ANL-7228.txt

ANL-7228

MASTER

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Argonne Aational Laboratorp

CATALOG OF NUCLEAR REACTOR CONCEPTS

Part . Homogeneous and

Section VI. Solid Homogeneous

(Semihomogeneous) Reactors
by

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

RELEASED FOR ANNOUNCEMENT !

IN NUCLEAR SCIENCE ABSTRACTS E

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Quasi-homogeneous Reactors
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DISCLAIMER

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

Portions of this document may be illegible In
electronic image products. Images are produced
from the best available original document.
LEGAL NOTICE

This report was prepared as an account of Government sponsored work. Neither the United
States, nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or implied, with respect to the accu-
racy, completeness, or usefulness of the information contained in this report, or that the use
of any information, apparatus, method, or process disclosed in this report may not infringe
privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for damages resulting from the
use of any information, apparatus, method, or process disclosed in this report.

As used in the above, ‘‘person acting on behalf of the Commission’’ includes any em-
ployee or contractor of the Commission, or employee of such contractor, to the extent that
such employee or contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides access to, any information pursuant to his employment or contract
with the Commission, or his employment with such contractor.

Printed in USA. Price $3.00 . Available from the Clearinghouse for Federal
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-

ANL-7228

Reactor Technology

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Development Report

ARGONNE NATIONAL LABORATORY

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9700 South Cass Avenue

Argonne, Illinois

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HC §:307

CATALOG OF NUCLEAR REACTOR CONCEPTS

Part I.
Quasi-homogeneous Reactors

Homogeneous. and

Solid Homogeneous (Semihomogeneous) Reactors

Section VI.

by

Charles E. Teeter,-James A. Lecky,

Martens

and John H.

RELEASED FOR ANNOUNCEMENT
IN NUCLEAR SCIENCE ABSTRACT;

Technical Publications Department
June 1966

LEGAL NOTICE

This report was prepared as &n account of Government sponsored work. Neither the United
States, nor the Commission, nor any person acting on behalf of the Commission:

A. Makes any warranty or representation, expressed or Impiied, with respect to the accu-
racy, completeness, or usefulness of the infor mation contained in this report, or that the use
of any information, apparatus, method, or process dieclosed In this report may not infringe
grivately owned rights; or

B. Assumes any llabilities with respect tc the use of, or for damages resulting from the
vee of any information, apparatus, method, or process disclosed in this report. -

As used in the above, ‘‘person acting on cmww_m of the Commission’® includes any em-
ployee or contractor of the Commission, or employee of such contractor, to the extent that

such employee or contractor of the Commission, of employee of such contractor prepares,

" disseminates, or provides access to, any informatioh pursuant to his employment or contract

* with the Commission, or his employment with such contractor.

Operated by The University of Chicagé

under
Contract W-31-109-eng-38

- with the
U. S. Atomic Energy Commission

vl

fyp

TABLE OF CONTENTS

Preface. . . . . . . .

~Plan of Catalog of Nuclear Reactor Concepts.

List of Reactor Concepts . . . . .

SECTION VI. . SOLID HOMOGENEQUS (SEMIHOMOGENENIIS) REACTORS.

Chapter 1. ‘Introduction. . . . . ., .
Chapter 2. Research and Test Reactors.

-.Chapter 3. Power and Breeder Reactors.

-+~ PREFACE .

Thls'report is an additional section in the Catalog of Nuclear Reactor
" Concepts fhat was begun with -ANL-6892 and continued in ANL-6909, ANL-7092,
ANL-7138,: and ANL-7180. As in;the previous reports, the material is_dividee
into chapters, each with.text and references, plus data sheets that cover the
individual concepts. The plan of the catalog, with the report numbers for the
secfiions already»issued is given on the next page which is followed‘by pages-
llstlng the concepts included in this sectlon |
Dr. Charles E. Teeter formerly employed by the Chlcago Operatlons Office

at Argonne, Illinois, is now. afflllated with the Southeastern Massachusetts
Technological Institute, New Bedford, Mass. Through a consultantship arrange-
ment witfi Argonne‘National Laboratory, helis continuing to help guide the
organizatiofi and compilation of this catalog. h :

- We wish to acknowledge the dssistance of Mlss Ellen Thro in the prepara-

tion of this section.

J.HM.
June, 1966
PLAN OF CATALOG OF REACTOR CONCEPTS'

general_lntroduction

Part I.

Part IT.

Part III. -

Homogeneous and Quasi-homogeneous Reactors

Section

Section

Section

Section:

Section

Section

Section
Section
Section
Section

Section

Section |\

ISection

Section

Section

IO
IT.

ITI.

Vi,

I.
II.
III.

Particulate-fueled Reactors

Reactors fueled with Homogeneous
Aqueous Solutions and Slurries

Reactors Fueled with Molten-~salt
Solutions

Reactors Fueled with Liquid Metals

Reactors Fueled with Uranium'Hexa-
fluoride, Gases, or Plasmas

Solid Homogeneous Reactors

"Heterogeneous Reactors

Reactors Cooled by Liquid.Metals

. Gas-cooled Reactors

Organic-cooled Reactors

Boiling Reactors

Reactors Cooled by Supercritical Fluids

Water-cooled Reactors
Reactors Cooled by Other Fluids
Boiling-water Reactors

Pressurized-water Reactors

Miscellaneous Reactor Concepté

ANL-6892

ANL-6892
ANL-6909

ANL-7092
ANL-7138

ANL-7180

This report

REACTOR .CONCEPTS DESCRIBED IN THIS REPORT

Name of Reactor.-- LT .. Chapter No. Data Sheet No.

Table Model Reactor 2 1
Homogeneous Graphite Reactor : 2 2
ANCO 201 Nuclear Reactor 2 3
AGN-201-  Reactor - 2 4
AGN-211- - Reactor 2 5 -
Training, . Research, and Isotope

Production Reactor, General Atomic , :
(TRIGA).  Mark I and Mark II 2 6
TRIGA Mark III Reactor 2 7
TRIGA Mark F Reactor 2 .8
TRIGA Variation 2 9
Research Reactor, General Atomic, - ' 3 '
REGA 10-30 - . _ 2 - 10
Isotope Production Reactor, General .
Atomic (IRGA) L2 L. 11
Low-Power Research Reactor 2 12
Siemens Unterrichts (Teaching)

Reaktor (SUR) -2 13
Homogeneous Subcritical Assembly 2 14
First Daniels Experimental Power Pile 3 1
Second Daniels Experimental Power Pile 3 2
Third Daniels Experimental Power Pile 3 3
Impregnated Graphite, Nitrogen-cooled . , _

Reactor .- . 3 4
Gas-Cycle Reactor | : 3

High-températpre Bismuth-cooled Power

Breeder .. - 3 -6
Solid Homogeneous Reactor for Open-Cycle .

Gas Turbine -3 7
Solid Homogeneous Type Gas Cooled Nuclear '

Power Reactor S 3 o ' ' 8
Semi-homogéfieous High-temperature Gas-

cooled Breeder Reactor . . -3 ' _ 9
Semi-homogeneous Critical Experiment (SHE) 3 . i 10
Semi-homogeneous Gas Cooled Breeder

Reactor (SHR) : 3 o 11

Peach Bottom High-temperature Gas-cooled
Reactor (HTGR) o 3 12
REACTOR CONCEPTS DESCRIBED IN THIS REPORT (Cont.)

—

Name of Reactor’ , ) ' Chapter No. A Data Sheet No.

TARGET 1000-MW(e) Power Reactor

(Possible Modification) ; 3 - 13
OECD High Temperature Gas Cooled . ' o
Reactor Project (DRAGON) ‘ 3 ' 14
Experimental Beryllium Oxide Reactor : o
(EBOR) : 3 o : 15
Maritime Gas-cooled Reactor (MGCR) . | 3 ;‘ | . 16
Ultra High Temperature Reactor - ) o .
Experiment (UHTREX) (Formerly TURRET) . 3 . ' 17
Australian High Temperature Gas-cooled ' :‘
Reactor (HTGC) 3 - - 18
Solid Homogeneous Reactor for MHD | ‘
(Magnetohydrodynamic)’ Power Generation ‘ 3 . . .19
Boiling-water Reactor With Solid .
Homogeneous Core , , -3 S - 20
Semi-homogeneous Bismuth Cooled Breeder

Reactor (SHR) : : 3 21
Terrestial Unattended Reactor Power .
System (TURPS) : . '3 : o 22
Solid Homogeneous Reactor for Rocket or. o ‘ .

Aircraft Propulsion . 3 . 23
Solid Homogeneous Reactor Moderated .
with Zirconium Hydride 3 247
XMA-I Reactor . 3 25
Pratt & Whitney Aircraft Reactor-II ‘
(PWAR-1II) 3 26
Hydrogen-Cooled, Solid Homogeneous . '

Reactor for Rocket Propulsion 3 27
Tory IIA-1 Reactor 3 28
Tory IIC Reactor 3 29
'SNAP Experimental Reactor (SER) -3 30
SNAP-2 Development Reactor (S2DR) 3 31
SNAP-2 Reactor 3 32
SNAP-8 Reactor 3 233
SNAP-10 Reactor 3 34
SNAP-10A Reactor 3

35
Hydride Moderated Boiling Reactor , S R '
(SNAP-4) : _ 3 - 36

REACTOR CONCEPTS DESCRIBED IN THIS REPORT (Cont.)

Name of Reactor

KIWI-A Reactop
KIWI-B Reactor
NRX-A Reactor

Chapter No.
3
3
3

Data Sheet No.

37
38
39
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PART I. HOMOGENEQUS AND QUASI-HOMOGENEOUS REACTQRS

'SECTTION VI. ~ SOLID HOMOGENEOUS (SEMIHOMOGENEOUS) REACTORS =

Chapter 1. Introduction

The concepts described in this section are for reactors ln which,thewsolld
fuel . and moderator are in a unlform comblnatlon or mixture, e.g., a uniform,
dlSper51on of fuel in a moderator matrix, Excluded are reactors in wh1ch the
fuel and moderator are discrete, as in some closely related ones in whlch the
fuel and moderator form alternate layers or zones. Both these and the solid
homogeneous reactors have also been named semihomogeneous or reactors of a mixed
fuel- moderator.type The reactors in which the fuel and moderator are discrete
will be described in Part II under thelr respective methods of coollng Some
solld homogeneous reactors have been descrlbed in thlS Catalog in Part I, Sectlon
I, Partlculate Fueled Reactors o ‘

Reactors covered in this sectlon have been de51gned for research and testing
as well as for power productlon and breeding. The research and test reactors
are descrlbed in Chapter 2, the power and breeder reactors in Chapter 3.

Uranlum ox1des are a common form of fuel 1ncorporated with the moderator
into a fuel element; carbides, alloys, and the metal 1tself have also been used.
In some fuel-moderators, the moderator is impregnated, by soaking, with a solu-
tion of a salt such as uranyl'nitrate,: Subsequent heating converts this to an
oxide or carbide. 1In some elements, partlcles of the fuel and moderator are
cold pressed and sintered or hot pressed. 1In some concepts no detalls of the
fabr1cat10n have been considered. ‘Graphite berylllum oxide, berylllum, zir-
conium hydrlde and polyethylene have been moderators with graphite and
berylllum being most common for power reactors and polyethylene used in some

low-power reactors ' Most of the solid homogeneous reactors are gas-cooled.
232 233

‘Breedlng 1s chlefly by the Th ---U cy cle.

The advantages of the solid homogeneous reactor were p01nted out in 1960
by Charple and Perry,1 who stated, '"The hlgh ‘temperature homogeneous or semi-
homogeneous reactors are at once the most dlfflcult and, in all probablllty,1
the most rewardlng members of the gas -cooled reactor famaly The advantages
given include: high fuel-moderator volumes and large heat-transfer surface,
which lead to high power densities and specific power; high gas pressure made

possible by the compact cores; high temperatures of the exiting coolant gas,

12

which mlght be hlgh enough to permit direct- cycle operatlon good neutron
economy ; and the pOSSlblllty of low fuel- -cycle costs.

Problems cited by these authors as chief difficulties at.the time are
development of fuel elements with good mechanlcal integrity and the capacity
for retaining fisSion:products and the need for rapld refuellng to take ad-
vantage of the high SpElelc power. _

Solid homogeneous reactors were among the first gas-cooled types to be
considered in the United States. By 1944-45 Farrington Déniels had evolved
designs for pebble-bed reaéfors (see Chapter 1, Section I, Part I) ihhwhich
the fuel-moderator is pebbles of uranium carbidefland graphite. 1In 1944 he
proposed the use of beryllium oxide‘in-a‘higfi-temperature pile. He and his
co-workers at the Metallurgical Labdrapbry, University of Chicago,‘étudied
this concept. Later, a group at thé Clinton Laboratofies ‘(now. the Oak Ridge
National Laboratory) continued . deve10pment through 1947, when the concept was
abandoned. 2 The fuel-moderator in the 1947 conceépt was a dlsper81on of uranium
oxide in graphlte or beryllium oxide, in the form of ‘tubes. For almost ten
years after 1947, there was comparatively little work~done on this type of-
reactor in the U.S., partly because of intereét in water-cooled afid liquid-~
metal-cooled reactors and also because of.the view that the heat transfer |
properties of gases would limit power den51ty and specific power. Such develop-
ments as successful foreign experlence with gas-cooled reactors and the applica-
tion of new materials led to a renewed interest in the fi,S; in gas-cooled reactors,
including those of the solid homégenéOUS‘type,3 K .

