FFR_chap01.txt

o
<

copo-xlgimrhw_tof—‘

Part 1

AQUEOUS HOMOGENEOUS
REACTORS

James A. Lang, Editor
Oak Ridge National Laboratory

Homogencous Reactors and Their Development
Nuclear Characteristies of One- and Two-Region Homogencous Reactors

. Properties of Aqueous Fuel Solutions
. Technology of Aqueous Suspensions
. Integrity of Metals in Homogencous Reactor Media

Chemical Processing

. Design and Construction of Experimental Homogeneous Reactors

Component Development

. Large-Scale Homogeneous Reactor Studies

Homogencous Reactor Cost Studies

AUTHORS
E. G. BouLMANN H. F. McDurrie
P. R. KAsTEN R. A. McNEEs
J. A. LANE C. L. SEGASER
J. P. McBripE 1. SPIEwAK
D. G. Taomas
CONTRIBUTORS
B. M. ApaMmsoN S. I. KarrLan
S. E. BeaLn N. A. Kroun
W. II. BrownNiNG C. G. LawsonN
W. D. BurcH R. E. LEuze
R. D. CHEVERTON R. N. Lyon
E. L. CoMPERE W. T. McDuFrrEE
(. H. GABBARD 1.. E. MoRSE
J. C. GRIEss S. PETERSON
D. B. HaLL R. C. ROBERTSON

E. C. HisE H. C. SAVAGE
G. H. JeNKs D. 8. Tooms
J. C. WILsON
PREFACE

This compilation of information related to aqueous homogeneous
reactors summarizes the results of more than ten years of research ana
development by Oak Ridge National Laboratory and other organizations.
Some 1500 technical man-years of effort have been devoted to this work,
the cost of which totals more than $50 million. A summary of a program
of this magnitude must necessarily be devoted primarily to the main
technical approaches pursued, with less attention to alternate approaches.
I'or more complete coverage, the reader is directed to the selected bib-
liography at the end of Part I.

Although research in other countries has contributed to the technology
of aqueous homogeneous reactors, this review is limited to work in the
United States. In a few instances, however, data and references pertaining
to work carried on outside the United States are included for continuity.

Responsihility for the preparation of Part I was shared by the members
of the Oak Ridge National Laboratory as given on the preceding page and
at the beginning of each chapter.

Review of the manuseript by others of the Oak Ridge Laboratory staff
and by scientists and engineers of Argonne National Laboratory and
Westinghouse Electric Corporation have improved clarity and accuracy.
Suggestions by R. B. Briggs, director of the Homogeneous Reactor Project
at the Oak Ridge Laboratory, and S. McLain, consultant to the Argonne
Laboratory, were particularly helpful.

Others at Oak Ridge who assisted in the preparation of this part include
W. D. Reel, who checked all chapters for style and consistency, W. C.
Colwell, who was in charge of the execution of the drawings, and H. B.
Whetsel, who prepared the subject index.

Oak Ridge, Tennessee James A. Lane, Edqtor
June 1958
CHAPTER 1
HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT*

1-1. BACKGROUNDT

1-1.1 Work prior to the Manhattan Project. Nuclear reactors fueled
with a solution or homogeneous mixture of fuel and moderator were among
the first nuclear systems to be investigated experimentally following the
discovery of uranium fission. In fact, it was only slightly more than a
yvear after this discovery that Halban and Kowarski at the Cavendish
Laboratory in England performed experiments which indicated to them
that a successful self-sustaining chain reaction could be achieved with a
slurry of uranium oxide (U30Os) in heavy water.

In these experiments, reported in December 1940 [1], 112 liters of heavy
water mixed with varying amounts of UzOg powder were used inside an
aluminum sphere 60 em in diameter, which was immersed in about one ton
of heavy mineral oil to serve as a reflector. (Mineral oil was chosen to
avold contamination of the D20 in case of a leak in the sphere.) By meas-
uring neutron fluxes at varying distances from a neutron source located in
the center of the sphere, Halban and Kowarski calculated a multiplication
factor of 1.18 4 0.07 for this system when the ratio of deuterium atoms to
uranium atoms was 380 to 1, and 1.09 4+ 0.03 when the D/U ratio was
160 to 1.

Other experiments conducted at the same time by Halban and Kowar-
ski [1]1, using U30g and paraffin wax, indicated that with a heterogeneous
lattice arrangement it would be possible to achieve multiplication factors
as high as 1.37 in a system containing about 100 atoms of deuterium per
atom of uranium.

It is interesting to note that the D2O supply used in the experiments
had been evacuated from France. The D20 originally came from the lab-
oratories of the Norwegian Hydroelectric Company, and with the destruc-
tion of this plant and its D20 stockpile in 1942, this was the sole remaining
supply of purified D20O. However, it was not enough to allow a self-
sustaining chain reaction to be established with natural uranium.

*By J. A, Lane, Oak Ridge National Laboratory.

1This section is based on material supplied by W. E. Thompson, Oak Ridge
National Laboratory.

1See the list of references at the end of the chapter.

1
2 HOMOGENEOQUS REACTORS AND THEIR DEVELOPMENT [cHAP, 1

Even earlier (in 1939) Halban and Kowarski, as well as other experi-
mentalists, had fairly well established that self-sustaining chain reactions
with U30g and ordinary water are not possible [2,3,4]. Homogeneous sys-
tems of uranium with carbon, helium, beryllium, or oxygen were also con-
sidered, and were rejected as not feasible either for nuclear, chemical, or
engineering reasons,

In November 1942, Kowarski, with Fenning and Seligman, reported
more refined experiments which led to the conclusion that neither homo-
geneous nor heterogeneous mixtures of UzOg with ordinary water would
lead to self-sustaining chain reactions, the highest values of the multiplica-
tion factor being 0.79 for the homogeneous system and 0.85 for the hetero-
geneous system.

Because it was clear even by early 1942 that the only feasible homo-
geneous reactor using natural uranium would be one moderated with D»0,
and because no D20 was available at that time for use in reactors, interest
in homogeneous reactor systems was purely academic. The atomic energy
program, which was then getting well under way, devoted its attention to
heterogeneous reactors. By using a heterogeneous lattice arrangement
with a core of uranium metal slugs spaced inside graphite blocks and a
periphery containing UzOg slugs (used after the supply of uranium metal
ran out) spaced inside the graphite, the first successful self-sustaining chain
reaction was achieved on December 2, 1942.

1-1.2 Early homogeneous reactor development programs at Columbia
and Chicago universities. Interest in homogeneous reactors lagged until
early in 1943, when it became clear that American and Canadian efforts to
produce large quantities of heavy water would be successful. At that time
the group under H. C. Urey at Columbia University directed its attention
to the development of slurried reactors utilizing uranium oxide and D20,

In March 1943, Urey and Fermi held a conference to review the situa-
tion with respect to homogeneous reactors. They noted the value of 1.18
that Halban and Kowarski had obtained for the multiplication factor in a
U30s-D20 slurry reactor and pointed out that the value caleulated from
theory was only 1.02. They realized, however, that neither the theory nor
the experiment was free from serious objections, and that insufficient data
were avallable to allow a trustworthy conclusion to be reached as to the
feasibility of homogeneous systems.

If the results of Halban and Kowarskl were correct, then a homogeneous
system containing a few tons of heavy water would be chain reacting. On
the other hand, if the theoretical estimates were correct, the order of
100 tons of D20 would be required.

Urey and Fermi recommended [5] that the earlier U3zOs-D20 experi-
ments be repeated with the improved techniques then known, and that
1-1] BACKGROUND 3

consideration be given to incorporating a mixture of uranium and heavy
water into the pile at Chicago to determine its effect on the pile reactivity.

From the theoretical considerations of E. P. Wigner and others, it ap-
peared that the most favorable arrangement for a UzOg-D20 reactor
would be one in which the slurry was pumped through a lattice of tubes
immersed in D20 moderator. This was especially true because the neutron
absorption cross section assigned to heavy water at that time made it ap-
pear that more than 200 tons of D20 would be required to reach criticality
in an entirely homogeneous system in which the UzOg and moderator were
mixed. With a heterogeneous system it seemed likely that a much smaller
quantity of D20 would suffice and every effort was directed toward pre-
paring a design that would require about 50 tons of DO [6].

It was estimated by E. P. Wigner that the uranium concentration in the
slurry would have to be 2.5 to 3 grams per cubic centimeter of slurry. It
became apparent immediately that no aqueous solution of a uranium com-
pound could be made with such a density, With pure Ul'g, 2.48 grams of
uranium per cubic centimeter could be obtained, and piles utilizing this
compound were considered. However, the corrosion problems in such a
system were believed to be so severe that the development of a reactor to
operate at a high power level would be extremely difficult, if not impossible.

Other compounds, such as uranyl nitrate dissolved in D20, were ex-
cluded because in the case of nitrate the neutron absorption of nitrogen
was too high and in other cases sufficient densities could not be obtained.
Thus the initial phase of the research at Columbia was directed toward the
development of high-density slurries [6].

