Publ. Co., Reading, Mass., 1958, pp. 21-23.
E. P. Wigner, "Neutronic Reactor," U. S. Patent 3,070,529, December
25, 1962. Filed 1944.

A. W. Kramer, Boiling Water Reactors, Addison-Wesley Publ. Co.,
Reading, Mass., 1958, p. 3l.

R. N. Lyon, "Preliminary Report on the 1953 Los Alamos Boiling Reactor
Experiments,'" CF-53-11-210, ORNL, Nov. 30, 1953. Decl. Feb. 14, 1957,
(See also ORNL-1221, p. 133 ff,)

R. V. Bailey, "Progress Report on Power Removal from Homogeneous
Boiling Reactors," CF-53-11-65, ORNL, Nov. 4, 1953, Decl. April 8, 1957,
W, E., Thompson, comp., "Homogeneous Reactor Project. Quarterly
Progress Report for Period Ending October 31, 1953," ORNL-1653, ORNL,
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J. A. Lane, "The Bubble Problem in the Homogeneous Reactor,"
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W. E. Thompson, comp., '"Homogeneous Reactor Project. Quarterly
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Ibid., for period ending May 15, 1951, ORNL-1057, ORNL, Oct. 10, 1951,
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Ibid., for period ending November 15, 1951, ORNL-1221, ORNL, April 1,
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Ibid., for period ending July 1, 1952, ORNL-1318, ORNL, Sept. 19, 1952,
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Ibid., for period ending October 1, 1952, ORNL-1424 Del., ORNL,

pp. 13, 20.

J. M. Stein and P. R. Kasten, '"Boiling Reactors; A Preliminary
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P. R, Kasten to J. A. Swartout, "Boiling Reactors--Variation of
Reactor Stability with Operating Power Level and Average Bubble Size,"
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P. R. Kasten, "Reactor Physics of TEAPOT," Pt. I, CF-52-6-78, ORNL,
June 5, 19523 Pt. II, CF-52-6-118, ORNL, June 20, 1952; Pt. III,
CF=-52-7-38, ORNL, July 16, 1952; Pt. IV, CF-52-8-128, ORNL, Aug. 15,
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Ref., 9, pp. 31-32,

Ref. 1, pp. 8, 21=22,

W. E. Thompson, comp., "Homogeneous Reactor Project. Quarterly
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Reactor Power Package for Use in Remote Locations," CF-53-10-23,
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H, J. Kamack, P. J., Flickinger, F. C. Haag, P, B, Haga, R. G. McGrath,
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B. A, Mong, J. E., Colgan, R. A, D'Elia, J. S. Mooradian, G. K. Rhode,
and P, M. Wood, "A Design Study of a Low Power Aqueous Homogeneous
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Kesselwerke AG., "Homogener Siedereaktor, bei dem die
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164

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165