Unlike most concepts previously discussed in Part I, the solid homogeneous
reactors have been developed into several commercial-size units. These reactors
have been considered partlcularly attractive because of the high steam tempera-
tures that are feasible with the homogeneous fuel moderator Bromlnent ¢urrent
developmentsAln the United States include the Peach Bottom (PenngyiVania).High-
temperature Gas-éooled Reactor (HTGR), the Ultra High Temperature Reactor .
Experiment (UHTREX), and the Experimental Beryllium Oxide-Reéctor (EBOR). Foreign
development has been extensive; in line with the earlier interest in gas-cooled
reactors. Programs in the United Kingdom, Australia, and Japén have included

solid homogeneous reactors. Several have been built, and others are planned.

-

<
13

References

R.A. Charpie and A.M. Perry, Gas-cooled Reactors--A Summary, Gas-cooled

Reactors.” A Symposium Sponsored Jointly by the Franklin Institute and the

American Nuclear Society, Delaware Valley Section, Monograph No. 7, Journal

" of The Franklin Institute, May 1960, pp. 9-11.

C.R. McCullough et él.,'Summary Report on Design and Development of High-
temperature Gas=cooled Power Pile, MonN-383, Clinton Laboratories, Sept. 15,
1947. Decl. May 10, 1957. |

Ref. 1, p. 4.

(4]

.

LY
15

Chapter 2. Research and Test Reactors

Reactors described in this chapter are low-poWer ones used to give experi-
mental data, primarily through the radiation they produce and for testing
reactor materials., 1In most, the fuel, an oxide of uran1um is dispersed in the
moderator (polyethylene, zirconium hydride, or graphite). Sofie of these reactors
have been widely used. ' ' |

An early‘(1952) model of a solid homogeneous research reactor was the Table
Model Reactor, described by Biehl, Hetrick, and Bennett. L The fuel-moderator
consists of very fine partlcles of uranium dioxide mixed homogeneously with
polyethylene and pressed into a mold. The core radius is 11.8 cm. ‘The authors
listed some advantages of a solid system over a fluid one: e.g., gas evolution
would be avoided if the moderator were sufficiently stable, because the solid
material would tend’to'trap fission fragments while it would not give off appre-
ciable amounts of decomposition gases. Also a smaller critical size, with smaller
core size and shielding, would be poseible° Tests showed that the uranium-
impregnated polyethylene would withstand long operation at high power. Poly~ -
ethylene has a somewhat hlgher hydrogen “atom.’ den31ty than ‘does water (7.8 x
10?2 nuclel/cm; versus 6.7 X 1022 for water), “

The Homogeneous Graphite Reactor, described in 1954 by Stelle,2 was intended
both as a research tool and a source of service irradiations to produce isotopes.’
Ite safety features aré a negative temperaturelcoefficient and control and safety
rods. The fuel, enriched uranium in U308’ is dispersed in graphite. Heavy water
circulates around the fuel-moderator as a coolant. A power of 135 kW(t) and an
average neutron flux oflO12 nentfons/cmz/secvwere planned.

The ANCO-201 reactor (1955) is a compact, low-power reactor designed for
training and research.3 The fuel-moderator is 207% enriched uranium dioxide
particles.dispefséd in discs of polyethylene. The originators, staff members
of the Applied Ndoleonics Corporation, stated that this reactor could be used
with only 450 g uranium-235 and it is safe enough to allow it to be installed
in highly populated areas. The maximum thermal neutron flux is 6 x 106
neutrons/cm /sec. ' |

The AGN-201 and -211 series of reactors of Aerojet-General Nucleonics
cloeely resemble each'other.é’5 "Many have been installed at universities.

.In 1956,'Biehl'et'a1.4 described AGN-201 as a compact; low-cost reactor
designed for maximum safety and high sensitlvityc The fuel, uranium dioxide

dispérsed in polyethlene, is in the form'of discs of different thicknesses.,
In addition to the negative temperature coefficient and rods for safety and
control; the reactor has an added safety feature. The core is divided into
two halves separated by a fuse of.polyethylene that has a high uraniufi-235
content. If a high flux level occurs, the fuse melts, causing the iower half
of the core to fall and separate from the other half. Of neglible power, these
reactors are cooled by natural convection, but they'can be modified to operate
at a maximum power of 20 MW(t). The core and a graphite reflector are in a tank
of water for shielding. - | |

The AGN-211 reactore’7 is similar in'many ways to the AGN-20l1. The fuel is’
a dispersion of uranium dioxide in polyethylene, but the fuel elements are feeF-
long sections, with a graphite reflectof at each end. The elements are clad with
polyethylene to prevent contact of the coolant (water) with the fuel. The reactor
has a power of 100 W(t). This reactor was considered desirabie for training ifi
nuclear engineering; for exampie, different arrangements of core 1attices mighfi
be tfied. | | :
| The TRIGA reectors (Training, Research, and Isotope Production Reactor,
General Atomic) introduced in 1958, have been widely used in the United States
and abroad for training, irradiatiofi, and isotope production. There are four
models; Mark I, Mark II, Mark III, and Mark F.8-14 All have fuel-moderator
elements of uranium mixed with zirconium hydride, and a coolant-shield of a pool
.of water. The fuel elements are cylindrical and are clad. Mark I, II, and III
have a graphite reflector and can operate either in the steady state or as
pulsed reactors: -In_steady-state operatioh, these feactors are low power, but
in puleing they can produce 200 MW(t). 1In the Mark F, specifically designed
for pulsing, the graphite reflector is eliminated; it has more fuel elements,
which QFovide higher heat capacity; and it has a large water tank, for more
versatility of radiation.10 A large prompf negative temperature coefficient of
reactivity is chiefly depended upon for safety, although control rods are pro-
vided. Advantages claimed for the reactors include: economy, versatility, and
safety, with no special containment building needed.

An extrapolation of TRIGA technology for a reactor.for pewering a bathy-
scaphe, research craft, or power buoy was suggested in 1960.15 |

. Two reactors with the same core as the TRIGA Mark I are General Atomic's

| REGA (Research Reactor, General Atomic) 10-30 and IRGA (Isotope Production
Reactor, General Atomic) reactors.8’13

In 1959, Thompson and Fahrnerl6.described a design for a low-power research

reactor in which the fuel-moderator is graphite impregnated with enriched uranium
17

as‘U308vand the coolant is heavy water. The.fue;‘elements are clad with aluminum
and filled w?th helium, with concentric tubes in each element for flow of coolant.
The power is 200 kW(t). The authors indicated problems would be the large amount
. of heavy water needed (about 400 gallons); the limitation on temperature by the
boiling of'the.heavy water; and difficulties in deéign of the cooling system, -
even at moderate pressures. | |

Siemenstchuckértwerke A.G. announced in 1961 plans to build a solid homo-
geleous reactor as a prototype for a model to be offered for sale.17 Few details

- were given. The reactor is fueled and moderated with a suspension of powdered

0,0

30g in polyethylene and has a graphite reflector and shielding of lead and water.

The -homogeneous subcritical subassembly18 designed by staff members at
Aerojet-General Nucleonics is included here because it is intended for use in its
own right to measure neutron fluxes, rather than as a step in designing another
.reactof. It has a cylindrical core of a homogenéouslmixture of 207% enriched
uranium and polyethylene and includes a meutron source. It can be used without

a reflector or with one of either graphite or polyethylene.
18

. THIS PAGE
WAS INTI:.NTIONALLY

LEFT BLANK

DATA SHEETS

RESEARCH AND TEST REACTORS

19
20

N N O T
SR N A .o
S L O PR

R P4 TR R s

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3 - y : '1 ?H l“‘ ’ ‘P : " ; “E y | ;2 h

WASINTENTIONALLY

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L ——— e e e — __\

No. 1-: Table Model .Reactor .

North. American -Aviation; Inc.. .

. Reference: TID-2503 (NAA-SR-Memo-352).

Originators: A.T. Biehl, D.C. Hetrick, and G.A. Bennett. - °

Status: Concept, June 1952,
Details: ‘Thermal.neiitrons, steady state, burner. Fuel-moderator: - '93% enriched

U as Uozbmixed-homogenebusly with polyethYLene.‘ Coolant:. - ambient: air. -Re- .=
flectotr: - BeO and:'graphite-;-'uUO2 particles, not more than* 10 g in size, mixed
with:powdered polyethylene and  pressed in“a mold. -Impregnation density: 50 mg
'U/cmgg: Core radius: 11.8'em. ' Control: ' negative temperaturé coefficient of
reactivity. - | |

Code: 0313..°. 16.-+ 31714 . 44 . 5932. - 711 84677 :921. 105

No. 2 Hdmogeneous Graphite Reactor

North American Aviation, Inc.

Reference: Science, 115, pp. 15-21, Jan. 1, 1954,

Originator: A.M. Stelle.

Status: Design, 1954,

Details: Thermal neutrons, steady state, burner. Fuel-moderator: enriched
U as U308’ ?niformly dispersed in graphite. Coolant: DZO' Core: Dblocks of
graphite - U308' DZO circulates around fuel-moderator. Control: negative
temperature coefficient; vertical safety and control rods of stainless-steel
tubes filled with B4C. . Reflector: 2 ft graphite around core.12Core ténk: )
cylinder 46 in. diam., 93 in. high. Average thermal flux: 10 neutrons/cm”/
"sec. Core will last 20,000 hours. Power: 135 kW(t).

Code: 0313 12 31102 4X 5932 711 84677 9xXX 105

81X11

'No. 3 ANCO 201 Nuclear Reactor

Applled Nucleonlcs Corp

‘Reference: NP-7578.

Originator: ' Staff members.

: Status: Preliminary design, '1955. | '
Details:. Thermal neutrons, steady state, boroef‘ 'Fuel-modefator' 450 g 20%-
enrlched uranlum as UO2 partlcles, 10 20 [, embedded in discs of radlatlon-‘
stabilized polyethylene Reflector graphite. ~Coolant: "~ ambient air. . Core:
right' circular cylinder, 9 in. diam.-andIS% in. high, with Al structure. Re-
flector surrounds core. Control 2 safety and 1 control rod contdin fuel--
reactor is subcr1t1ca1 when they are w1thdrawn, second control rod contalns Ccd
and operates in usual manner.  Neutron flux.- 6 x 106 neutrons/cm /sec Power:.
100 milliwatts. _ | o L
Code: 0313 ~ 16 _‘ 31714 43 5932 711 - 84677- - 921 105
| | o - 81112

83119
23

No. 4 AGN—ZQl-_. Reactor

Aerojet-General Nucleonics.
rd

References: Directory of Nuclear“Reacfdrs, III,'1960;.Nucleonics, 14, No. 9,

pp. 100-103, Sept. 1956.

Originators: Staff members.

Status: AGN-201-

: Startup :
No. Uwner L Location (dismantled)

-100 Naval Posggraduate.Schbol. : Monterey; Calif. 1956 |

-101 Catholic{Un&vgrsity. : . . Wash.,.-D.C. . 1957 ?;

-102 Oklahoma A&M - Stillw;ter oo

’ P1103 Aerojet-GengréLiNucleonics San Ramon, Calif. "

-104 U. of Akroé N Akron, O. ":: L "o
M-105 - Nationéi Néval Med. . Center Bethesda, Md. . = . 1957 (1962) .
' . New York University | New York ‘ 1964 -

-106 Tekas ASM = - " College Station 1957

-107 - U.nof'Utah ) '_' o -_ Salf'Lake.City e .‘ woo

108 aNL ' Argomme, T1I. w7
-109 Colorado State U, =~ Ft. Collins R "
-110 - U, of Palermo - * Palermo, Italy 1960

111 U. of Gefievd B o Geneva, Switzerland "

~112 U. of Calif. xr‘t - ""'fierkeiey S o 1957

-113 U. of Delaware ‘ Newark ' 1958

-114 Oregon State U.; .  Corvallis ' "

Details: Thermal neutrons, steady state, burner. Fuel-moderator: 20% enriched

U as U02 dispersed homogehebusly in polyethylene. Coolant: ambient air, natural
convection. Reflector: graphite. Core structure: fuel-moderator discs 10 in.

in diameter and of different thicknesses, four discs 1.5 cm thick, 3 discs..75

cm, and 2 discs .375 cm. Core and reflector in tank of H,O serving as shielding.

Control: safety and shim rods of polyethylene and UOZ; cire divided into two
halves cofinected by high-U235-content polyethylene fuse, which melts in case of
high flux level, causing lower half of core to fall 2 in.; negative temperature
coefficient of reactivity. Power: negligible except for AGN-201-P-103 (20 Watts)
and ACN-201-M-105 (5 Watts). '

Code: 0313 16 31714 43 5931 711 83119 921 105
84677
85XX9

Orig. Atoms for Peace Reactor, Rome & 2nd Geneva‘Cohf.