The reactor visualized by the Columbia group was one in which an ex-
tremely dense suspension of uranium in D20 would be pumped through a
large number of pipes arranged inside a heavy-water moderator. It was
planned that both the slurry and the moderator would be circulated
through heat exchangers fer cooling [6].

Then, in July of 1943, the experiments of Langsdorf [7] were completed,
giving a much lower cross section for deutertum than was known earlier.
As a result, the homogencous reactor became much more attractive, since
the eritical size (neglecting external holdup} could then be reduced to about
30 tons of DO with about 6 tons of uranium as oxide in an unreflected
sphere [8]. This favorable development allowed emphasis to be shifted to
less dense slurries, greatly simplifying the problems of maintaining a sus-
pension of dense slurry, pumping it, and protecting against erosion. Ex-
periments were directed toward developing a reactor design which would
permit operation without continuous processing of the slurry to maintain
its density [6].

By the end of 1943 preliminary designs had been developed at the
University of Chicago Metallurgical Laboratory for several types of heavy-
4 HOMOGENEQUS REACTORS AND THEIR DEVELOPMENT  [cHAP. 1

water reactors, all using slurry fuel but differing in that one was com-
pletely homogeneous [9], one was a light-water-cooled heterogeneous ar-
rangement [10], and another was a D20-cooled heterogeneous reactor [11].
These reactors were proposed for operation at power levels of 500 Mw or
more (depending on external power-removal systems) and were intended
as alternates to the Hanford piles for plutonium production in case satis-
factory operation of the graphite-natural uranium, water-cooled piles
could not be achieved.

At this point one might ask why it was that homogeneous solution
reactors were not given more serious consideration, especially in view of
the newly discovered cross section for deuterium, which permitted con-
siderably lower concentrations of uranium. The answer is that the only
known soluble salts of uranium which had a sufficiently low cross section to
enable the design of a reactor of feasible size and D20 requirement were
uranyl fluoride and uranium hexafluoride. (Enriched uranium was not
then available.) These were considered, but rejected principally because
of corrosion and instability under radiation. A second factor was the evi-
dence that D20 decomposition would be more severe in a solution reactor
where fission fragments would be formed in intimate contact with the
D20 rather than inside a solid particle as in the case of a slurry.

Research on homogeneous reactors was undertaken at Columbia Uni-
versity in May 1943, and continued with diminishing emphasis until the
end of 1943, at which time most of the members of the homogeneous re-
actor group were transferred to Chicago, where they continued their work
under the Metallurgical Laboratory.

At the Metallurgical Laboratory, the principal motivation of interest in
homogeneous reactors was to develop alternate plutonium production
facilities to be used in the event that the Hanford reactors did not operate
successfully on a suitable large scale, and studies were continued through
1944. With the successful operation of the Hanford reactors, however,
interest in homogeneous plutonium producers diminished, and by the end
of 1944 very nearly all developmental research had been discontinued. The
results of this work are summarized in a book by Kirschenbaum [12].

1-1.3 The first homogeneous reactors and the Los Alamos program.
During the summer of 1943 a group at Los Alamos, under the leadership
of D. W. Kerst, designed a ‘‘power-boiler’” homogeneous reactor, having
as its fuel a uranyl sulfate-water solution utilizing the enriched uranium
which was expected to become available from the electromagnetic process.
However, this design was put aside in favor of a low-power homogeneous
reactor designed by R. F. Christy. The low-power homogenecous reactor
was built and used during the spring and summer of 1944 for the first of a
series of integral experiments with enriched material (see Chapter 7).
1-1] BACKGROUND 5

There were two reasons for choosing UO2804 instead of uranyl nitrate
as the fuel: there is less neutron absorption in the sulfate than in the ni-
trate, and the sulfate was thought to be more soluble. The latter reason
was considered important because it was feared that with the maximum-
enrichment material from the electromagnetic process, 1t might be dificult
to dissolve the critical mass in the desired volume [13]. These objections
to the use of uranyl nitrate, however, were subsequently found to be
mvalid.

After gaining experience in operating the low-power reactor, “LOPO,”
the Los Alamos group revised its plans for the higher power homoge-
neous reactor, known as the “"HYPO,” and after extensive modification
of the design, the reactor was built and put into operation in December 1944
with uranyl nitrate as the fuel.

In April 1949, rather extensive alterations to the HYPO were begun in
order to make the reactor a more useful and safer experimental tool. The’
modified reactor, known as “SUPO,” is still in operation. The present
SUPO model reached local boiling during imitial tests, due to the high
power density. A slight increase in power density above the design level
produces local boiling between cooling coils, even though the average so-
Iution temperature does not exceed 85°C.

Interest in solution reactors continued at Los Alamos, and improved
designs of the Water Boiler (SUPO Model 1) were proposed [14]. These,
however, have not yet been constructed at Los Alamos, although similar
designs have been built for various universities [15].

The work on water boilers at Los Alamos led to the design of power
reactor versions as possible package power reactors for remote locations.
Construction of these reactors, known as Los Alamos Power Reactor Fx-
periments No. 1 and No. 2 (LAPRE~1 and LAPRE-2), started in early
1955. To achieve high-temperature operation at relatively low pressures,
LAPRI-1 and =2 were fueled with solutions of enriched uranium oxide in
concentrated phosphoric acid. The first experiment reached criticality in
March 1956 and was operated at 20 kw for about 5 hr. At that time
radioactivity was noted in the steam system, and the reactor was shut
down and dismantled. It was discovered that the gold plating on the
stainless steel cooling coils had been damaged during assembly and the
phosphoric acid fuel solution had corroded through the stainless steel.
The cooling coils were replaced and operations were resumed in October
1956. However, similar corrosion difficulties were encountered, and it was
decided to discontinue operations. In the meantime, work on LAPRE-2
continued, and construction of the reactor and its facilities was completed
during the early part of 1958. The details of these reactors are given In
Chapter 7.
6 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT [crap. 1

1-1.4 Early homogeneous reactor development at Clinton Laboratories
(now Oak Ridge National Laboratory). With the availability of enriched
uranium in 1944, the possibility of constructing a homogeneous reactor
became more attractive because, by using enriched uranium, the D.0
requirement could be greatly reduced, or even ordinary water could be
used. The chemists at Clinton Laboratories (now ORNL), notably C. D.
Coryell, A. Turkevich, S. G. English, and H. S. Brown, became interested
in enrichéd-uranium homogeneous reactors primarily as a facility for pro-
ducing other radioisotopes in larger amounts, and a number of reports on
the subject were issued by various members of the Chemistry Division
(D. E. Koshland, Jr., W. J. Knox, and L. B. Werner).

In August 1944 Coryell and Turkevich prepared a memorandum [16]
recommending the construction of a 50-kw homogeneous reactor containing
5 kg of uranium enriched to 12497 U%35 or about 500 g of plutonium. The
fuel proposed was to be in the form of salt solution in ordinary water. The
following valuable uses of such a reactor were listed in this memorandum
and enlarged upon in a later memorandum by Coryell and Brown [17]:

(1) The preparation of large quantities of radioactive tracers,

(2) The preparation of intense radioactive sources.

(3) Studies in the preparation and extraction of U233,

(4) The preparation of active material for Hanford process research.

(5) Study of chemical radiation effects at high power levels.

(6) Accumulation of data on the operating characteristics, chemical
stability, and general feasibility of homogeneous reactors.

The physicists were also interested in the homogeneous reactor, partic-
ularly as a research facility which would provide a high neutron flux for
various experimental uses. The desirability of studying, or demonstrating,
if possible, the process of breeding had been made especially attractive
by the recent data indicating that U233 emitted more neutrons for each
one absorbed than either U235 or Pu®*, and the physicists were quick to
point out the possibility of establishing a U2%3-thorium breeding cycle
which would create more U??3 fro:i the thorium than was consumed in the
reactor. These potentialities were very convineingly presented in No-
vember 1944 by L. W. Nordheim in a report entitled “The Case for an
Enriched Pile” (ORNL-CF-44-11-236).

The power output of such a breeder with a three-year doubling time is
about 10,000 kw, and this was established as a new goal for the homoge-
neous reactor. The reactor, then, was conceived to be a prototype homo-
geneous reactor and thermal breeder; in addition, it was conceived as an
all-purpose experimental tool with a neutron flux higher than any other
reactor.

Work on the 10,000-kw homogeneous reactor was pursued vigorously
through 1945; however, at the end of that year there were still several
1-1] BACKGROUND 7

basic problems which had not been solved. Perhaps the most serious of
these was the formation of bubbles in the homogeneous solution. These
bubbles appear as a result of the decomposition of water into hydrogen and
oxygen by fission fragments and other energetic particles. Because the
bubbles cause fluctuations in the density of the fuel solution, they make it
difficult to control the operating level of the reactor. Nuclear physics
calculations made uat the time indicated that under certain conditions it
might be possible to set up a power oscillation which, instead of being
damped, would get larger with ecach cycle until the reactor went completely
out of control. Minimizing the bubble problem by operating at elevated
temperature and pressure was not considered seriously for two reasons:
first, beryllium, aluminum, and lead were the only possible tank materials
then known to have sufficiently low neutron-absorption characteristics to
be useful in a breeder reactor. Of these metals, only lead was acceptable
because of corrosion, and lead is not strong enough to sustain elevated
temperatures and high pressures. Second, there had been essentially no
previous experience in handling highly radioactive materials under pres-
sure, and consequently the idea of constructing a completely new type of
reactor to operate under high pressure was not considered attractive.

Other major unsolved problems at the end of 1945 were those of corro-
sion, solution stability, and large external holdup of fissionable material.
Because 1t appeared that the solution of these problems would require
extensive research and development at higher neutron fluxes than were
then available, it was decided to return to the earlier idea of a hetero-
geneous reactor proposed by E. P. Wigner and his associates at the Metal-
lurgical Laboratory. Experimental investigations in this reactor, it was
hoped, would yield data which would enable the homogeneous reactor
problems to be solved. The extensive effort on this latter reactor (later
built as the Materials Testing Reactor in Idaho) forced a temporary
cessation of design and development activities related to homogeneous
breeder reactors, although basic research on aqueous uranium systems
continued.

1-1.5 The homogeneous reactor program at the Oak Ridge National
Laboratory. FEarly in 1949, A. M. Weinberg, Research Director of Oak
Ridge National Laboratory, proposed that the over-all situation with
respect to homogeneous reactors be reviewed and their feasibility be
re-evaluated in the light of knowledge and experience gained since the
end of 1945. Dr. Weinberg informally suggested to a few chemists, physi-
cists, and engineers that they reconsider the prospects for homogeneous
reactors and hold a series of meetings to discuss their findings.

At the meeting held by this group during the month of March 1949, it
was agreed that the outicok for homogeneous reactors was considerably
8 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [CHAP. 1

brighter than in 1945 and that effort directed toward the design of a small
experimental reactor should be resumed. By July 1949, interest in homo-
geneous reactors had increased further as a result of the preliminary studies
which had been started, and it was decided to establish a small develop-
ment effort on homogeneous reactors. A Homogeneous Reactor Com-
mittee, under the direction of C. E. Winters, was formed and reactor
physics and design studies were undertaken on a somewhat expanded
scale. By the latter part of August 1949, a preliminary design of the major
components had been developed.

Construction of the reactor (Homogeneous Reactor Experiment No. 1)
was started in September 1950, and completed in January 1952. After a
period of nonnuclear testing with a natural-uranium fuel solution, HRE~-1
reached criticality on April 15, 1952. Early in 1954 it was dismantled
after successfully demonstrating the nuclear and chemical stability of a
moderately high-power-density circulating-fuel reactor, fueled with a
solution of enriched uranyl sulfate.

During the period of construction and operation of HRE-1, conceptual
design studies were completed for a boiling reactor experiment (BRE)
operating at 150 kw of heat and a 58-Mw (heat) intermediate-scale homo-
geneous reactor (ISHR). Further work on these reactors was deferred
late in 1953, however, when it became evident from HRE~-1 and the asso-
ciated development program that construction of a second homogeneous
reactor experiment would be a more suitable course of action.

The main reason for this decision was that HRE-1 did not demonstrate
all the engineering features of a homogeneous reactor required for con-
tinuous operation of a nuclear power plant. Thus a second experimental
reactor (Homogeneous Reactor Test, HRE-2), also fueled with uranyl
sulfate, was constructed on the HRE-1 site to test the reliability of ma-
terials and equipment for long-term continuous operation of a homo-
geneous reactor, remote-maintenance procedures, and methods for the
continuous removal of fission products and insoluble corrosion products.
Construction of the reactor was completed late in 1956 and was followed
by a period of nonnuclear operation to determine the engineering charac-
teristics of the reactor. This testing program was interrupted for six to
nine months by the need for replacing flanges and leak-detection tubing
in which small cracks had developed, owing to stress corrosion induced
by chloride contamination of the tubing. The reactor was brought to
criticality on December 27, 1957, and reached full-power operation at
5 Mw on April 4, 1958. Shortly thereafter, a crack in the core tank de-
veloped which permitted fuel solution to leak into the D3O blanket.
After consideration of the nuclear behavior of the reactor with fuel in both
the core and blanket, operation was resumed under these conditions in
May 1958.
1-1] BACKGROUND 9

TasLe 1-1

L.eveLs or Errort oN HoMoGENEOUS
Reacror DeveLopMENT AT ORNL

 

 

 

i Millions Man-years

Fiscal year of dollars (technical}
1949 0.15 5
1950 0.54 15
1951 2.2 75
1952 4.1 127
1953 3.4 119
1954 3.9 133
1955 7.7 219
1956 9.1 238
1957 10.0 316
1958 11.5 333

 

 

 

 

 

The ten-vear growth of the ORNL effort on homogeneous reactors is
indicated by Table 1-1, which summarizes the costs and man-years de-
voted to the program through fiscal year 1958.

Following the completion of construetion and beginning of operation of
HRIE-2, the ORNL Homogeneous Reactor Project directed its attention
to the design of a 60-Mw (heat) experimental aqueous thorium breeder
reactor, designated as HRE-3, with the objective of completing the con-
ceptual design during the summer of 19538. Work on slurry development
and component development was accelerated to provide the information
necessary for the start of construction of HRE-3 at the earliest possible
date.

1-1.6 Industrial participation in homogeneous reactor development. In-
dustrial participation in the homogeneous reactor program started with a
number of studies to evaluate the economic potential of such reactors for
large-scale power production [18-22]. The opinion of some who compared
homogeneous breeder reactors with solid-fuel converters is reflected in the
following excerpts from Ref. 19: “The two reactor types that offer the
greatest possibilities for economic production of central station power are
the thermal U233 breeders of the circulating fuel type and fast plutonium
breeders containing fuel easily adaptable to a simple processing system . .
The self-regulating features of fluid-fuel reactors and low fission-product
inventory due to continuous chemical processing give these reactors the
greatest possibility of safe and reliable operation . . . Both the pressurized
10 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [cHAP. 1

water and sodium-graphite systems suffer from the inability to consume
(in a single cycle) a large fraction of the uranium necessary to result in
low fuel costs that are attainable with breeder systems.”

During late 1954 and early 1955, Westinghouse and Pennsylvania
Power and Light Company, operating under Study Agreements with the
Atomic Energy Commission, made a joint study [21] aimed at determining
the economic feasibility of aqueous homogeneous-type reactor plants. The
study indicated that a two-region solution-slurry plant and a single-region
slurry plant appeared to have excellent long-range possibilities for pro-
ducing competitive electric power. The study also indicated, however,
that considerable development work would be required before the tech-
nical feasibility of either type of plant could be determined with any
degree of certainty. The results of this and other continuing studies led
the two companies to set up the Pennsylvania Advanced Reactor Project
in August 1955. An initial proposal to build a 150-Mw (electric) power
station financed with private funds was made to the A.E.C. by the Pennsyl-
vania Advanced Reactor group at that time. This proposal was later modi-
fied and resubmitted as part of the power demonstration reactor program.

In spite of the formidable development program which appeared to be
assoclated with the construction of a full-scale homogeneous reactor
power plant, a second industrial group proposed building a homogeneous
reactor as part of the power demonstration program in cooperation with
the government. This proposal (made in response to a request by the
Atomic Energy Commission for small-scale reactors) by the Foster Wheeler
and Worthington Corporations in January 1956, considered construction
of an aqueous homogeneous burner reactor. Plans were for a reactor and
associated oil-fired superheater with a net electrical capacity of 10,000 kw
for the Wolverine Electric Cooperative, Hersey, Michigan. Although this
proposal was accepted in principle by the Atomic Energy Commission in
April 1956, and money was appropriated by Congress for carrying out the
project, in May 1958 the Atomic Energy Commission announced that
plans had been canceled due to increases in the estimated cost of the plant
(from $5.5 million to between $10.7 and $14.4 million).

The second proposal submitted to the Atomic Inergy Commission
jointly by the Pennsylvania Power and Light Company and Westinghouse
Electric Corporation was determined by the Commission on February 26,
1958, as acceptable as a basis for negotiation of a contract but was later
recalled, following a review by the Joint Congressional Committee on
Atomic Energy. The proposal called for the construction of a reactor of
the homogeneous type with a net electrical output of 70,000 to 150,000 kw
to be operated on the Pennsylvania Power and Light Company system.
The reactor would use a thorium-uranium fuel as a slurry in heavy water.
Under the proposal, the Atomic Energy Commission would assume the
1-2] GENERAL CHARACTERISTICS: HOMOGENEOUS REACTORS 11

cost of research and development planned for 1958 and 1959, at which
time a decision would be made either to begin actual construction of a
plant or terminate the project. The cost of the project, scheduled for com-
pletion by December 1963, was estimated at $108 million. The Westing-
house and Pennsylvania Power and Light Company’s share of the cost
included $5.5 million for research since 1955, $57 million for plant con-
struction, and $16 million for excess operating costs during the first five
years of operation. The Atomic Energy Commission was asked to provide
the additional $29 million, including $7 million for research and develop-
ment in 1958-1959, $18 million for research and development following a
decision to construct the plant, and $4 million for fuel charges during the
first five years of operation.