24

"No. 5 AGN-211- Reacpor

~ Aerojet-General Nucieonics

References: Direéto:y"dffiNqélear Reactors, III, 1960; Proc. Second U.N. Int.

Conf., 10, pp. 368-74. o

Originators: . Staff members. . ‘ P

Status: AGN-211- (all 4 are.idéntical) _ _
No. g ' ggggr ’ ' o Location A Startug
-100 U. of Baéel.(at Brussels Intl. Expos. 1958) Basel, Switzerland 1958
-101 Rice U. : ' | S Houston, Texas 1959
-102  uU. df’Oklahoma | | ‘ ' Norman : . 1958
-103  U. of West Va. D 'Morgafitown | 1959

Details: Thermal neutrons, steady state, burner. Fuel-moderator: 20% enriched

U as UO2 dispersed homogeneously in.polyethylene. Coolant: HZO' Reflector:

‘graphite. Fuel element: 12 in. long, approx.. 3 in. square fuel-moderator sec-

tion with 6 in. graphite reflector at each end. Element clad with bolyethylene
to prevent'contacf-between fuel and coolant. Euel elements surrounded by
graphite reflector elements. Core in tank of poél of.H20. Cdntrol; 2 Boral
safety rods, 1 Cd and 1 Al coarse control rods, and 1 stainless-steel fine
control rod. Power: 100 W(t). Specific power: 0.13 kW/kg.' Power density:
5 x 1072 kW/liter. | ‘ R |
Code: 0313 16 31101 43 5931 711 84677 921 105

' 81111

81112
25
No. 6 Training, Research, and Isotope Production Reactor,
General Atomic (TRIGA). Mark I and Mark II

oot -

General Atomic, Division of General Dynamics Corp. .

References: Brochure, General Atomié; 1958; bifectory of Nuclear Reactors, II,
pp. 223-26 ff.; Programming and Utilization of Research Reactors, 2, pp. 91-115;
NP-10188; NP-10714; Paper 57-AIF-3, 4th Ann. Conf. of AIF, 1957. ' o

Originators: Staff members.

Status: © ¥ Mark I reactors o '
— ~ . Power, .. Startup
Qwner - " Loctation - h kW(t) - (Dismantled)

Gen. Dynamics La Jolla, Calif. 250 1958
Omaha Vet. Admin. . Omaha,. Nebraska ' 18 . 1959

U. Arizona | Tucson . e ... 100 . ' 1958

U. Tekas - - Austin o . 250 1963 ‘
U. Minas Gerais ' Belo Horizonte, Brazil . 30 1960 °

U

. Lovanium " Leopoldville, Congo Republic 250 - 1959

Mark II. reactors

Beingfbuilt.

‘Columbia U. ; ~New York City - : 250
Gen. Dyn.-World . . o
Agricul. Fair ° San Diego ' ' 50 1960. (1960)

U. Illinois’ Urbana-Champaign 100 1960
Cornell U. . Ithaca, N.Y. 10 1963 .

- Kansas State U. . Manhattan . 10 1962
Ntl. Cmte. for | | ‘1_ i _ _ _
Nucl. Research Rome, Italy - 100 1960
J.Stefan Nucl.Inst. Ljubljana, Yugoslavia 250 being built
Musashi U. ' Kawasaki City, Japan 100 1962
Rikkyo U, Yokosuka City, Japan 100 1961
Korean AEC Seoul, Korea 100 1962
Inst. Nucl. Res. Dalat, Vietnam 250 1963
Inst. AE Bandung, Indonesia . 250 1964
U. Pavia Italy 250 planned
Vienna Poly. Inst. Vienna, Austria 100 1962
Inst. Tech. ‘ Helsinki, Finland 100 1962
U. Mainz Germany 100 being built
Details:

‘Mark I: Thermal neutrons, steady state (but can be pulsed), burner. For
training and irradiation. Fuel-moderator:

mixed with zirconium hydride. Coolant: H20.
flector: graphite.
cylindrical rods 28.4 in. long, 1.48 in. diameter.

uranium, 20% U235, homogeneously
Additional moderator: H0. Re-
Fuel-moderator elements clad in aluminum, in shape of
Elements arranged in circular

lattice in right cylindrical core, about 14 in. in diameter, 14 in. high. Core
surrounded by reflector set in tank or pool of coolant-moderator Hp0. Control:
B4,C powder clad with Al shim-safety, safety, and regulating rods; samarium burn-
able poison discs at each end of fuel elements; prompt negative temperature
coefficient of reactivity.

Mark II: Essentially similar to Mark I reactors.
31101 43 5981

Code: 0313 17
0323 13

84677
81111
81164

105

26

No. 7 TRIGA Mark III Reactor

General Atomie )

Reference: - GA-4339 Rev.

~ Originators: vStaff members.

Status: Reactors being built at Univ.'Califo:nie (Berkeley) and Natl. Inst.

Nuclear Energy: (Salazar, Mexico).

Details: Thermal neutrons, steedy state or puleed; burner.. Fuel-moderator:

U (20 w/o- U ) ZrH --8.5% U, 89.9% 2r, 1.6% H. Fuel-moderator clad with

stalnless‘steel Coolant HZO -Reflector: graphite, H20 " Core arrangement:

cyllnder of about 80 cyllndrlcal fuel-moderator elements (28 37 in. long, l.47

in. diam. ), 41 graphlte dummy elements, and control rods in pool of H20 Re-

flector consists of graphite end sections on fuel elements and radlal reflector

of H20 and of graphite in thermal- column position. Control. large prompt

negative temperature coefficient, borated graphite rods for shim, safety,
regulating, and transient control. :Power: steady state, 1 MW(t); pulsed,
200 MW(t). Neutron -flux (nedtrons/cmz/sec): .steady state -- 3 x 1013; pulsing:

transient bursts up to 1017 6 X 1016 with 10 msec pulse width.

Code: 0313 17 31101 43 5981 711  84677° 921 105

0323 13 B 81111

No. 8 TRIGA Mark F Reactor

General ‘Atomic -

Programming and Utilization of. Research Reactors,’

. Graphlte reflector ellmlnated

Core contains more fuel elements

2

1
2
3. Larger H
4. Entire core is movable

Reference: 2,vpp. 91-115..
Originators: ' Staff members. ’ ;
Status: S ST
R L - oo - ' Power; . ) .
~ Owmer . Locatlon _ CokW(t) . . Startup.
Gen. Dynamics La Jolla Calif. 1500 1960
Northrop Corp. . Hawthorne, Calif. 100 1963
Diamond Ordnance .
Fuze Lab.” Rad. 'National Bureau of
Facil. (US Army) Standards,
(DORF) Washington, D.C. 100 1961
Armed Forces Radio- | | |
biol. Res. Inst.
(DASA) .
(AFRRI) Bethesda, Md 100 ; 1962
Details: Thermal neutrons, pulsing, burner. Built specifically for pulsing
purposes. D1fferences from Mark I: . '

with higher heat capacity.

0 tank for more versatility of irradiations.

Al support structure hung from.

carriage that moves on rails on reactor-room floor.

Mark F prototype can be pulsed to about 1500 MW(t).

Reactivity insertion '
: Max;ޤ6Wer level

Prompt -energy release

Min., pulse width

Min., period

Max. repetition rate

Shorter neutron lifetime in H

2
and higher max. power.
Code: 0323 17 . 31101 43 5981
13

*
Now the Harry Diamond Laboratories

Pulsing pafameéers&u
2.2% Bk/k | '
2000 MW

24 MW - sec | ’

10 msec

3 msec

12 per hr

O-reflected reactor allows smaller pulse width

84677 921 105
81111
81166

27
28

No. 9 TRIGA Variation -

General Atomic

Reference: = Nucleonics, -18, No. 12, p. 31, De;;_L960.

Status: . Proposal, 1960.

Originators: . Staff members.

Details: Extrapolation of TRIGA technology‘for small reactor producing 25-50
kW(e) to deliver 35-70 hp for powering a bathyscaphe, research craft, or bower
buoy for the U.S. Navy. | o | '
Code: 0313 17 31101 43 . 5981 711 84677 - 921 105

' 13 o | 81111 |

. No. 10 Research Reactor; General Atomic REGA 10-30

General Atomic

Reference: Paper 57-AIF-3, 4th Annual Conf., ATF.

Originators: Staff members.

Status: Design, 1957. |

Details: Core identical with TRiGA-Mark I core. _

Code: 0313 17 31101 43 5981 - 711 84677 921 105
' 13 - 81111

29

No. 11 1Isotope Production-Réactor,~General¢Atomic (IRGA)

General Atomic.

Reference: Brochure, General Atomic; Paper 57-AIF-3,- 4th Annual Conf.,. AIF. B

Originators: Staff members.

Status: Design, 1957. -

Details: Copé;iQentical with TRIGA-Mark I core.., - . . | .

Code: - 0313 17, 31101 _ 43 . 5981  711.. 84677 921 105
| 13 . 81111

‘No. 12 Low-Power Research Reactor

North American Aviation, Inc.

Reference: NAA-SR-34. _ ‘ S L .

Originators: A.S. Thompson and T. Fahrner.

Status: Design, 1959. ‘

Details: * Thermal neutfdns,'steédy state, burner. Fuél-moderator: éraphité
impregnated with enriched U ‘as USOé. Coolant: 'DZO. Reflector: - graphite.': |
Core: ~hexagonal prism 4% ft high. 37 Al-clad, helium-filled fuel elements;

" edch contain hexagonally shaped fuel-moderator blocks. Alternatively, fuel-
moderator blocks could be cylindrical or wedge-shaped. Coolant passes through
concentric Al tubes located axially in each fuel element; inlet temp. 142°F,
outlet 156°F. Control and safety rbds in annular gap between core tank and

reflector. Reflector, 30 in. thick graphite, surrounds core. Power: 200 kW(t).

Code: 0313 . 12 31102 4X 5931 711 8111X 921 105
No. 13 Siemens Unterrichts (Teaching) Reaktor (SUR)

Siemens-Schuckertwerke A.G.

Reference: At. Ind. Forum Memo, 8, No. 11, p.- 14, Nov. 1961,

Orignator: Staff members.

Status: Construction proposed, 1961, _
Details: Thermal nefitrons, burner. Fuel-moderator: U308 susPended in
polyethylene. Reflector: graphite. Shielding: Pb, HZO' Power: 0.1, 1,
or 10.W. , ’ ,

Code: 0313 16  3XXXX  4X 593X 711  8XXXX 92X 10X

No. 14 Homogeneous Subcritical Assembly

Aerojet-Genefal.Nucleonics

Reference: TID-7619, pp. 80-87.

Originators: Staff members.
Status: Two in 6peration, 1961 at The University of Te#as and Pensacola Junior
College, Florida. _ - _
Details: Thermal neutrons, subcriticallaséembly. Fuel-moderator: homogeneous
mixture of polyethylene and 20%-enriched U as oxide. Coolant: ambient air.
Reflector: graphite, polyethyléne,_or none. Neutron source: Pu-Be in core or
an accelerator target source., Core: right cylinder. Purpose of assembly: to
measure neutron fluxes.
Code: 0333 16 31714 43 5932 711 8XXXX 911 105

‘ ' 921 .
10.

11.

12.

31

References

. "A.T. Biehl, D.C. Hetrick, and G.A. Bennett, A Tablé'Mbdel‘Reactor, Nucl.

Sci. Technol. (Extracts from Redactor Sci. Technol., .2, Nos. .1-4,. Apr:.-

Dec. 1952), TID-2503 (NAA-SR-Memo-352).

A.M. Stelle, The Engineering Design of the North American Aviation

Homogeneous Graphite Research Reactor, Science, 115, pp. 15-21, Jan. 1,

1954.

The ANCO 201 Nuclear Reactor. - Preliminary Design, NP-7578, Applied
Nucleonics Corp., Nov. 15, 1955,

A.T. Biehl, F. Fayram, R. Henley, G.A. Linenberger, H. Montgomery, R.

"Mainhardfi; G. Néwby; J.'Randali; D. Sawle, and J. Sherwin, Cofipact, Low-

Cost Research Reactor Efiphasizes Safety; Nucleonics,:l&; No. 9, pp. 100-
103, Sept. 1956. o ' -

Directory of Nuclear Reactors, Vol. III, Research, Test and Experimental

Reactors, IAEA, Vienna, 1960, pp. 205-209 ff,
Tbid., pp. 239 ff.

A.T. Biehl, R.P. Geckler, M. Harvey, and R. Henley, The AGN-211 Homo-

geneous Polyethylene-UO, Pool-Type Reactor, Proc. Second U.N. Int. Conf.

2
on Peaceful Uses of Atomic Energy, 10, pp. 368-74, United Nations, N.Y.,

1958.
Brochure, General Atomic, Division of General Dynamics Cofp., 1958.

Directory of Nuclear Reactors, Vol. II, Research, Test and Experimental

Reactors, IAEA, Vienna, 1960, pp. 223-26 ff.

Frederic de Hoffmann, Research Programs with the TRIGA Reactors, Program-

i .
ming and Utilization of Research Reactors, Proceedings of a Symposium held

in Vienna, October 16-21, 1961, Vol. 2, pp. 91-115, Academic Press, New
York, 1962.