1-2. GENERAL CHARACTERISTICS OF HOMOGENEOUS REACTORS

1-2.1 Types of systems and their applications. Because of the large
number of possible combinations of mechanical systems and compounds
of uranium and thorium which may be dissolved or dispersed in H20 or
D20, there exists in principle an entire spectrum of aqueous homogeneous
reactors. These may be classified according to (a) the type of fissionable
material burned and produced (U%35 burners, converters, breeders), (b) the
geometry or disposition of the fuel and fertile material (one-region, two-
region), or (c) the method of heat removal (boiling, circulating fuel, and
fluidized suspension reactors). The possible materials which can be used
in these various reactor types are given in Table 1-2; all combinations are
not compatible.

TasLE 1-2

HoMocENEOUS REACTOR MATERIALS

 

 

 

F Fertile Moderator Corrosion—registant
uel . metals of primary
material coolant .
interest,
U02804 + H2S04 1238 galt D20 Austenitic stainless
steels
UO:Fs+ HF U238 oxide H-0 Zircaloy—2
UO2N1305+ HNO3 ThO» Titanium
U080, + LiaS0, Platinum
UO3 + alkali oxide + CQOg Gold
U03+ H3PO4, UO2+ H3 PO,
UOs + HzCrOy4
UQO,, UO3, UsOs

 

 

 

 

 

 
TABLE 1-3

HomocEnrOUS REACTOR TYPES AND APPLICATIONS

 

Reactor designation

Power level range,

Fuel solution or suspension

Application

 

 

 

Mw heat
Water boiler 0-0.05 Enriched U02804 or TUO2(NOj3)2 [University nuclear research and
in H,0 training
Homogeneous research reactors 800-2000 Enriched U02804 in D20 Nuclear research at ultra-high
thermal-neutron fluxes
U235 burners 40-500 Enriched U02804 in H20 or D20 | Small- to large-scale power plants
in high-fuel-cost locations; mo-
bile power plants
LAPRE type power reactors 1-100 Enriched UOj dissolved in 60 w/o | Remotely located small- and inter-
phosphorie acid mediate-scale power plants
Enriched UO2 dissolved in 95 w/o
phosphoric acid
One-region power converters 500-1000 Slightly enriched UQ3 in D20 Large-scale power production
One-region Pu producer 1000-2000 Slightly enriched U02S04 in D20 | Dual-purpose power plus pluto-
[with or without added Li%(504)] nium production
Two-region Pu producer 500-1500 Enriched U02S04 in D20 (core) Dual-purpose power plus pluto-
Depleted U0 2504 1in D20 (blanket) nium production
Single-region thorium breeder 500-1500 Enriched U235 or U233 oxide plus | Large-scale power production
ThOz in D 20
Two-region thorium breeder, 200-1000 Enriched U235 or U233 a5 U0,S0, | Large-scale power production and
solution core in D20 (core) plus ThOQ2 in DO U233 breeding or U235 to U233
(blanket) conversion
Two-region thorium breeder, 200-1000 Enriched U235 or U233 oxide plus | Large-scale power production and

slurry core

 

 

ThOgz in D20 (core) plus ThOq
in D20 (blanket)

 

U233 breeding or U235 to [233
conversion

 

 

Gl

INHKRJOTHAHd YIHHL NV SHOLOVHY SNOUNHADHONWOH

I 'dvHD]
1-2] GENERAL CHARACTERISTICS. HOMOGENEOUS REACTORS 13

The terms used in classifying homogeneous reactors may be defined as
follows: Burner reactors are those in which fissionable fuel is consumed
but virtually no new fuel is generated. To this class belong the water
boilers, homogeneous research reactors, U235 burners, and LAPRE-type
reactors. Converter reactors produce a different fissionable fuel than is
destroyed in the fission process, such as in the dual-purpose plutonium
producers or single-region converters, while breeder reactors produce the
same fissionable fuel as that which is consumed. One-region reactors con-
tain a homogeneous mixture of fissionable and fertile materials in a moder-
ator. Generally, these have large reactor diameters, in order to minimize
neutron losses, and contain fuel plus fertile material in concentrations of
100 to 300 g of uranium or thorium per liter of solution or slurry. Two-
region reactors are characterized by a core containing fissionable materials
in the moderator surrounded by a blanket of fertile material in moderator,
These reactors may have comparatively small diameters with dilute core-
fuel concentrations (1 to 5 g of uranium per liter) and a blanket containing
900 to 2000 g of fertile material per liter. Boiling reactors are reactors in
which boiling takes place in the core and/or blanket and heat is removed
by separating the steam from the solution or suspension. Fluidized sus-
pension reactors are those in which solid particles of fuel and fertile ma-
terial are fluidized in the core and/or blanket, but are not circulated
through the cooling system external to the reactor pressure vessel.

A summary of homogeneous reactor types and the primary application
of each is given in Table 1-3.

1-2.2 Advantages and disadvantages of aqueous fuel systems. Aqueous
fuel systems possess certain advantages which make them particularly
attractive for numerous nuclear-reactor applications ranging from small
reactors (for mobile units or package-power plants) to large, high-power
reactors (for large-scale production of plutonium, U233 and/or power).
These advantages stem partly from the fluid nature of the fuel and partly
from the homogeneous mixture of the fuel and moderator; i.e., an aqueous
homogeneous reactor combines the attributes of liquid-fuel heterogeneous
reactors with those of water-moderated heterogeneous reactors. If practical
methods for handling a radioactive aqueous fuel system are developed, the
inherent simplicity of this type of reactor should result in considerable
economic gains in the production of nuclear power and fissionable material.

However, many apparently formidable practical problems are associated
with continued operation and maintenance of systems involving radio-
active fuel solutions. It is believed, therefore, that extensive experience in
a series of small- to large-scale reactor installations will be required to
demonstrate the reliability of aqueous homogeneous reactors; this will
necessitate a long-range development program. In addition, the choice of
14 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [cHaP. !

water as the fuel-bearing medium limits both the fuel concentration and
operating temperature to values which may be less than optimum for pro-
duction of power and fissionable material.

The principal advantages of aquecus fuel systems are:

(1) High power density. Because of the homogeneous nature of the
reactor fuel-fluid, virtually no heat-transfer barrier exists between the fuel
and coolant. Thus reactor power densities of 50 to 200 kw/liter mayv be
possible, being limited by considerations other than heat transfer, such as
radiation-induced corrosion and chemical reactions.

(2) High burnup of fuel. In heterogeneous reactors, burnup is limited
by radiation damage to fuel clements or loss of reactivity. In liquid-fuel
reactors, continual removal of poisons is possible, as well as continual
additions of new fuel, thereby permitting unlimited burnup.

(3) Continuous plutonium recovery. Continuous removal of neptunium
or plutonium is possible in a liquid-fuel reactor. This yields a product with
a low Pu24? content and Increases the value of the plutonium [23].

(4) Simple fuel preparation and reprocessing. The use of aqueous fuel
solutions or slurries eliminates the expensive fuel-element fabrication step
and simplifies the reprocessing of depleted fuel.

(5) Continuous addition or removal of fuel. Charging and discharging fuel
can be accomplished without shutting down the reactor and without the
use of solid-fuel charging machines.

(6) High neutron economy. Neutron economy is improved by eliminating
absorption of neutrons by ecladding and structural material within the
reactor core. Also, there is the possibility of continuously removing
Xel!35 and other fission-product poisons. In addition, an acqueous fuel
system lends itself readily to a spherical core geometry, which minimizes
neutron leakage.

(7) Stmple control system. Density changes in the moderator create u
sensitive, negative temperature coefficient of reactivity which makes this
system self-stabilizing. This eliminates the need for mechanically driven
regulating rods. In addition, shim control can be achieved by changing the
fuel concentration.

(8) Wide range of core sizes. Depending on concentration and enrich-
ment, critical H2O and D20 homogeneous reactors range from 13 ft to as
large as is practicable. Correspondingly, there is a wide range of applica-
tion for these reactor systems.

The prineipal problems of aqueous fuel systems are:

(1) Corrosion or erosion of equipment. The acidity of fuel solutions and
abrasiveness of slurries at high flow rates creates corrosion and erosion
1-2] GENERAL CHARACTERISTICS: HOMOGENEOUS REACTORS 15

problems in the reactor and its associated equipment. Special provisions
must therefore be made for maintaining equipment.

(2) Radiation-induced corrosion. 'The presence of fission radiation in-
creases the rate of corrosion of exposed metal surfaces. This limits the per-
missible wall power density, which in turn restricts the average power
density within the reactor.

(3) Lxternal circulation of fuel solution. Removal of the heat from the
reactor core by circulating fuel solution, rather than coolant only, through
external heat exchangers increases the total amount of fuel in the system
and greatly complicates the problems of containment of radioactivity and
accountability of fissionable material. The release of delayed neutrons in
the fuel solution outside of the reactor core reduces the neutron economy
of the reactor and causes induced radioactivity in the external equipment,
resulting in the need for remote maintenance.

(4) Nuclear safety. The safety of homogeneous reactors is associated
with the negative density coeflicient of reactivity in such systems; how-
ever, by virtue of this coeflicient, relatively large reactivity additions are
possible through heat-exchanger mishaps and abrupt changes in fuel cir-
culation rate. In boiling reactors changes in the volume of vapor within
the reactor core may lead to excessive reactivity changes.

(5) Liomited wranium concentration. In solution reactors, uranium con-
centration 1s limited by solubility or corrosion cffects, and in slurries, by
the effective viscosity and settling characteristics. In HeO-moderated
reactors, in particular, a high uranium or thorium concentration is neces-
sary for a high conversion ratio. Concentrations up to 1000 g/liter, how-
ever, may be considered for solutions and up to 4000 g/liter for fluidized
beds.

(6) Luomated operating temperatures. At the present time the operating
temperatures of aqgueous solution systems appear limited because of cor-
rosion problems at ~225°C and phase stability problems above 300°C.
Pressures encountered at higher temperatures are also a problem.

(7) Lxplosive decomposition product. Radiation-induced decomposition
of the moderator can produce an explosive mixture of hydrogen and oxygen
in the reactor systen. This hazard means that special precautionary design
measures must be taken. To prevent excessive gas formation and reduce
the requirement for large recombiners, a recombination catalyst such as
cupric ion may be added. Disadvantages associated with this addition are
the neutron poisoning effects and changes in chemical equilibria which
oceur.,

A comparison of the advantages and disadvantages of specific homo-
geneous reactors is given in Table 1-4.
TaBLE 14

ComprarisoN or HomoGceNeEoUs Reactor TYPESs

 

Reactor types

Advantages

Disadvantages

 

 

 

One-region U?23% burner,
H>0 or D20 moderator

Two-region breeder,
solution or slurry core

One-region ThOg slurry

One-region UO3 slurry

One-region UOeS0y solution

 

Possible elimination of chemical process-
ing plant

Elimination of D20 requirement
(H20 moderator)

Low fissile-material inventory
(D20 moderator)

High neutron economy and low fuel costs

Low fissile-material inventory

Possible fission-product removal from core
solution

High neutron economy and low fuel costs

Elimination of zirconium as a construction
material

Relatively low fissile- and fertile-material
inventory

Elimination of zirconium problems

Elimination of slurry handling problems

 

 

Relatively high fuel costs (due to burning of
enriched uranium with no regeneration) com-
pared to homogeneous breeders and converters

Radiation corrosion of zirconium core tank limits
power density (may be more serious with
solution core compared with slurry core)

Slurry handling problems

Startup and shutdown of reactor may be
difficult

Slurry handling problems

Startup and shutdown problems

Slurry handling problems

May require all-titanium system

Plutonium does not stay in solution and may
deposit on walls of equipment

 

INHNWJOTIAHCI YITHL NV SHOLIVAY SNOHNHIOWOH g7

1 'dVHD]
1-3] U235 BURNER REACTORS 17

1-3. U2% Bur~NER RBACTORS

1-3.1 Dilute solution systems and their applications. One-region re-
actors fueled with a dilute solution of highly enriched uranium or "'burner
reactors” are ideal as a concentrated source of neutrons, since the critical
mass and size of the core of this type of reactor can be very small. Many
low-power research reactors are in operation which use this fuel system,
and very-high-flux research reactors of this type are heing considered [24],
The principal advantages of solution reactors for this latter application are
the small amount of U2 required for criticality and the ability to add
fuel continually.

One-region burner reactors are applicable for both small- and large-
scale nuclear power plants. Such plants can operate for very long periods
of time (20 years or more) without necessity for removal of all the fission
products. Corrosion product buildup, however, must be limited to prevent
uranium precipitation. The fuel concentration would be dilute, increasing
with time of reactor operation if no fuel processing is carried out. Either
light or heavy water cun be used as the moderator-coolant; the fuel con-
centrations would alwuys be higher for the light-water-moderated reactors.
An advantage of these syvstems is that they utilize fuel in the concentration
range which has been studied most extensively. Fxperience in circulating
such solutions, however, indicates that careful control of operating condi-
tions and the concentrations of the various fuel constituents, such as
H2S504, CuS04, NiSOy, H202, Og, ete., is necessary to avoid problems of
two-phase separation, uranium hydrolysis, and oxygen-depletion precipi-
tation of uranium.

I'or power production, homogeneous burner reactors can be considered
as possible competitors to the highly enriched solid-fuel reactors, such as
the Submarine Thermal Reactor and the Army Package Power Reactor.
By eliminating fuel-element fabrication, fuel costs in homogeneous burners
with either 1320 or Hz0 as the coolant-moderator are in the range of
4 mills/kwh at present Atomic Energy Commission prices for enriched
uranium [25].

Possible fuel systems for the dilute, highly enriched burner-type reactors
are UO2804 in HaS04, TO2(NO3)2 in HNOj, UOsF: in HF, and UOs-
alkali metal oxide-COy in H20. These fuel systems are compared in
Chapter 3.

1-3.2 High-temperature systems. TFuel systems of enriched uranium
dissolved in highly concentrated phosphoric acid have been suggested for
homogeneous power reactors because of the high thermal stability and low
vapor pressure of such systems. This permits operation at higher tempera-
tures than is possible with dilute acids, with accompanying higher thermal
efficiencies. Fuel systems of this type include TOj3 in 30 to 60 w/o (weight
18 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [cHAP. 1

percent) phosphoric acid, UO2 in 90 to 100 w/o phosphoric acid, and U3
in concentrated chromic acid. The UQO3-H3PO4 system, used in the Los
Alamos Power Reactor Experiment No. 1 (LAPRE-1), must be pressur-
ized with oxygen to prevent uranium reduction. Solutions containing
phosphate-to-uranium ratios of 4/1 to 10/1 are stable up to 4350°C. How-
ever, the neutron economy is poor and these solutions are corrosive to ull
metals except platinum and gold. The UOo—-H3PO4 systems, pressurized
with hydrogen, have somewhat better corrosion characteristics and copper
may be used at least in regions which are kept below 250°C.

1-4. CONVERTER REACTORS

1-4.1 Purpose of converters. In converter reactors, U?3% is burned to
produce U233 or Pu??® by absorption of excess neutrons in fertile material.
Thus the purpose of converter reactors is the production of power. fission-
able material, or both. Since homogeneous reactors have to operate at
temperatures above 225°C and pressures above 1000 psi because of proh-
lems of corrosion and gas production, homogeneous converters are thought
of as dual-purpose reactors for the production of power and fissionable
material or power-only reactors. Such reactors are also considered mainlv
in connection with the UZ235-17238-Py?39 fuel cycle, whereas the homo-
geneous breeder reactors are associated with the thorium fuel eycle.

1-4.2 One-region converters. One-region converter reactors mav be
fueled with a relatively concentrated solution (100 to 300 ¢ liter D20} of
slightly enriched uranium for plutonium and power production or with a
suspension of slightly enriched uranium oxide for power production only.*
The principal advantage of the solution-type converter for plutonium pro-
duction is the insolubility of plutonium in the high-temperature uranium
sulfate system (see Chapter 6). This opens the possibility of separating the
plutonium by centrifugation rather than by a solvent extraction or ab-
sorption process. The costs of this method of recovering the plutonium.
which contains only small amounts of Pu?*°, should be considerably less
than i1s possible with solid-fuel reactors and conventional processing tech-
niques. Indications are, however, that the plutonium formed in the fuel
solution is preferentially adsorbed on hot metal surfaces in contact with
the solution and 1s difficult to remove (see Chapter 6). Other problems with
the solution-type converter are the highly corrosive nature of concentrated
uranyl sulfate solutions and the lower temperature at which the two
liquid phases separate. An all-titanium high-pressure system mayv bhe

*Early work at Columbia and Chicago was aimed at a low-temperature version
of such a reactor for plutonium production only; however, present-dayv considera-
tions are limited to high-temperature systems.
1-5] BREEDER REACTORS 19

necessary to contain these solutions, which will lead to considerably higher
equipment and piping costs. The addition of lithium sulfate to the solution
would reduce corrosion and raise the phase-separation temperature so
that it might be possible to use stainless steel; however, the neutron
economy with normal lithium is poorer and separated Li? would be costly.