Hazards Summary Report for Diamond Ordnance Radiation Facility, NP-10188,

Dept. of the Army, Ordnance Corps, 1960,

Final Safeguards Report for Armed Forces Radiobiology Research Institute,

NP-10714, National Naval Medical Center, Sept. 1961.
32

13,

14,

15.

16.

17.

18.

Frederic de Hoffmann, A New, Safe, Research and Isotope Reactor, Paper

57-AIF-3, The 1957 Nuclear Industry -- Problems and Progress, 4th Annual

Conference of Atomic Industrial Forum, Oct. 28-31, 1957.

- TRIGA, Mérk'IIIAReactor Description, GA-4339 Rev., General Atomic, Division

of General Dynamics, Dec. 1963.

" Nucleonics, 18, No. 12, p. 31, Dec. 1960.

A.S. Thompson and T. Fahrner, Design Study of a Low-Power Research

Reactor, NAA-SR-34, North American Aviation, Inc., Oct. 14, 1959,

At., Ind. Forum Memo, 8, No. 11, p. 14, Nov. 1961.

N. Little, A Homogeneous Enriched Assembly, Proceedings of the University

Subcritical Assemblies Conference Held at Gatlinburg, Tennessee, August

28-30, 1961, TID-7619, pp. 80-87, USAEC, Jan. 1961.

33

Chapter 3. Power and Breeder Reactors

Most of the concepts described in this chapter are for higthémperaturé, f
gas-cooled reactors, some of which have been developed into Iarge-scaie plafifs.
A few are liquid-cooled reactors for special'purposes, such as the SNAP reactors.
Typicaily, the reactor core is composed of fuel-moderator elements around which
the coolant flows. | ‘

. Gréphite or beryllium oxide is the moderator for nearly ali power reactors,
with graphite being used for most. Beryllium metal was suggested early.in the
development of solid homogeneous reactors, but it has been little used. The
advantages of gfaphite and those of berYllium oxide have been.given by several
afithors. Faris1 discuséed'graphite. Jaye et al.,2 and Roberts,3 gave the ad-
Qéntages of beryllium oxide. Desirable characteristics listed for graphite as
a moderator in solid homogeneous reactors are: a negafive temperature coef-
'fiCient, which'givesrsafety in‘low;temperatufe applications; desirable mechanical
prOpertiés at high temperatures; comparatively simple decohtaminatiofi; and.
simplicity of fabrication and récovery processes. Many advantages are claimed
for beryllium oxide. Because it is a better moderator than'grqphite, conver-
sion ratios are higher afid cores are smaller fiith beryllium oxide than with
graphiteifdr the same fuel load and specific power. Higher‘temperatures for
the gas coolant leaving the reactor are poésible; thus more effiéient steam
cycles can be used and the temperatures might be high enough for efficient gas
turbines. Beryllium oxide is compatible with carbon dioxide at high temperatures,
so that this gas can be used instead of helium. With carbon dioxide there are
fewe; problems with gas-tightness in the system. The fairly high thermal con-
ductivity of beryllium oxide makes it suitable for solid homogeneous fuels. It
resists release of fission products from the fuel-moderator, and it fiay have a
Along lifetime without serious loss of essential properties. Jaye et 51.2 have
suggested that a berylliufiuoxide spine might be used in a moderator element of
graphite. One obvious major difference between the two moderators is the con-

siderably higher cost of beryllium oxide.
Gas-cooled Reactors

" Studies of the use of beryllium oxide in a high-temperature power reactor
were begun by Farrington Daniels in 1944 at the Metallurgical Laboratory,
University of Chicago. Calculations by Daniels and co-workers established the

feasibility of such a reactor, and problems in.its design were investigated at
34

the Metallurgical Laboratory until 1946, when this work was transferred to a
group at the Clinton Laboratories. An aim of the program was developing a high-
temperature, helium-cooled, powér reactor fueled with enriched uranium and
moderated with beryllium oxide. A boiler for steam generatiofi and a generator
to produce electrical power were also included ifi the specifications for the
reactor plant. Emphasis was put on a design that could be developed in a fairly
short time. ‘ , _

In November 1946, a preliminary design was proposed,5 -In this 40 MW(t)
reactor, the fuel~m§derator is in the form of enriched uranium oxide mixed with
beryllium oxide to form cylinders, which are within axial holes in hexagonal
prisms of beryllium oxide. These prisms form a cylindrical core. A reflector?
a thorium breeder-blanket, and the core vessel surround the core. The coolant,
helium, flofis inté the bottom of the reactor, through the core, and out the top,
whence it passes to a boiler.

At the suggestion of Wigner, a group studied a reactor utilizing beryllium
metal instead of beryllium oxide, This study was published in February 1947.6
The design is very similar to the earlier one with beryllium oxide. Plates of
beryllium-uranium alloy are contained in a skeleton of beryllium metal. Fertile
material, thorium metal, is in channels in the side and end reflectors of beryl-
lium metal. The power is the same as for the oxide reactor, 40 MW(t). The
results showed no outstanding differences between the moderators that would make
the metal preferable. Also, the metal has such disadvantages as poor tensile
and creep properties at high temperatures, and the knowledge of fabrication tech-
niques for it was limited. Thus it was decided that use of the metal would give
no advantage, and the development of a beryllium oxide pile was continued,

Investigations showed that reducing_the volume of the pile would give a
considerably smaller critical mass and would increase the conversion ratio. Also,
tests on beryllium oxide indicated loss of thermal conductivity under irradiation,
causing unqertainfies in performance of the pile. It was thus decided to design
a smaller, lower-power pile to give a better facility for study of such problems.
In June 1947 the USAEC notified the Laboratory that it would not authorize de-
sign or construction of the power pile, but investigations on problems should
continue at a lower priority,4 In September of the same year, the Power Pile
Division (the Clinton Laboratory group assigned to the project), issued a final
report summarizing the status of the developmént:.4 The design advanced in this
report is generally similar to that for the earlier design with beryllium oxide.

The power is lower, 12-20 MW(t), and beryllium metal or graphite are considered

35

as alternatives to beryllium oxide. Concepts for higher power were ¢onsidered,
as was a design for a horizontal reactor, intended to facilitate loading and
unloading of elements. These concepts differed from the 12-20 MW(t) concept
~chiefly in:dimensions; fuel loadings, and engineering.

,A 1950 concept by Daniels, the Impregnated Graphite, Nitrogeén-Cooled Reactor,
was designed to use the étandard“equipment and materials available in 1950.7 The
author said that such a power pile could be'builtrquiekly and at low cost. It
'was intended to be a "one-shot' experimental reactor with expendable parts. The
fuel-moderator is uranium oxide disPersed>in'graphite in the form of blocks that
make up the core,. Nitrogen was recommended as a coolant gas because it could be
used to run standard turbineS‘ but helium could be used. The reactor was de-

signed as a breeder, with fully enriched uranium to be used for breeding U 233

" and sllghtly enriched uranium for breeding plutonium. For breedlng U 33, thorium
nitrate would be in the fuel-modérator or in the graphite reflector. ‘Plutonium
would be bred:from the U238 in the fuel. The coolant flows from the top of the
core to the bottom through channels in the graphite blocks and exits at 1100°F
The maximum power is '11.5 MW(t).

Another concept by the same author in 1956°'was the Gas-Cycle‘Reacror,Sfin
which the ffiel-moderator coesiSts of uranium carbide, which could be in dif-
ferent forms, such as rods or pellets. This reactor was also designed for high
temperatures, with en exit cpoiant temperature of 1350°F. ‘'The power is 20 MW(t).
| Slegel in 1951, suggested a power breeder reactor in which the fuel-

“ moderatdr is graphite impregnated with uranium dicarbide. 2 The coolant, helium
or liquid bismuth, passes through ‘axial coolant channels. Thorium fertile |
material is in either the core or a breeder blanket. The power is 1400 MW(t).

Investigations at the Studebaker-Packard Corporation, repOrfed by Thompson
in 1956;10 resulted in a design for a gas-eodled reactor, with solid homogeneodus
fuel-moderator, intended for use with a gas turbine power plant...The fuel -
moderator is a ceramic of uranium dioxide and graphite. Additional moderation.
is provided in the roughly spherical core by a central island of graphite or
beryllium and a reflector of the same material. The fuel, ‘with passages for
flow of the heélium coolant, is in the annulus between the island and thé re-
flector. The power 'is 60 MW(t). " |

© In a 1960 patent, applied for in 1956,11 Fortescue and Lockett described
a reactor with solid homogenéous fuel-moderator elements. The materials of -
the “eléments are not givefi, nor is ‘the 'gas cooldnt. The Ffuel elements are sur-

rounded by a reflector of graphite. The coolant gas passeés downward between

36

the feflector and the pressure shell of the reactor, then upward between core
elements.
In Japan, work on 'semi-homogeneous' reactors (which utilize solid homo-
géneous fuel-moderators) inciudés designs for two power reactors. For one design,
a critical experiment has been operated.
Inoue et al,12 in 1958 proposed that enriched uranium oxide, impervious
graphite, and thorium oxide be incorporated into elements for a breeder reactor
with a powér of 600 MW(t). A blanket of thorium metal surrounds the core. Few
details wefe given for the structure of this reactor, which is cooled by carbon
dioxide, The outlet temperature of the coolant gas (700°C), according to the
authors, would make possible,fise of the reactor with a gas or steam turbine.
The Semi-homogeneous Critical Experiment (SHE)]"'}“15 Was-reported in 1961
by Inoue:-and co-workers. Thé fuel-moderator is a uniform mixture of 20%-enriched
UO2 and graphite, formed into discs. . The discs -are placed, within a graphite
tube, in sequence with discs of graphite to form rods. The rods, in hexagonal
horizontal arrangement, are in two halves of the assembly. The halves are
brought together for criticality. The experiment operates at room temperature.
Discs of other compositions, e.g., thorium oxide for breeding, can be added to .
vary theAcore composition. Control rods are provided. 'The first loading was
in 1961; and operation of the expefiment was continuing in 1964. .
~ The Semi-homogeneous Gas Cooled Breeder Reactor (SHR)16 is very similar to
the critical experiment, but ‘it is designed for a power of 25 MW(t) and is cooled
by helium;..The fuel-moderator, UO2 6r UCvaarticles smaller than 6 p evenly dis-
persed in graphite, is in the form of pellets. They are contained in an imper-
vious graphite sheath. A blanket of thorium surrounds the core.
The Peach Bottom (Pennsylvania)-High-temperature Gas-cooled Reactor
(urer) /2%
temperatures.for the gas coolant (1360°F or higher); high fuel burnup; high

utilizes a design intended to have the advantages of high outlet

power density; and good neutron economy. The Peach Bottom reactor is intended
as a prototype for reactors of this kind.

The breliminary design and development.of the HTGR began early in 1957 at
General Atomic Divisibn, General Dynamics. The project'is supported by High
Temperature Reactor Develoément Associates, a group of 52 private utility
companies. Research and development is being carried out by General Atomic
under a contract with the USAEC. . The Philadelghia Electric Company is owner
and operator of the plant. Construction began in 1962. Fuel loading. and start-

up were scheduled for early 1965 but were delayed by a fire in the containment
. Reactor, Gas-cooled,; Exploiting Thorium (TARGET).

37

shell;zé The reactor went critical on March.31, '1966. Power operation is now

scheduled for late '1966.° The reactor, designed as a converter; utilizes U235

as initial fuel and willuuse.U233 after the breeding cycle has begun..' Helium-
is the_coolant, and thorium dicarbide is the fertile material. The cylindrical
fuel-moderator element is of compound structure, with graphite as moderator,
cladding, and. fuel matrix containing the fuel and fertile material in homo-
geneous dispersion. The fuel-fertile material is a mixture of uranium and
thdridm‘carbides,_as particles coated with pyrolytic carbon to control escape
of fission products. There is an upper reflector, a fuel-bearing middle section,
a bottom reflector, and an internal fission-product trap. The fuel elements are
aligned vertically in a.triangular pitch. The coolant gas flows upward in spaces
between fuel elements. The power is 115 MW(t); 40 MwW(e).

Advanced forms of the Peach Bottom reactor are under consideration 2525 a
300- to SOO-MW(e) reactor is the subject of a fesearch and development program
being carried out jointly by General Atomic and the Empire.State Atomic Develop-
ment Associate, Inc. (ESADA). -With the same type of ‘fuel elements as are in the
Peach Bottom plant, the advanced deéign is for a reactor to produce steam for a
2400 psi, 1000°F superheat, 1000°F reheat steam cycle. Development and testing
on extending fuel life are also in progress.. . A modified form of the Peach Bot-
tom fuel is being evaluated for a 1000-MW(e) reactor plant, the Thermal, Advanced
19,25 'The-prografi is being
carried out by General Atomic under contract to the USAEC. This concept is being

considered for possible construction in the 1970's. The reactor would have about

the same surface temperatures as in the Peach Bottom prototype, but a more ef-

- ficient steam cycle would be used, to give a steam pressure of 3500 1b psi, super-

heat to 1050°F, and two reheats, each to 1000°F. In the most highly developed
design,25 the fuel, uranium carbide, is a bed of particles within a graphite
container. Thus it is not a solid homogeneous reactor. In a suggested alterna-

tive, features .are the same except that the fuel-moderator-fertile material

‘consists of microscopic particles (1 p or less in diameter) dispersed in a
graphite matrix. Still another possibility being evaluated is operation with

‘recycled U2v33 as fuel and thorium fertile material in the fuel-moderator elements

to achieve core breeding.