A single-region converter fueled with natural or slightly enriched uranium
oxide as a suspension avoids the problems of plutonium precipitation,
phase separation, and corrosion mentioned above. The advantage of such
a converter reactor for power production is the elimination of radiation
damage and fuel burnup problems encountered with solid-fuel elements;
however, the problem of radiation damage to the reactor pressure vessel
must be considered.

1-4.3 Two-region converters. Two-region homogeneous converters may
also be fueled with either D0 solutions or slurries; in these reactors, how-
ever, the U#3° is in the core and the fertile material in the blanket. Con-
verters of this type become breeders if the bred fuel is subsequently burned
in the core and there is a net gain in the production of fuel. A two-region
converter with a dilute enriched-uranium core solution and a concentrated
depleted-uranium blanket solution shows promise of producing more eco-
nomical power and plutonium than the one-region converter reactors
mentioned previously [26] because of the lower inventory charges and
the better neutron economy. Although the power density at the wall of
the titanium-lined pressure vessel is lower in the case of the two-region
machine, which minimizes the possibility of accelerated corrosion rates,
there is some evidence [27] that titanium corrosion will not be severe in
any case. The major materials problem in the dilute-solution core converter
will be that of zirconium corrosion, which may be above 30 mils/year at
power densities necessary for economic production of power and fission-
able material.

Two-region converters fueled with a uranium oxide slurry in the core
may be a possibility as an alternative to the solution-slurry system; how-
ever, not much is known about the corrosion resistance of zirconium in
contact with fissioning uranium oxide or about the engineering behavior
of such a slurry.

1-5. BREEDER REACTORS

1-5.1 The importance of breeding. If present projections [28] for the
growth of the nuclear power industry in the United States are correct, the
installed capacity of nuclear electric plants in 1980 may be as much as
227 million kilowatts and may be increasing by 37 million kilowatts an-
nually. Even assuming optimistic figures for fuel burned in then-existing
plants and fuel plus fertile material for inventories in new plants [29],
20 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [cHaPp. ]

the annual requirement of fissionable material will be approximately
420,000 kg in 1980. This fissionable material will have to come tfrom nut-
ural sources (i.e., uranium mined from the ground) or be produced from
neutrons absorbed in fertile material in a reactor (i.e., breeding or con-
version). Since presently known reserves of high-grade ores of uranin
and thorium in the United States [30] contain 148,000 tons of uraniur.
and 60,000 tons of thorium, respectively, and these in turn contain only
106 kg of fissionable material, it is obvious that conversion of a significuns
fraction of the fertile material contained in the reserves will be nece<sury,

Although such a conversion will not reduce the inventory requiremern:
of fuel and fertile material for new plants starting up, this amounts o
only about 259, of the burnup requirement. On this basis. the goal of
nuclear industry should be to develop reactor designs and associuted tuel
systems which achieve a consumption of at least 5% to 109 of the torul
fertile material, as well as the initial fissionable material. At thix poin:
the annual burnup requirement would become small compared with the
inventory requirement. This corresponds to a total burnup of ubont
50,000 Mwd/ton. Although such a fuel consumption might be obtuine
in high-neutron-economy converter reactors through recycling of the fuci.
it seems likely that even the best such reactor may fall short of this goul
and that both fast and thermal breeders will be needed.

In the long term, therefore, the development of breeding system=x 1=
must. In the short term, where emphasis is on fuel costs rather than on
neutron economy and fertile-material utilization, converters rather thun
breeders may predominate.

1-5.2 One-region thorium breeders. Since U233 does not occur in nature.
homogeneous thorium breeder reactors will probably start out as con-
verters, with U235 as the fuel and thorium as the fertile material. One-
region reactors of this type utilize a suspension of 100 to 300 g per liter of
thorium oxide plus enriched U?35 ag oxide and D20 as the moderator. [n
order to maintain a breeding ratio greater than 1, fuel processing is neces-
sary to remove fission-product poisons. Also, to reduce losses due to neutron
leakage, the diameter of the reactor should be at least 12 ft.

1-5.3 Two-region breeder reactors. Two-region breeder reactors would
have thorium oxide suspensions (500 to 1500 g/liter) in the blanket re-
gion and could have either a highly enriched uranyl sulfate solution
(1 to 10 g/liter) or a thorium oxide-uranium oxide slurry (200 g ThO.
liter and 10 g UO3/liter) in the core region. Use of a solution-type core
permits the continuous removal of insoluble fission-product poisons by
means of hydroclones, while a slurry-type core leads to higher breeding
ratios. Because of these compensating factors, estimated fuel costs are
1-6] MISCELLANEOUS HOMOGENEOUS TYPES 21

approximately the same in both types of reactors (see Chapter 10). While
the use of a suspension in the core may minimize the problem of radiation-
induced corrosion of the zirconium, not much is yet known about the
behavior of zirconium in a thorium-uranium slurry-fueled reactor. Cal-
culations summarized in Chapter 10 show that both solution- and slurry-
fueled two-region breeders have higher breeding ratios and lower fuel costs
than one-region breeders.

Numerous studies of large-scale two-region breeder reactors have been
carried out [20,26,31-42], some of which are described in detail in Chap-
ter 9.

1-6. MiscELLANEOUS HoMoGENEOUS TYPES

1-6.1 Boiling reactors. In May 1951, following completion of the con-
struction of HRE-1, a group at the Oak Ridge National Laboratory
focused its attention on the possibility of removing heat from a homo-
geneous reactor by boiling, rather than by circulating the fuel solution,
in recognition of the advantages of a boiling reactor. These are: (a) more
rapid response to sudden reactivity increases, minimizing power excur-
sions, (b) elimination of fuel circulating pumps, (¢) increase in the tem-
perature of steam delivered to the turbine for a given reactor operating
pressure, and (d) reduction or elimination of problems of corrosion and
induced radioactivity associated with the circulation of fuel and fertile
material through an external heat-removal system. However, at that
time, questions of the nuclear stability of a boiling, liquid-fuel reactor
and the maximum specific power, in terms of kilowatts per liter, that could
be extracted from a given size core remained to be answered.

Experiments on bulk boiling at atmospheric pressures in a 1-ft-diameter
cylindrical tank indicated that power densities up to 5 kw/liter might be
achieved. It soon became apparent, however, that high-pressure power-
density measurements would be required, and the design of a boiling reactor
experiment (BRE) called the “Teapot” was initiated. To answer the
question of the nuclear stability of such a reactor, a combined group from
the Oak Ridge National Laboratory and Los Alamos operated the SUPO
under boiling conditions in October 1951. The reactor was operated at a
total power of 6 kw and solution power densities of 0.5 kw/liter were
obtained.

This removed one of the important obstacles to the construction of an
experimental boiling reactor, and in January 1952 the Oak Ridge National
Laboratory made a proposal to the Atomic Energy Commission to con-
struct the Boiling Reactor Experiment (BRE) to answer the question of
maximum specific power at higher pressures and to investigate the operat-
ing characteristics of boiling reactors. The proposed reactor was to operate
at a power level of 250 kw of heat and pressures up to 150 psi. The re-
22 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [cHAP. 1

actor was estimated to cost approximately $300,000, including the building
to house it. A one-year effort involving nuclear and engineering calcula-
tions, completion of the BRE conceptual design, and experiments on
bubble nucleation in the presence of radiation resulted from the proposal,
In January 1953, however, the problem of maintaining sufficient oxidizing
conditions to prevent reduction and precipitation of the uranium in a boil-
ing uranyl sulfate solution became apparent, and construction of the reactor
was deferred pending outcome of solution-stability experiments. These
experiments, completed in Octeber 1953, indicated that at oxygen concen-
trations likely to be encountered in a boiling reactor (6 to 7 ppm), reduction
of the uranium would occur in solutions in contact with stainless steel.*
Since the metallurgy of titanium or zirconium was not sufficiently advanced
to construct a reactor using these alternate metals, it was decided to
abandon the BRI itself and continue experimental work on the problems
of solution stability, steam separation, and power densities at high pressure.
Work on steam separators and experimental measurements of the move-
ment of air and steam through heated solutions at high pressures were
carried out under contract by the Babcock & Wilcox Company [44]
These results and theoretical calculations of the power removal from boil-
ing reactors [45,46] provide a basis for estimating the obtainable power
density of such reactors under varying core heights, operating pressures,
and moderator density decreases. Values range from 10 to 40 kw liter
with an average of 18.5 kw/liter for a 15-ft-high core, operating at 2000 psi,
and a density decrease due to steam of 0.4. Although the effect of such
a void fraction on nuclear stability is not known, if tolerable, boiling re-
actors may be able to achieve average power densities comparable to those
estimated for large-scale nonboiling circulating-fuel reactors operating
under a similar pressure [47]. In this latter case, the holdup of solution
in the external circulating system lowers the power density of the core
only, to an average of about 10 to 20 kw/liter. The two systems are
comparable, therefore, in terms of obtainable power densities, and boiling
reactors cannot be excluded on this basis.

The various applications of boiling as a method of heat removal from
homogeneous reactors include a one-region boiling solution or slurry re-
actor, a two-region reactor with a nonboiling core and a boiling blanket,
and a two-region reactor with a boiling core and a boiling blanket. The
problem of maintaining a sufficiently oxidizing solution in a boiling uranyl
sulfate—1s0O reactor, although serious in a stainless-steel system, can be
eliminated if all surfaces in contact with the solution are made of titanium
and oxygen is supplied continuously. Solutions containing 10 g of uranium

*Tn more recent tests [43] with nonboiling solutions, in which oxygen concentra-
tions were held at 2 to 3 ppm, no reduction and precipitation of uranium occurred.
1-6] MISCELLANEQUS HOMOGENEOUS TYPES 23

per liter have been successfully boiled at 325°C in titanium-lined pipe [48].
Experiments have not yet been carried out with higher uranium concen-
trations in titanium. Continued interest in boiling homogeneous reactors
has led to a number of studies of large-scale reactors of this type
[33,36,37,49-51]. The actual construction of a boiling slurry reactor
1s under way in the USSR [52].

Use of boiling as o method for removal of power from the ThOs slurry
blanket of large two-region homogeneous reactors appears to present no
major difficulties [53]. The apparent advantages are that no circulating
pump would be required to handle slurry, containment of the highly active
slurry in the reactor vessel, and the possibility of operating the blanket
at the core pressure and using the blanket power for heating the secondary
steam [54]. One major problem is that of keeping the slurry suspended
during startup.

1-6.2 Gaseous homogeneous reactors. Although this book deals pri-
marily with aqueous svstems, some mention should be made of other
types of fluid-fuel homogeneous reactors in which the fuel and moderator
are mixed and can be circulated. The existence of UFg, which has a low
parasitic capture cross section and is a gas at ordinary temperatures,
makes possible the consideration of gaseous reactors. UFy boils at 56.4°C
at atmospheric pressure and has a critical temperature of 252°C at 720 psi.
Although Ul is corrosive to most metals, it can be contained in nickel
and monel. However, the effect of radiation on the integrity of the pro-
tective film has not been studied. Considerable experience has been
gained in the handling of UFs in metal containers at high temperatures
and pressures,

Pure Ul is not a practical possibility for a gaseous homogeneous re-
actor because fluorine is not a good enough moderator. Addition of helium
makes such a reactor possible, and calculations by D. E. Hull in a report,
“Possible Application of UFg in Piles” [55], show that the eritical mass of
a graphite-reflected, He 4 ULy, reactor is 84 kg of U235, About 15 tons of
helium 1n a 60-ft-diameter core would be required. In a recent investiga-
tion [56] of reflector-moderated gaseous reactors (Plasma Fission Reactor),
the critical mass of gaseous U235 in a 2-ft-diameter cavity surrounded by
D20 was calculated to be less than 1 kg. Such reactors would have to
operate at extremely high temperatures (3000°K) which many feel are
beyond the realm of present technology.

Mixed UFg gas and dispersions of solid moderators such as beryllium or
graphite have been suggested, as well as beds of moderator particles
fluidized with circulating UFg [55]. However, these proposals have no
apparent advantages compared with gas- or liquid-cooled fluidized systems
described in the following section.
24 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [cHAP. 1

1-6.3 Fluidized systems. A variant of the fixed-bed or solid-moderator
homogeneous reactors consists of subdividing the fuel and/or moderator
to the point where the particles can be fluidized by the flowing gas or
liquid. Gas-cooled reactors of this type have received considerable atten-
tion because of the higher heat-transfer rates obtainable compared with
fixed-bed reactors. A number of studies of gas-fluidized reactors have been
carried out by ORSORT groups at the Oak Ridge National Laboratory
[57,58] and by other groups [59,60].

In an Oak Ridge study [61] various types of fluidized-bed reactors were
compared. Systems investigated were (a) a sodium-cooled fast reactor,
(b) a gas-cooled system, (¢) an organic-moderated reactor, (d) a heavy-
water-moderated reactor, and (e) a light-water system. A detailed study
of this latter system was carried out to compare its characteristies and
performance with solid-fuel heterogeneous pressurized-water reactors.
The results indicated that both the light-water-moderated and organie-
moderated fluidized reactors showed promise, while the gas-cooled, the
D20-cooled, and the fast (unmoderated) reactors were found to be less
satisfactory for application of the fluidized-bed technique.

Systems using ThO2, fluidized by gas or D20, were described by the
Dutch at the Geneva Conference on Atomic Energy in 1955 [62,63].

Calculations show that a typical, one-region, 400 thermal Mw reactor
having a core diameter of 15 ft and a temperature rise of 50°C would re-
quire particles in the 40- to 60-micron range [(4], whereas a two-region
reactor with a liquid-fluidized blanket would require particles in the 200-
to 600-micron range [65], and if the particles were confined to a 6-in.
annulus next to the core the particle size required would be in the 0.10-
to 1.5-cm range [65].

The disadvantages that may be observed with fluidized suspension
systems include the possibility of particle attrition [65], and instabilities
due to channeling during steady-state operation and due to settling if a
circulating pump failed.

Room-temperature attrition tests using 0.1-in.-diameter X 0.1-in.-long
ThO» and ThOz + 0.59, CaO cylinders (cylinders prepared by calcination
at temperatures of both 1650 and 1800°C) fluidized in water gave an
attrition rate of 12 to 159, weight loss per week [66]. However, circula-
tion tests using 10- to 20-micron ThO2 spheres (calcined at 1600°C) 1n
toroids at superficial velocities up to 26 ft/sec and water temperatures of
250°C showed essentially no attrition for periods up to 200 hr [67].

This appears to indicate that attrition rate is at least a function of par-
ticle size, without giving any indication as to the effect of void fraction,
slip velocity, particle shape and density.
REFERENCES 25

REFERENCES

1. H. Harean and L. Kowarski, Cambridge University, March 194].
Unpublished.

2. H. Haupan et al,, Number of Neutrons Liberated in the Nuclear Fission
of Uranium, Nature 43, 680 (1939).

3. H. L. ANpErsoN et al., Neutron Production and Adsorption in Uranium,
Phys. Rev. 56, 248 (1939).

4. . Harean et al.,, Mise en Evidence d’une Reaction Nucléaire en Chaine
au Sein d'une Masse Uranifére, J. phys. radium 10, 428 (1939).

5. E. Fermr and H. C. Urey, Memorandum of Conference Between Prof. E.
Fermi and Prof. H. C. Urey on March 6, 7, and 8, 1943, USAEC Report A-554,
Columbia University, Mar. 8, 1943.

6. C. F. Hiskry and M. L. EmiNorr, The Heavy-Water Homogeneous Pile:
A Review of Chemical Researches and Problems, USAEC Report CC-1383,
Argonne National Laboratory, Feb. 28, 1944,

7. A. Lanasporr, Slow Neutron Cross Section of Deuterium, USAEC Report
CP-902, Argonne National Laboratory, Aug. 30, 1943.

8. I. Karrax, Theory and Calculations of Homogeneous P-9 Piles, USAEC
Report A-1203. Columbia University, Sept. 10, 1943.

9. L. A. OHLINGER, Argonne National Laboratory, 1943, Unpublished.

10. L. A. OnLINGER, Destgn Features of Pile for Light-Water Cooled Power
Plant, USAEC Report CE-805, Argonne National Laboratory, July 16, 1943.

11. H. D. SyitH et al.,, Argonne National Laboratory, 1943. Unpublished.

12. I. KirsuenBauMm et al. (Kds.), Utilization of Heavy Water, USAEC
Report TID-5226, Columbia University, 1957.

13. C.P. BAkER et al., Water Boiler, USAEC Report AECD-3063, Los Alamos
Seientific Laboratory, Sept. 4, 1944.

14. L. D. P. King, Proceedings of the International Conference on the Peaceful
Uses of Atomic Energy, Vol. 3. New York: United Nations, 1956. (P/488)

15. J. W. Crasran (Ed.), U. S. Research Reactors, USAEC Report TID-
7013, Battelle Memorial Institute, August 1957.

16. C. D. CorverL and A, Turkevich, Oak Ridge National Laboratory, 1944.
Unpublished.

17. C. D. Coryerr and H. 8. BrowN, Oak Ridge National Laboratory, 1944.
Unpublished.

18. CommoNwEALTH EpisoN Company (NucLEAR Powrr GrouP) aND PuBLIC
SErvICE COMPANY OF NORTHERN [LLiNoIs, 1952. Unpublished.

19. A Survey of Reactor Systems for Central Station Power Production, USAEC
Report NEA-5301(Del.), Foster Wheeler Corporation—Pioneer Service and
Engineering Company, October 1953.