The OECD High Temperature Gas-Cooled Reactor Project (DRAGON),ZE'-28 had its

origin in abodt:l956, when a team of the United Kingdom Atomic Energy Authority,

~under Fortescue, began preliminary studies. The first report was- the 1958 paper

by Shepherd et a1.267.1h‘thatvyéar, a committee was set up under the auspices of
38

the OECD (Organization for Economic Cooperation and Development) to study col-
laboration among members of the organization, with a high~temperature gas-cooled
reactor chosen for study. -In 1959 an agreement was concluded by Austria;
Denmark, Euratom, Norway, Sweden, Switzerland, and the United Kingdom for work

on this type of reactor. Development of the reactor experiment is carried out

" at Winfrith Heath, England, by staff members from the parent organizations.

The objectives of this reactor development are those of others utilizing solid
homogeneous fuel, including the attainment of exit-gas temperatures (1382°F)
higher than possible before.

The core has two zones, inner and outer, It is so divided because it is
difficult to test, in a smali‘gore that .has much higher neutron leakage, units -
comparable with those of a large reactor. In the inner zone the fuel, uranium

carbide,'is diluted with thorium carbide; in the outer it is diluted with zir-

conium carbide. The fuel-moderator is in the form of particles, which are

pressed into cofipacts that are inserted intb graphite fuel tubes. Seven tubes
make up a,fuel-element. The core, which is made up of 37 of these fuel elements
in a triangular lattice, is surrounded by a graphite reflector.l The coolant
gas, helium, passes upward through the core. The power is 20 MW(t).
| The reactor became critical in August.1964° I

The Experimental Beryllium Oxide Reactor (EBOR) is the land-based proto-
type for the Maritime Gas-cooled Reactor (MGCR), which is to be of higher power,
but otherwise based on E].A:’,OR.zg_31 The purpdse is to demonstrate the technology
of beryllium oxide as moderator for gas-cooled reactors’to be used in central
station and marine power plants. EBOR, which was designed by the staff of General
Atomics, is at the National Reactor Testing Station in Idaho. The fuel-moderator
consists of compacts of highly enriched uranium dioxide and beryllium oxide, which
are contained in Hastelloy-X fuel pins. The pins, which are arranged cylindri-
callj, are surrounded by blocks of beryllium oxide. The coolant gas, helium,
flows into and out of the reactor through two concentric passages at the side of
the reactor,. The exit temperature is 1300°F. The design of EBOR is for 10 MW(t).

32-37 formerly termed

The Ultra High Temperature Reactor Experiment (UHTREX),
TURRET, is being developed at Los Alamos Scientific Laboratory, with completion
scheduled for 1966. The design is for a coolant exit temperature (2400°F) appre-
ciably higherflthan any previouély used. The design (as TURRET) was originally
suggested for electrical generation with a high-temperature gas turbine, but-:it
was later chanéed to an'experifiental prototype for a process-heat reactor

operating at an unusually high temperature for the exit gas. One purpose was
39

to investigate operation at high'temperature e. g',“the hehaviorporiunclad porous
‘graphite fuel elements at'up to 3000°F. The fuel moderator consists of hollow
-graphlte cyllnders contalnlng uranium carblde as mlcrospheres coated with pyro-
lytic carbon, which delays escape of f1551on products 'The fuel-moderator
elements -are arranged in rows within channels in the cyllndrlcal graphlte core.,
‘.Mechanlcal rotation of the core over loadlng rams permlts reloadlng the core
during 0perat1on at full power. Helium coolant enters the core at the bottom
and, after the flow is divided into fuel channels leaves at the top for dis-
'charge The power for this experlment is 3 MW(t) - |

The use of beryllium oxide ‘as moderator in a high- temperature pebble -bed
reactor (the HTGC, High Temperature Gas Cooled Reactor) is being investigated
by workers of the Australian Atomic Energy Commission._,'38'-'llLl Beryllfum oxide
is the matrix for a solid ceramic fuel-moderator-fertile material, in which
plutonium and thorium oxides are dispersed{-MIt was designed to incorporate
the advantages of this moderator, which have been previously discussed in this
chapter. The pebble-bed concept was chosen because it permits high burnup (by
" recirculation of the pebbles) and because ho neutron absorbers need be ifcluded
in the fuel-moderator. The study of beryllium-moderated, high-temperature,
gas-cooled reactors was begun by the Australian.AEC'in 1960, and the concept
is being assesseo now for plutonium fuel; Uz'33 and U235 fuels will be considered
later. Currently, the design is for a cyllndrical core vessel containing the
pebbles, which are slowly added at the top'and’removed at the bottom. The car-
bon dioxide coolant flows up between the core ‘vessel and the pressure vessel to
the’ top then downward through’ the core, 1eav1ng the bottom at 700 800°C. The
power density is 10-20 kW/liter. '

A brief description of a helium-co6léd solid homogeneous reactor was
published in 1963 by Sternglass ét‘éi.42 :The reactor is part of a'systefi
designed to utilize MHD (magrnetohydrodynamic¢) power generation. 1In the 264-
MW(e) reactor, the core is a graphite cyllnder in which fuel-moderator elements,
of uranium carblde unlformly dlstrlbuted in graphlte are contained W1th1n
vertical channels. The homogeneous fuel-moderator, which is unclad, was chosen

because the high temperatures required by the MHD unit would prevent:the use‘of

clad fuel.
Other Reactors for Electrical Power -

Although most solid homogeneous reactors for electrical power are gas-

cooled, some concepts have been advanéed for reaétors cooled by other fluids.
40

In a 1960 United Kingdom Patent,43 applied for in 1956, the fuel, moderator,
and fertile material are combined as a mixture to form fuel rods. A typical mix-

233

ture would be U ""0,, BeO, and ThO2° Boiling water is the coolant, but because

’
only enough of it ti cool the fuel elements is used, it does not act as moderator.
In an alternative to the gas-cooled Japanese Semi-homogeneous Gas Cooled
Breeder Reactor described earlier,16 Inoue et -al. suggested that liquid bismuth,

rather than helium, be used as coolant.

A low-power [about 100 kW(e) ] reactor is being developed at the Martin

Nuclear Division of the Martin-Marietta Company for a direct-conversion system,

44,45 The system is for a

Terrestial Unattended Reactor Power System (TURPS).
self-regulating constant heat sourcevat about 1000°F. 1In an early report by
Murphy, thg coolant would be phosphorus sesquisulfide, P483; and alternatives
were given for fuel and moderator: wuranium dioxide as fuel and water as mod-
erator, or a fuel-moderator of Z;UHX. Boiling P483 was chosen as coolan; be-
cause this compound has low neutron absorption and resists radiation; little

is known, however, of its flow and heat-transfer characteristics. The criteria
were for a ldw-cost, portable, system of medium size that would be reliable, be
simple toIOperate, and have low capital and operating costs. A plant with a net
power of 1.3 MW(e) was originaliy'proposed. A few details of a more recent
design have begn published. The coolant is the same, P453. The fuel-moderator
element has two segments: one is an alloy of uranium and zirconium hydride; the
other is zirconium hydride alone. 1In the.configuration developed, hydrogen can
migrate from the unfueled segment to the fueled segment to act as moderator.
Control could be by reduction of heat load in the unfueled section, which would
reduce hydrogen migration and thus reduce activity. The coolant circulates By

natural convection in a boiling-condensing cycle. A contract for investigation

of the concept has been awarded by the United States Air Force.
Reactors for Aircraft and Space Vehicles

Several reactors designed for use in airc}aft, as power sources for space
propulsion,;or as auxiliary power for space vehicles utilize solid homogeneous
fuel. Some of the concepts have been proposed for aircraft, while the Tory,
SNAP, Kiwi, Phoebus, and NERVA reactors have been prominent in the deveiopment
of space vehicles. ' - |

In 1959 patents, filed in 1949;46’47 Grebe described a reactor for propul-
sion in which the porous fuel-moderator element is in the form of porous seg-

ments of fuel-moderator, separated by coolant passages. Because of the structure

41

]

of this core, the name '"Cabbage Head'" has' been applied to it.481-The segments

, converge‘to close at one end and are joined to adjacent segments at the other
end. The segments bow outwardly, the bowing increasing as the distance from
the center increases, so that a roughly spherical or pear-shaped core is formed.
Gas flows into passages, through the porous segments, out coolant channels, and
through a jet nozzle. The patent claims that the design permits close contact
of the coolant with fissionable material and the flow of a large volume of gés
without excessive pressure loss. In a modification, the use of different mod-
erator matrix materiéls in different zones was proposed. A refractory moderator,
such as graphite, would be used in the hot zone near the nozzle, and a.less-
refractory moderator, of better nuclear properties, would be used in-the cooler
inlet zone.

An early (1952) concept-by staff members at the Bendix'Corporation'isfifor
a low-power water-cooled reactor in which U235 is distributed in zirconium
hydride.49

A design for a high-temfierature‘reactor utilizing a ceramic of highly
enriched uranium dioxide in beryllium oxide with yttrium hydride (which has- a
much higher hydrogen content than zirconium hydride) as additional moderator
was proposed by Leeth in 1958,50 The U02-Be0 and the yttrium hydride are
arranged homogeneously in concentric polygons. The coolant, air, has an exit’
temperature of between 1800°F and 2000°F.

-In 1960, Levengood, Dissler, and Kalinowski described a reactor, the PWAR-
II'(Pratt & Whitney Aircraft Reactor - II), in which the fuel-moderator is
uranium dioxide in beryllium oxide.s1 Lithium coolant flows around the hexa-
gonal containers for the fuel-moderator pins. Beryllium oxide is also the-
reflector. Designs for 510 MW(t) and 200 MW(t) reactors were given.

Newgard and Leval described in 1960 a design for rocket propulsion in
which uranium-loaded graphite plates form a core through which hydrogefi propel-
lant flows.52 The reflector is made up of beryllium oxide blocks, .and control
is by movement of control plates in the reflector.

Aé part of the PLUTO program.to develop a reactor for a- nuclear ramjet
engine, two solid homogeneous reactors, Tory IIA-1 and Tory IIC, were developed

53, 5 The reactors are similar in design. In both,

and successfully tested.
the cylindrical core consists of bundles of many fuel-moderator elements (100,000
for Tory IIA-1 and several hundred thousand for Tory IIC) of highly enriched
uranium dioxide homogeneously mixed with beryllium oxide. A graphite reflector,

with rotating cylinders contadining poison rods, is used for part of the control.

42

Cooling is by single pass, straight-through flow of air from inlet, -through the
reactor, and out the exhaust. :

Tory IIA-1 first operated on Maf 14, 1961. Full power, 155 MW(t), opera-
tion took place later in 1961l. The operation of Tory IIC (May 20, 1964) at |
fulllpower was also successful. 1In that yeaf, emphasis on the project. was
reduced because of the decision that the work by the USAEC was essentially
completed. The reactors were inactivated.

Work on the SNAP (Systems for Nuclear Auxiliary Power)SS.60 reactors
originated in the need in advanced space vehicles for compact, long-lived, and
high-power nuclear sources of electrical power. Atomics International, a
division of North Amefican Aviation,'Inc.,~began_investigating the problem in
1953, and by 1955 the type of reactor was chosen. The bases'used in the choice
included minimum weight and suitable operating temperature. 1In the general
reactor design, the fuel is U235, moderated with zirconium hydride. The cool-‘
ant is a 1iquia metal, which transfers heat to a mercury Rankine cycle or a
direct-conversion thermoelectric unit for power conversion. In 1956, the
USAEC established the SNAP program.

A homogeneous thermal reactor with highly enriched fuel and a reflector.
was chosen for several reasons. The homogeneous core with enriched fuel per-
mits small size and efficient use of the fuel. A thermal reactor is preferable
to a fast reactor because of the lower uranium inventory and less-complex
control problems. Adding a reflectoer: (beryllium)‘reduces the minimum critical
mass 6f uranium in the core. The:use of the zirconium hydride as moderator
also gives a smaller core size., To restrain thermal dissociation of this
hydride, it is clafi with a material that is a barrier to diffusion of hydrogen.
Control of the reactor is aécomplished'by varying the thickness or effectiveness
of the reflector, or‘By remdving it to make the reactor subcritical.