20. H. G. Carson and L. H. Laxoruwm (Eds.), Preliminary Design and Cost
Estimate for the Production of Central Station Power from a Homogeneous Reactor
Utilizing Thorium— Uranium-233, USAEC Report NPG-112, Commonwealth
Edison Company (Nuclear Power Group), February 1955,

21. WesTINGHOUSE ELECTRIC CorPORATION and PENNsYLvaNiaA POWER AND
LigaT CoMPaNY, 1955. Unpublished.
26 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [CHAP. 1

22. Single-fluid Two-region Aqueous Homogeneous Reactor Power Plant—
Conceptual Design and Feasibility Study, USAEC Report NPG-171. Common-
wealth Edison Company (Nuclear Power Group) and Babcock & Wilcox Com-
pany, July 1957.

23. AEC Plutonium Price Schedule, AEC Release, June 1957, Federal Register,
June 6, 1957.

24, P. R. KasteN et al., Aqueous Homogeneous Research Reactor—Feasibility
Study, USAEC Report ORNL-2256, Oak Ridge National Laboratory, April 1957,
25. AEC Price Schedule for Enriched Uranium, Nucleonics 14(12), {1956).

26. R. B. Brigas et al., OQak Ridge National Laboratory, 1954. Unpublished.

27. H. F. McDurrir, Corroston by Aqueous Reactor Fuel Solutions, USAEC
Report CF-56-11-72, Oak Ridge National Laboratory, November 1956.

28. W. K. Davis and L. H. Rovpis, AEC’s Fast Reactor Program, Vucleonics
15(4), 67 (1957).

29. J. A. LanE, Determining Nuclear Fuel Requirements for Large Scale
Industrial Power, Nucleonics 12(10), 65-67 (1954).

30. J. C. Jounson, Uranium Production To Match Needs, Chem. Eng. News
35, No. 48, 70-75 (1957).

31. J. A. La~E et al., Oak Ridge National Laboratory, 1950. Unpublished.

32. J. A. Lane et al., Oak Ridge National Laboratory, 1951. Unpublished.

33. L. C. Wippoes et al., Oak Ridge National Laboratory, 1951. Unpublished.

34. G. Purnawm et al., Reactor Design and Feasibility Problem; U-233 Power
Breeder, USAEC Report CF-51-8-213, Oak Ridge National Laboratory, 1951.

35. H. F. Kamack et al., Botling Homogeneous Reactor for Producing Power
and Plutonium, USAEC Report CF-54-8-238(Del.), Oak Ridge National Lab-
oratory, 1954.

36. H. C. CraiBorNE and M. Tosias, Some Economic Aspects of Thortum
Breeder Reactors, USAEC Report ORNL-1810, Oak Ridge National Laboratory,
October 1955.

37. D. C. Hamivton and P. R. KasTEN, Some Economic and Nuclear Character-
istics of Cylindrical Thorium Breeder Reactors, USAEC Report ORNL-2165,
Oak Ridge National Laboratory, October 1956.

38. M. F. DureT and I. L. WiLson, A Two-zone Slurry Reactor, Report
AECL-249, Atomic Energy of Canada Limited, December 1953.

39. CoMmmoNwEALTH EpisoN Company (NucLear Power Grour), A Third
Report on the Feasibility of Power Generation Using Nuclear Energy, USAEC
Report CEPS-1121(Del.), June 1953.

40. Paciric NorTHWEST Powkr Group, Aqueous Homogeneous Reactors.
USAEC Report PNG-7, February 1956.

41. CoMMONWEALTH EpisoNn CompaNY (NucLEAR PoweEr GroUP) AND Bap-
cock & WiLcox CompaNy, Evaluation of a Homogeneous Reactor, Nucleonics
15(10), 64-71 (1957).

42. American Standard and Sanitary Corporation.

43. J. C. Griess and H. C. Savacg, in Homogeneous Reactor Project Quarterly
Progress Report for the Period Ending July 31, 1956, USAEC Report ORNL-
2148(Del.), Oak Ridge National Laboratory, 1956. (p. 77)

44. T. A. Hucugs, Steam-Water Mixture Density Studies in a Natural Circula-
REFERENCES 27

tion. High-pressure System, USAEC Report BW-5435, Babcock & Wilcox Com-
pany, February 1958,

45. P. C. Zyora and R. V. BaiLey, Power Removal from Boiling Nuclear
Reactors, USALE.C Report CF-55-7-43, Oak Ridge National Laboratory, 1955.

46. L. G. Avexanprr and 3. Jayg, 4 Parametric Study of Rate of Power
Removal from Homogeneous Bovling Reactors, USAEC Report CF-55-9-172,
Oak Ridge National Laboratory, 1955.

47. J. A. Lane et al., Aqueous Fuel Systems, USAEC Report CF-57-12-49,
(Oak Ridge National Laboratory, 1957,

48, I. Sriewak, in Homogeneous Reactor Project Quarterly Progress Report for
the Pertod FEnding Oct. 31, 1957, USAEC Report ORNL-2432, Oak Ridge
National Laboratory, 1957. (p. 14)

49. J. M. Stei~y and P. R, KasTexN, Boiling Reactors: A Preliminary Investi-
gation, USAEC Report ORNL-1062, Oak Ridge National Laboratory, December
1951.

50. R.J. Rickert et al.. 4 Preliminary Destgn Study of a 10-Mw Homogeneous
Boiling Reactor Power Package for use in Remote Locations, USAEC Report
CF-53-10-23, Oak Ridge National Laboratory, 1953.

51. H. R. Zrrtuix et al., Boiling Homogeneous Reactor for Power and U-233
Production, USAEC Report CF-55-8-240, Oak Ridge National Laboratory,
August 1954,

22, A. I. ArvicHaxow et al.,, A Boiling Homogeneous Nuclear Reactor for
Power, in Proceedings of the International Conference on the Peaceful Uses of
Atomic Energy. Vol. 3. New York: United Nations, 1956. (p. 169)

53. P. C. Zxora, Comments on Botling Slurry Blankets for Homogeneous
Reactors, USAEC Report CF-55-9-125, Oak Ridge National Laboratory, Sept.
27, 1955.

54, P. C. Zxyova, Boiling Blanket for TBR-Power Utilization, USAEC Report
CF-55-3-57. Oak Ridge National Laboratory, Mar, 9, 1955.

35. D. E. Huwrr, Possible Applications of UFg, USAEC Report MonN-336,
Oak Ridge National Laboratory, 1947.

56. 8. A. Corgate and R. L. Aamont, Plasma Reactor Promises Direct
Electric Power, Nuecleonies 15(8), 50-55 (1957).

57. R. R. Harix et al., Oak Ridge National Laboratory, 1952. Unpublished.

58. H. W. Graves et al.,, Oak Ridge National Laboratory, 1953, Unpublished.

59. C. C. SivversTely, Fluidized Bed Reactor Cores, Nucleonics 15(3), 101
(1957).

60. J. B. Morrrs et al., The Application of Fluidization Techniques to
Nuclear Reactors, Trans. Inst. Chem. Engrs. London 34, 168-194 (1956).

61. C. L. Teerer et al.. Fluidized Bed Reactor Study, USAEC Report CF-57-
8-14, Oak Ridge National Laboratory, August 1957.

62. J. J. WeENT and H. ne Bruyn, The Design of a Small Scale Prototype of
a Homogenecous Power Reactor Fueled with Uranium Oxide Suspension, in
Proceedings of the International Conference on the Peaceful Uses of Atomic Energy,
Vol. 3. New York: United Nations, 1956. (P/936)

63. J.J. WenT and H. pE Bruyn, Power Reactors Fueled with “Dry’”’ Suspen-
sions of Uranium Oxide, in Proceedings of the International Conference on the
28 HOMOGENEOUS REACTORS AND THEIR DEVELOPMENT  [cHAP. 1

Peaceful Uses of Atomic Energy, Vol. 3. New York: United Nations, 1956
(P/938)

64. J. A. Lang, Oak Ridge National Laboratory, personal communication
Oct. 22, 1957.

65. P. R. CrowLey and A. S. Kirzes, Feasibility of a Fluidized Thorium,
Oxtde Blanket, USAEC Report CF-53-9-94, Oak Ridge National Laboratory,
Sept. 2, 1953.

66. I. Seiewaxk and J. A. Harrorp, Abrasion Test of Thorta Pellets, USAEC
Report CF-54-3-44, Oak Ridge National Laboratory, Mar. 9, 1954.

67. S. A. ReEp, Summary of Torotd Run No. 153: Circulation of Slurries of
Thoria Spheres at 250°C and 26 Fps Using Pins of Type 347 8.8., SA-2120-B
Steel, Titantum-75A, and Zircaloy-34, USAEC Report CF-57-11-46, Oak Ridge
National Laboratory, Nov. 11, 1957.