The general design was useq‘in a series of reactors: SNAP Experimental.
Reactor (SER); SNAP-2 Development Reactor (SDR or S2DR); SNAP-2; SNAP-4; SNAP-8;
SNAP-10; and SNAP-10A. The first two were preliminary developments for the
SNAP-2, SNAP-8, SNAP-10, and SNAP-10A reactors. Both operated for test periods,
SNAP-2 was intended as a low-power [3 kW(e) ] reactor for space application for
use with a mercury- Rankine power convérsion cycle; it was a predecessor of -
SNAP-8. This project was redirected. in 1963, although the Rankine cycle was
used in later developments. SNAP-8 is a joint project between the USAEC and-
the Nationai Aeronautics and Space Administration for developing a reactor of

600 kW(t), 35-50 kW(e), for space applications with a mercury Rankine cycle.
43

SNAP-10 and SNAP-10A are direct-Conversionlsystems. In SNAP-10, the heat is

. converted by conduction from the core to a thermoelectric system. In SNAP-10A,
the heat from a liquid metal coolant is converted by means of a thermoelectric
system. : , _

The SER, SNAP-2, SNAP-8, SNAP-10, afid SNAP-10A differ. chiefly in power, -

dimensions, heat conversion, and some other details, but they are basically of
the general type that has been discussed. ’

The basic core configq;ation for the SNAP reactors consists of the fuel-
moderator rods, of uranium and zirconium hydride in homogeneous mixture, ar-
ranged in a triangular pattern to make up a hexagonal core. Rods are clad with
steel or Hastelloy: stainless steel for the SER and Hastelloy-N for the later
‘reactors. The beryllium reflector is in segments and contains control drums,
which can be rotated to change neutron leakage. For scram, three beryllium
safety plates fall away. The NaK coolant flows axially among the fuel elements.
In SNAP;10A the sfiaces Between the'angles of the hexagonal core and the cylin-
drical core vessel are filled with beryllium metal.

The SE/R,SS’56 50 kW(t), operated from September 1959 until the.end of the
test program, in November 1960. During more than a third of the operating time
the outlet temperature of the coolant was 1200°F. | |

SNAP-2 Development Reactor (S2DR, SDR)55-57 was a second-generation design,

which was to approach the requirements of a space reactor. It operated suc-

cessfully from April 1961 until December 1962.

A prototype of SNAP-_8,55-57 with a higher power, 600 kW(t),.was undergoing
test operation in 1965. o
The design for SNAP-1058’59 incorporates cooling by direct conduction of

heat from the core to thermoelectric devices on the reactor face. Thus a
liquid-coolant s&stem, with pumps, etc., is eliminated. The core consists of
stacks of uranium-zirconium hydride plates, with beryllium plates alternating
to increase heat transfer. This reactor was to be controlled by a strong
negativé temperature coefficient; for startup or shutdown; halves of the
reactor are joined or separatedl This concept originated from a requirement
of the U.S. Air Force for a device to produce 200-250 Watts (electrical). 1In
1958-59 Atomics International developed this concept, with work continuing until
1960. Some réasofis for redirecting the project were: the temperatfire rafige
desired (1100°F) was too high for the thermoelectric material then used (lead
telluride); and there was little prospect of the solid-fin-type heat-rejection

radiator  being developed to increase the power much beyond 300 Watts. With
44

a liquid-metal coolant, as used in the SNAP-2 reactor, the higher p5wgf8=“?;&s
are possible, - | - '

'SNAP-10A56’57{59
kW(t), and lower outlet temperature for the coolant, 1010°F instead of 1200°F.

is essentially the SNAP-2 concépt with lower power, 34

The bulk of the refléctor is outside the core, and it can be éompletély‘re-

‘moved without affecting the coolant system. The direct-conversion system’

utilizes silicon-germanium alloy as the thermoelectric material. The conver-
sion system has an overall efficiency of 1.63%. SNAP-10A is the first nuclear
reactor to be orbited in space. Work on this reactor was begun in 1961 and
performance testing was begun in December 1962. A SNAP-10A reactor (FS-3)
operated in a ground test for 10,000 hours.from January 1965 to March 1966. On
April 3, 1965, another SNAP-10A (FS-4) was orbited in a.SPacéCraft. It operated
untii May 16, 1965, when the system shut down. The cause given as most pro;
bable for the shutdown was a failure in the electrical system of the satellite
carrying the SNAP-10A, rather than in the reactor itself.:

SNAP-4 differs from the other SNAP reactors in purpose and in the coolant

~used. A project on this reactor began in 1959, with the aim of developing a

reactor and turboelectric system that would be very compact and capable of long
unat tended éperation under water- or in a remote land locatibn.6 " The power
level is 1-4 MW(e). The design is very similar to that of the SNAP-2 ‘and SNAP-8.
The fuel-moderator is hydrided zirconium-uranium alloy as rods in bundles that

make up a vertical hexagonal core,62 Unlike the other SNAP reactors, SNAP-4

‘utilizes boiling water as coolant, Water boils in the core and the steam goes

to a turbine. Advantages claimed for the design are compaétness;,feasibility
(built with existing technology); low cost; safety; no maintenance during core
life; versatility in application; and automatic, unattended operation. |

A prominent application of solid homogeneous fuels: in space propulsion is
in some ‘of the Kiwi' reactors and reactors developed from them.

‘Under Project Rover, which began at the Los Alamos Scientific Laboratory
in 1955,63 development of solid-fueled reactors as potentiél sources of flight
power was begun. An aim was to supply basic reactor design for a flight engine.

The réactors_wduld produce high temperatures and would pose no unexpected pro-
64

"blems in startup, stable operation, or shutdown.

64-66 This reactor was

The first integral reactor test was of Kiwi-A:
successfully tested on July 1, 1959. A power of 70MW(t) and an exit-gas temp-
erature of 3200°R were achieved. - The fuel élement is  made up of graphite plates

containing enriched uranium oxide.. These plates, stacked within cylindrical

45

boxes, are in an annular zone around a central island contalnlng heavy water

and control rods. A graphite flow- separator cyllnder d graphite outer re-_'
flector, and a double-walled, water-cooled aluminum presShre'vessel surround

the fuel zoné. The gaseous hydrogen coolant enters the reactor at the upper%
end of the reflector, passes down into the core -up between the fuel plates,l
into an exhaust, and out the nozzle. o ' R

After the testing of Kiwi-A, several'designé were dereloped and tested,
with the aim of proaucing a 1000-MW(t) reactor. Liqfiid, instead of geseous,
hydrogen was the coolant used in some later develepments, and some changestifi
core design were .introduced. Even the early Kiwi tests showed'thenpotefitiei“
of such reactors and stimulated interest in developing a»flight engine.63

In mid-1961 the NERVA (Nuclear Energy for Rocket Vehicle Application)
program was begun under the direction of the NASA-AEC Space Nuclear‘Prdpulsidn
Office. The Aerojet'Generel'Corporafieh was designated as’ the prime con-
tractor, with Westinghouse Electrie Corporatlon as the pr1nc1pa1 subcon-
tractor for development of the reactor,’ shleldlng, and reactor controls. 63

" In the development of the Kiwi reactors, Kiwi-Bl was designed'fdr liquid
hydrogen, but the first Kiwi-B reactor that was operated (Kiwi-B-1A) utilized
gaseous hydrogen.65 It was tested on December 7, 1961, attaining a power of
330 MW(t). The test of Kiwi-B-1B (1962) was the first test in which the reactor
was started and brought to high power with liquid hydrogen as coolant. Testing
continued, with Kiwi-B-4E reaching 1000 MW(t). With Kiwi-B-4E, improved fuel
elements were utilized. The uranium is in the form of extremely small beads of
uranium carbide coated with pyrolytic carbon. The power for the test of ‘this
reactor (August and September 1964) was nearly 900 MW(t), and the temperature
of the coolant leaving the fuel elements was about 4000°R,65 Among the develop-
ments in the Kiwi-B series was the demonstration that the reactors could be
controlled by drums in the reflector,68

The next step in the development was a reaetor designed and built by
Westingheuse Electric and tested by Aerojet-General Nucleonics. This design
was based on Kiwi-B-4A.

Two of the chief problems in designing a reactor for flight were in develop-
ing a structure and fuel elements that would withstand such conditions as:
extremely high temperatures, vibration, and corrosion. The reactor, NRX (NERVA
Reactor Experiment), was given a non-nuclear test (NRX A-l) in the spring of
1964 to test stability and structure. NRX-A2, a nuclear test, was successfully

carried out in September and October, 1964063
46

In the NRX-A reactor, the fuel elementlis enriched uranium carbide, coated
with pyrolytic carbon, uniformly dispersed in graphite. The fuel-moderator
elements contain passages for flow of the coolant (hydrogen), and they are
grouped in clusters. There is an inner reflector of graphite and an outer one
of beryllium. Control is partly by 12 cylindrical beryllium drums, within the
outer reflector, that rotate to turn a poison strip to or afiay from the core.
There are also control rods in the core and the outer reflector.69

Concurrent with the NERVA developments has been the Phoebus program at
LASL.65’68 The Pfioebus reactors are advanced versions, with many improvements,
of Kiwi reactors. They are large, high-power graphite reactors intended to
provide long operating times, high'specific impulse, and restart capability.
Clustering of engines for‘pr0pulsion-is part of this program. Phoebus-1

70,71 Performance

operated in July, 1965 at its design power, 1000 MW(t).
parameters of the Phoebus-2 design illustrate some of the conditions sought:

a power of 5000 MW(t), a hydrogen flow of 285 pounds per second, and an exit-
gas temperature (from the fuel elements) of 4500°R. Above 5000°R, . graphite-

based reactors, with solid-to-gas heat exchange, will probably not be feasible.

65
DATA SHEETS

POWER AND BREEDER REACTORS

47
48

| THIS PAGE
WAS INTENTIONALLY
LEFT BLANK

49

No. 1 First Daniéls:Experimgntai.Power Pile-

Clinton Laboratories

Reference: MonN-188.

Originators: Farrington Daniels suggested pile; development.completed by Power
Pile;Group (later Power Pi}e Division) at‘Clinton Labopatqries,
Status: Preliminary design, 1946.

Details: Thermal neutrons, steady state, breeder. Fuél-moderator: ;U0

23360
containing 2% of 50% enriched uranium. Coolant: helium. . Fértile;matepial:_ Th
metal. Reflector: BeO and_graphite. Core structure:. elements;of'UQé-BeO>in

form of hollow cylinders, 1% in. OD, 1l in. ID, 4% in. long. Cylifi@ers within
2-in. éxiallholeslin hexagonal,@qo prisms, 3 in. across flats, 45 ifi. long. . BeO
prisms form cylindrical core, 6 ft-diam., 5% ft high. 517 channels - in core::
504 for fuel elements, 13 for.safety and cqntrol:rods;‘ Helium:flqwsuinto bottom

of reactor (at:5009F), through. annular spaces between QO -Be0 cylinders and.BeO

prisms,vand out top (at 1400°F) qo go to steam boiler, éore surrounded by .inner
réflector, breeder blanket, outer reflector, stainleés—steel.liner, insulation,
and carbon-steel pressure shell. Control: 6 safety rods and.7 control rods; 6
control rods in circle 15 in. from center of .core .and one at center. .. Power
(max.):l 40 MW(f); 10.MW(e). L

Code: 0312 15 . 31716 43 5932 726  8111X- 941 105
50

No. 2 See0nd Daniels Experimeéntal Power Pile
Clinton Laboratories

Reference: M-4157. 3 o | L o )

'OriginatorS° Farrlngton Danlels suggested original p11e, E.P. Wigner proposed

- Be metal plle, development by Power P11e Group (later Power Pile Division) at
Clinton Laboratorles. '
Status: De51gn study; 1947 _ A
' Details : Thermal neutrons steady state converter; Fuel -moderator: U235 in
U Be alloy. Coolant hellum Reflector°‘_Be. -Fertile material Th. Fuel.
elements' short hexagonal prlsms made of plates of U-Be alloy, 3/16 in. thick.
Elements 8 5 in, long, 7.9 in. in dlameter ~ Core: hexagonal horizontal
‘cylinder w1th Be skeleton 1ncorporat1ng 37 fuel channels parallel to the axis,
4.60 ft dlameter 4.25 ft long. Helium at 10 ‘atm. flows parallel to fuel
elements; 1nlet‘temperature 500°F, outlet temperature 1400° - Containment:
austen1t1c steel inner support. for Be structure; low alloy-steel outer support,
which also acts ~as pressure shell Fertlle material contained in channels in
side reflector and possibly also in end reflector 1ncorporated into ends of
fuel” elements Maxrmum conversion ratio: 0.9. IControl. 8 vertical rods.
Power: 40 Mw(t)t | B '

Code: 0311 . 15 ' 31716 - 44 - 5922 726  8111x 941 - 105

3l

No. 3. Third Daniels Experimental.Power Pile . (-

Clinton Laboratories

Reference: MonN-383.

Originators: Farrington Daniels sfiggested original pile; development by

members of Power Pile Group (later Power PileVDivisidn):at'Clinton Laboratories.
Status: Désign; abandoned, 1947. |

Details: Thermal neutroué, steady State, breeder. Fuel-moderator: 30%-enriched
uranium or uranium COmpound_disperged in-matrix of BeO, Be, or graphite. Coolant:

helium. Fertile material: ThO Reflector: BeO. .Fuel-moderator elements fit

(with some clearance) within 2-§n° axial holes in hexagonal BeO prisms,'3 in. -
. across flats and 6 in. long.. Thes€ -.prisms may, instead of the fuel-moderator
elements, contain BeO alone or fertile material, depending upon whether they are
to be. fuel elements, reflector elements, or fertile elements. Prisms form 517
vertical channels, -2 -in. diam., 76 in. long. Core is cylinder 72 in. diameter
and .76 in. long. Cylinder retained by shell and supported at bottom by plate,
Cylindrical steel pressure shell (8 ft 0D, 1% in. thick, 22 ft 6 in. long) sur-
founds this assembly. Helium flows; at '10 atm., into pressure shell at top at
500°F, flows downward through annulus inside shell, passes upward -through
channels in core, to a plenum at the top inside the pressure shell, leaving the |
reactor at 1400°F.. It is then passed to boilers. Control: 1 control rod at
- core center; 6 shim rods in ting arognd center; 6 safety rods in outermost ring.
Power: 12-20 MW(t); 2.4-4 MW(e). Designs for 60 and 250 MW(t) possible, with
different dimensions, critical 1oadifigs, etc.; designs essentially similar to
this concept. Horizontal reactor stfucturé,'similar in-mostrrespects to this
concept, proposed to facilitate loading and unloading fuel elements.,
Code: 0312 12 ‘31716 43 5912 726 8111X 941 105 -

15 - 59%2 ' -

52

No. 4 Impregnated Graphite, Nitrogen-cooled Reactor

L]

Reference: ANL-4448 (AECD-4095).

Originator: Farrington Daniels.

Status: Design, 1950.
Details: Thermal neutrons, steady state, -breeder. TFuel-moderator: UQ, dis-

2
persed in boron-free graphite; graphite impregnated with UO,(NO and heated

)
235 .. 233 2372
to form uranium oxide; fully enriched in U if U is to be bred, slightly

enriched if Pu is to be bred. 10-15 kg U235 required, Coolant: N, or helium,

.rods, pellets in cartridges, or UO

Reflector: 20 tons of graphite 1% ft thick, at sides, top, and botfiom, Fertile
material: for U233, Th(NOB)4 in moderator or reflector; for Pu, U238 in fuel.
Core: 370 impregnated graphite moderator blocks 4 in. x 4 in. x 45 in. Coolant,
at 5 atm., flows into top of core, down through vertical channels in moderator
blocks, and out bottom to turbine for power production. Coolant inlet tempera-
ture, 558°F; outlet temperature, 1100°F. High temperature allows both power
production and breeding. Arrangement: impregnated graphite blocks rest on |
vertical slabs of BeO. Reactor is 6 ft diam., 5% ft high. Containment:
stainless-steel shell 9 ft diam. and 12 ft high. Control: boron-steel control
rods in holes lined with BeO. Power: 11.5 MW(t); 1.9 MW(e) (designed for use
with gas turbine available at time of proposal).

Code: 0312 12 31718 44 5931 736 81111 941 105 .
31716 42 723

No. 5 Gas-Cycle Reactor

Reference: Nucleonics, lfi,_No. 3, pp. 34-41, March 1956.

Originator: Tarrington Daniels.

Status: Conceptual design, March 1956.

Details: Thermal neutrons, steady state, burner. Fuel-moderator: 107 enriched
U contained in graphite as uranium carbide. Fuel-moderator could be in form of
2 and graphite powders could be mixed in
cylinders and heated, producing uranium carbide. Coolant: helium. Pressure
shell contains coolant at 225 psi. Coolant enters at 743°F and leaves at 1350°F.
Control: reguiating rods containing B or BAC in Mo tubes. Power: 20 MW(t).

Code: 0313 12 31716 43 5941 711 81111 9XX 105
impregnated with U

53

No. 6 High-temperature Bismuth-cooled Power Breeder

North American Aviation, Inc.

Reference: Sidney Siegel, unpublished report, Aug. 1951.
Originator: . Sidney .Siegel. ! . '

Status: "~ Proposed design,'1951.

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

23302. Coolant: helium or liqu;d Bi'f Fertile material: Th

or compound in core or breeding blanket.  Reflector: g;aphite.‘ Fuel-moderator
either solid graphite core' (with axial codlant‘chahnels)‘impregnated homogeneously
with fuel, or impregnafiéd'graphite tubeé_sfiSpendéd in coolant channels; Th-

fertile material: could be impregnated in.graphite, in a slurry of thorium

deuteroxide in D,0 or as thorium fluoride pellets in graphite matrix. - Core:

2
cylinder, 3.4-6.8 ft diam., 3.4-6.8 ft high. - Reflector around core. Coolant

'flows through axial coolant channels. Power: . 1400 MW(t); 470-700 MW(e).

Code: 0312 12 -31105 45 5942  7%6  8XXXX 941 - 105
31716 |

"No. 7 Solid Homogeneous Reactor for Open-Cycle Gas Turbiné

Studebaker—Packétd Corporation

Reference: AECU-3559. .

Originators: A.S. Thompson, staff members.

Status: Preliminary design, 1956.
Details: Thermal neutrons, steady state, converter. Fuel-moderator: compacted
ceramic of UC, and graphite; 10% énrichment if Be is used for additional mod- -

2 .
erator; 20% if graphite. Additional moderator: Be or graphite. Coolant:

"helium. Inlet temp., 1240°F; outlet, 1740°F. Pressure: 20 atm. Reflector:’

grabhite or Be. Corelstrucfure: cénfral_island (40 in. OD) of graphite or
Be and reflector (100 in. OD) of same material,'in_roughly spherical reactor;
fuel' in annulus' between island and reflector. Coolant flow through passages
in fuel-moderator. ‘Containment: horizontal Inconel pfessure shell, 2 in.
thick, 10 ft diam. Power: 60 MW(t).. ‘
Code: 0311 11 31716 = 43 5942 724  8XXXX 921 105

15 .
54

No. 8 Solid Homogeneous Type Gas Cooled Nuclear. Power.Reactor

-

UKAEA

Reference: United Kingdom Patent 850,014,

Originators: Peter Fortescue and G.E. Lockett.

Status: Patent granted, 1960; filed in 1956.

Details: Thermal neutrons, steady state, burner. Designed to operate at high

temperature for power production. Fuel: not specified; fuel elements solid

homogeneous, with moderator and fuel combined. Coolant: gas. Reflector:

graphite. Each of 61 fuel elements contained on a spike projecting from per-

forated platform structure resting on brackets within cylindrical stainless-

steel pressure shell. Fuel elements surrounded by ring of wedge-section graphite

reflector blocks, which form barrier for the gas coolant. Another wedge of

blocks sufrounds this. Core and reflector assembly: elongated core elements

pivotally supfiorted at lower end and drawn together at the upper end by radial

+ pressure differential. of coolanf.. Coolant passes downward betweeq reflector

and pressure shell and upward between core elements. Coolant path.determined B
by gas seals between core reflector, which are also maintained by pressure
differential, o

Code: 0313 1X 317XX 4X 5XXX 711 8XXXX 921 105 | .

No. 9 Semi-homogeneous High-temperature Gas-cooled Breeder Reactor

Japan Atomic Energy Research Institute

il

Reference: AEC-tr-3620.

. Originators: K. Inoue et al.

Status: Preliminary design, 1958.

Details: Thermal neutrons, steady state, breeder. Fuel: Initially 20% enriched

U235 in U02; after breeding begun, U233 used instead of U235. Moderator:

graphite, coated with silicon carbide to make it impervious. Coolant: COZ°

Fertile material: ThO,, Th metal. Fuel, moderator, and ThO, incorporated into

2° 2
elements that can be either rod- or slab-shaped and arranged either vertically

or horizontally. Th metal in blanket of graphite around .core. Coolant inlet -,

temperature: 300°C; outlet, 700°C. Power: 600 MW(t). _ N .

Code: 0312 12 31717 43 5932 726 8XXXX 941 105

o
‘diam. and 1 cm thiqk; with graphite-to-UO

55

No. 10 Semi-homogeneous Critical Experiment (SHE)

Japan Atomic Energy Research Institute

~

References: Trans. Am. Nucl. Soc., 4, No. 1, pp. 56-7, June 1961; JAERI-1032; .
JAERI-4014,

Originator: K. Inoue et al,

Status: First loading completed January 1961; in operation, 1964.
Details: Thermal neutrons, steady state, burner or breeder.. Critical assembly,

235

Fuel-moderator: U02(20% U "7) - graphite. Coolant: air, at ambient (room)

temperature. Reflector: graphite. Fertile material, if any: ThOz. Fuel- .

moderator homogeneous mixture of U0, and graphite in cold-pressed discs 4.5 cm

2
weight ratio of 10 to 1. Graphite

tube contains, in sequence for its 120-cm21ength: fuel disc, l-cm graphite
disc, and 0.5-cm graphife'disc._ Rods arranged in hexagonal,. horizontal array
in two halves of assembly, which come together to achieve criticality. To vary
core composition, discs of Th02, graphiFe-mixtures, and pure graphite can be
added. Control: rods containing B or Cd; 1 control and 3 safety rods in. each
half; large negative témperature coefficient. _
Code: 0313 = 12 31714 43 5931 711 81211 921 105

0312 726 . 81212

84677

No. 11- Semi-homégéneous Gas Cooled Breeder Reactor (SHR)

Japan Atomic Energy Research Institute

References: TAEA Symp. on Power Reactor Experiments, I, pp. 149-54.

Originator: K. Inoue et al.

Status: Design, 1961.
235

‘Details: Thermal neutrons, steady state, breeder. Fuel: initially U , ulti-

mately U233. Coolant: helium. Moderator: graphite. Reflector: graphite,.

Fertile material: thorium. Fuel-moderator of UO2 or UC2 particles, smaller than
6 u, uniformly dispersed in graphite as ring-shaped pellets. Pellets contained
in impervious graphite sheath. Thorium blanket surrounds core. Graphite re-
flector between core and blanket. Power: 25 MW(t). |

Code: 0312 12 31716 43 5931 726  BKXXX 941 105

45 5941

56

No. 12 Peach Bottom High-temperature Gas-cooled Reactor (HTGR)

Philadelphia Electric Co.

References: Proc. Third U.N. Conf., 5, pp. 101-115; J. Franklin Inst.,
Monograph No. 7, pp. 27-40, 1960; Nucl. Sci. Eng., 20, No. 2,

pp. 201-218, Oct. 1964; At. Energy Rev., 1, No. 3, pp. 119-139, 1963; TID-

7662, pp. 299-316; Nuclear Congress, N.Y., 1964, Preprint No. 91; Nuclear News,

8, No.'3, p. 9, March 1965; Atomics, Nov.-Dec., 1964, pp. 10-11.

- Qriginators: General Atomic staff members.

Status:” Critical, March 1966, power operation scheduled for late 1966,
Details: Thermal neutrons, steady state, converter. Fuel: initially 93.5%
enriched U235; Uzjj after breeding cycle has operated. Coolant: helium.

‘Moderator: graphite., Reflector: graphite. Fertile material: ThC Solid

homogeneous fuel element, of which graphite is moderator, cladding, §Uel matrix,
and core structure, consists of'upper reflector, fuel-bearing middle section,
Bottom reflector section, and internal fission-product trap. U235 and thorium
dicarbides (coated with pyrolytic carbon) dispersed in graphite in annular fuei
compacts 1.5 in. long, 2.75 in. in diameter, which are stacked on cylindrical
spine in fuel element. Overall length of fuel assembly: 7.5 ft. Core struc-
ture: 804 cylindrical elements-aligned vertically in triangular pitch. Coolant,
at 23.8 atm., enters reéctor at 660°F, cools reflector, flows upward in tri-
cuspid spaces between fuel elements, exits at 1380°F. Control: 36 control rods
of B4C dispersed in graphite matrix; ZrB2
erature coefficient, dispersed in graphite matrices for burnable poisons.
Power: 115 MW(t); 40 Mw(e).

Code: 0311 12 31716 44 5941 726 81111 921 105

45 84677

and Rh, which enhances negative temp-

57

*
No.. 13 TARGET 1000-MW(e) Power Reactor (Possible Modification)

General Atomic

Reference3° GA-4706; At° Energy Rev., 1, No. 3, pp. 119-139, Sept. 1963.

.Orlglnators Staff members

Status: Prellmlnary de51gn 1964, pbésible alternative fdel to allow release
of fission products. ' ‘ . |
. Details: Thermal neutrons, steady state, converter. Fuel-moderator-fertile
material: pyrp;arbonjcoated pérticles (L p or less in diameter) of carbidgé.bf

U235 and Th diépersed in graphite. Adaitionél moderator: BeO. Coolant: helium,
| Refléctor: graphite. Fuel element: main features include graphite fuel matrix
around BeO sfiine in cylindrical graphite element with an upper reflector and an
internal fission-productltrap., Element is 4.5 in. diameter, 20 ft.long. Helium
flows upward thfough channéls formed by the approx 5500 fuel elements and approx
240 control rods that are oriented vertically in triangular arrangement to form
cylindrical cbre. Active core height, 15.5 ft; diameter,'31.l ft.  Coolant inlet
temp., 720°F; outlet, 1470°F. Reactor Qessel: prestressed concrete lined with
steel, 56 ft, 76 in. ID, 76 ft, 8 in. outside height. Control: negative tempera-
ture coeff1c1ent' 242 vertical cylindrical rods, 3 in. by 20 ft; rod is stainless-
steel sheath holdlng boron carblde in graphlte (30 w/o boron) contalners Power:
2270 MW(t), 1000 Mw(e). | ,
Code: 0311 12 31716 44 5941 726 81111 921 . 105

| | | 84677

Thermal Advanced Reactor, Gas-cooled, Exploiting Thorium

58

* : ‘ ’ .1‘. -
No. 14 OECD High Temperature Gas Cooled Reactor Project (DRAGON)

European Nuclear Energy Agency

References: Proc. Second U.N. Int. Conf. 9, pp. 289-305; Proc. Third U.N. Int.
Conf., 1, pp. 318-325; Nucl. Power, 5, No. 46; pp. 112-117, Feb. 1960.
QOriginators: L.R., Shepherd et al. .

Status: Critical Aug. 23, 1964.

Details: Thermal neutrons, steady state, breeder. Fuel: 14 kg U235. Coolant:

" helium. Moderator;"graphite. Fertile‘matérialz Th. Reflector: -graphite.
Core divided into two zbhes: cenfral-zone fuel of UC2-ThC2 particles coated with

pyrolytic carbon and silicon carbide; surrounding-zone fuel UC,-ZrC pérticles

coated with pyrolytic carbon. Fuel particles mixed with resin%bonded graphite
pofider and pressed into compacts, which are inserted into graphite fuel tubes.
Seven tubes form a fuel element. Core composed of 37 fuel elements in tri-
angular 1a§tice forming appréximate cylinder within annular graphite reflector.
Surrounded by steel pressure vessel. Around pressure vessel are six branches;
each contains at the top heat exchanger and blower. ‘Inlet and outlet ducts |
arranged concentrically, andiflow‘is baffled, so that pressure vessel is ex-
posed bnly to coolant gas at inlet temperature. Coolant passes upward through
core, then to inner concentric duct of one of the heat-exchanger branches, then
goes to blower and to annular space between neutron shield and pressure vessel,
In reactor experiment, steam produced is bled off, not used for power in.tur-
bines. 1Inlet temperature of coolant in reactor: 350°C.(662°F); outlet tempera-
ture: 750°C (1382°F). Shielding: steel pressure vessel; water; concrete
building around steel vessel. Control: 24 Bac rods arranged in circle immedi-
ately surrounding core, 1 for fine control. Power: 20 MW(t). Power density:
14 MW/m3. :

Code: 0312 12 31716. 44 5941 726 81111 921 105

* .
Organization for Economic Cooperation and Development
59

No. 15 Experimental Beryllium Oxide Reactor (EBOR)

. General Atomic

References: Proc., Third U.N. Int. Conf., 6, pp. 323-332; Nucleonics, 19, No. 3,
p. 31, March 1961; GA-2603.. -~ - .« '

Orlglnators Staff members.

Status: Expected to begin Operatlon in 1966 land-based prototype for Maritime
Gas-cooled reactor. . _
Details: Thermal neutrons, steady state, burner. . Fuel-moderator: 70% enriched
U as UO2 combined with BeO'in compacts. -Coolant: helium. Reflector: BeQO. Core
structure: fuel-moderator pellets, 0.33 in. diam., 0.43 in. long, contained in
Hastelloy-X fuel pins, 178 péllets to a pin. Fuel elements contain spine of BeO,
~instrumentation tube, fuel pins, inlet and outlet ports for gas coolant, and
shroud.. 52 reflector elements are Hastelloy-X-clad_annular BeQ rings. 36 core
eiements arranged cylindrically. Core, 23.43 in,,squaré,.76 in. long, is sur-
rounded by 7 in. BeO. Coblant enters and leaves through two concentric passages
at side of reactor. Gas_fiow divides on entering reactor. One"stream‘fiows up-
wa;g between pressure vessel and thermal s@ield,atheh downward between core
container and thermal shield. Other stream flows downward between pressure
vessel and.thermal shield. Streams join below core and flow upward through

core and intoiaq'upper_plenum before leaving_reactor, . Coolant inlet temp.,
750°F; qutlet temp., 1300°F; pressure 1120 psia. Cylindrical pressure vessel

of low Cr-Mo alloy steel. Control: & cruciform shim-safety and regulating rods
of:dySProsium oxide in glumina,tiles contained_in Hastelloy-X structure. Power
density (avg.): 14.4 kW/liter. Power: 10 MW(t). ‘

Code: 0313 15 . 31716 43 5931 711 81134 921 105

No. 16 Maritime Gas-cooled Reactor (MGCR)

General Atomic

References: Nucleonics, 19, No. 3, p. 31, March 1961; GA-2603.

Orlglnators Staff members, General Dynamlcs Corp.

.Status:  Land-based prototype (EBOR) being constructed June 1965.
Details: To be based on experience with EBOR. Power: 53-77 MW(t).
Code: 0313 15 31716 43 .. 5931 711 81134 921 105 .

60

No. 17 Ultra High Temperature Reactor Experiment (UHTREX)
(Formerly TURRET)

LASL

References: LA-2198; LAMS-2469; LAMS-2623; J. Franklin Inst., Monograph No. 7,
pp. 127-132, 1960; Trans. Am. Nucl. Soc., 2, No. 2, pp. 148-149, 1959; Atomics,

Nov.-Dec., 1964, p. 13.

Originators: R.P. Hammond et al.

Status: Completion scheduled for 1966.

Details: Thermal neutrons, steady state, burner. Fuel-moderator: 93% enriched
U235 as carbide in graphite. Coolant: helium. Reflector: graphite; ungraphit-
iéed carbon.. Hollow cylinders of commercial graphite, 1 in. OD, % in. ID, 5%‘in.

long, are impregnated with U02(N0 then dried and heated to convert nitrate to

3)ys
oxide, predominently UOZ' Subsequent heating in core completes conversion to
carbide. Carbide is present as microspheres coated with pyrolytic carbon.
Coating delays release of fission products. Core: vertical graphite cylinder

70 in. diam. Fuel-moderétor elements positioned in horizontal, radial holes,
which pierce core at 15° intervals in 13 equally spaced horizontal planes, making
a total of 312 channels. Core can be reloaded while operating at full power.

Each channel contains 4 elements end to end. - New element introduced through gas

lock and rammed into core. Inner-most element falls down a slot into central

‘core plug and is removed through gas lock and conveyor. Core mechanically

rotated and indexed at 15° intervals so each vertical row of fuel elements is

aligned in turn with the single row of stationary loading rams. Original reactor

name, TURRET, came from this feature. Coolant enters core at bottom, at 1600°F,

through 12.5 in. diam. annulus; flow divides into multiple radial fuel channels,
and is collected in plenum at core periphery for discharge from reactor at 2400°F.
Core surrounded by reflector ofliin. graphite and 12 in. dense, ungraphitized
carbon. Cofitainmenf: 14 ft diam. carbon-steel spherical pressure vessel. Temp-
erature held to about 600°F by internal refractory insulation of porous carbon
brick. Conducted heat radiated to water-cooled shielding walls. Control: large
negative temperature coefficient of reactivity allows fuel loading and coolant
flow to control temperature and power during reactor operation; Mo tubes filled
with ZrB2 inserted into stationary part of core to compensate for about 20%. % k/k.

Power: 3 MW(t).

. Code: 0313 12 31716 44 5942 711 84677 921 105

81211
material:. pebbles in which PuO, and ThO, are dispersed in BeO. - Coolant: CO

61

No. 18 Australiqq;fligh'Tgmperature Gas-cooled Rea@tor_(HTGG)

Australian Atomic Energy Commission

. References: J. Franklin Inst., Monograph Nd.-?, pp. 81-86, 1960; Nucl. Eng., 9,

No. 92, p. 9, Jan. 1964; Proc. Third U.N. Int. Conf., 11, pp. 329-340; J..Nucl.
Materials, 14, pp. 29-40, 1964,

Originators: Staff members.

Status: Design studies in progress.
Details: .Thermal neutrons, steady state, breeder. Fuel-moderator-fertile

2 2 2°

Reflector: side, graphite; end, BeQO.. Core: cylindrical vessel containing -

pebbles, which are élowly added at top and removed at bottom. Containment:. .

spherical pressure vessel. Coolant flows up between core and pressure vessel

‘to top of core, then downward through core; leaves at bottom to go to heat

exchéfiger. Coolant enters core at about 350°C (662°F); leaves at 700-800°C

" (1292-1472°F). Fuel surface temperature: 1100°C (2012°F) max. Coolant gas

pressure: > 500 psi. Control: 2 vertical rods. Power density: 10-20 ..
kW/liter. . % ; | .
Code: 0312 15 31717 46 . 5932 726 8l11X 921 105

No. 19 Solid Homogeneous Reactor for MHD,(Magnetohydrodyhamicfii

Power Generation

Westinghouse Electric Corporation

Reference: Nucl. Energy, March,'l963,“pp. 60-66..

Originators: E.J. Sternglass, T.C. Tsu, G.L. Criffith, and J.H. Wright.
Status: Preliminary design, March 1963. | '

Details: Thermal Aeutrohs, sfieady state, converter; Fuel-moderator: slightly
enriched U as U02 particles distributed uniformly in matrix of graphite.
Coolant: helium. Reflector: graphite. Core: cylinder of graphite moderator
and reflector, with Jértical channels for fuel-moderator elements. Coolant
enters bottom of core at 1690°F, passes upward through it, and leaves at 2500°F

to go‘to a MHD genérator. Fueling machine permits refueling of reactor while .

on load. Power: 600 MW(t); 264 MW(e).

Code: 0311 © 12 31716 42 5942 723  8XXXX 921  105°
62

No. 20 Boiling-water Reactor With Solid Homogeneous Core - ' .

Brown, Boveri & Cie:., Aktiengesellschaft

Reference: United Kingdom Patent 854,291.

Originators: Staff members.

Status: Patent granted Nov. 16, 1960; applied for Aug. 4; 1956 .in Germany,
Details: Thermal neutrons, steady state;,breéder. Fuel-moderator-fertile -

material: rods containing all three; typical mixture would be U2330 BeO,

2’

and ThO,. Coolant: boiling H 0,_bn1y in quantity to flow around fuel elements

to remoie heat by boiling but iot enough to act as moderator. ‘Reducing amount
of water reduces absorption of thermal neutrons and gives greater breeding
.efficiency. ' '

Code: 0312 15 3210Y - 45 5432 - 726 8XXXX 9XXX 105.

1X . 4. 5XX2 - 72X

No. 21 Semi-homogeneous Bismuth Cooled Bréeder_Reacfor (SHR)

Japan Atomic Energy Research Institute

Reference: IAEA Symp. on Power Reactor Experiments, I, pp. 149-54.

Originator: I. Inoue et al.

Status: Design, 1962, |
Details: Same as Semi-homogeneous Gas Cooled Breeder Réactor described in Data
Sheet No. 11, except that liquid Bi rather than helium is used as coolant.
Code: 0312 12 31105 43 5131 726  8XxxX' 941 105

45 5941
63

No. 22 Terrestial :Unattended Reactor Power System (TURPS)

Martin Nuclear Div., Martin-Marietta Corp.

References: - Nucleonics,.23, No. .8, p. 46, Aug. 1965; Trans. Am. Nucl. Soc., b,
pp. 320-321, Nov. 1963.

Originators: C.E. Murphy, staff members. -

- Status: Under investigation, 1965; USAF contract.

Details: Thermal neutrons steady state, probably burner. Fuel-moderator: U

(probably.enrichefl) mixed with Zer. Originally, enriched U as a fuel and H20
as moderator were suggested as ‘alternatives. Coolant: boiling P433. Fuel-

moderator of*U-Zer and segment of unfueled Zer make up 2-segmernt element.

Hydrogen migrates from unfueled segment to fueled segment to act as moderator.

Control: reduction in heat load on unfueled segment reduces hydrogen migration,

with consequent reduction in moderation and activity; Coolant in natural circu-
lation boiling-condensing .cycle. . Temperature: - 1000°F. Power: 100 kW(e)

minimum.

‘Code: 0313 . 17- 32113 . 44 5981 711 84677  9XX 105

64

No. 23 Solid Homogeneous Reactor for Rocket or Aircraft Propulsion

. (""Cabbage Head')

Referénces:, U.S. Patents 2,894,891 and 2,917,443; personal communication,

Originator: J.J. Grebe.

Status: . Patents granted July 14, 1959, Dec. 15, 1959; first patent filed
Oct. 3, 1949. S : _ "

) ' 233 235
Details: Thermal neutrons, steady state, burner. Fuel-moderator: U y U R
or Pu239 dispersed in LiH, Be(O, BeH

BeZC, or graphite, Coolant: H, or air,

2° 2
No reflector. Fuel-moderator: tapered segments of two layers separated by
coolant passages. At one end of the core, each segment layer converges and
closes. At other end, ێach segment is joined to the adjacent segment. Fuel-
moderator is porous to gas flow. Segments are bowed outwardly to form a'roughl&
spherical or pear-shaped core, the outer segments being larger than the inner
ones. Outlet ducts from thé elements merge into a central coolant channel,
which itself becomes larger and leads into a‘jet.nozzle, The coolant flows
into inlet ducts, passes through the porous fuel-moderator into coolant channels,
and leaves through the central coolant channel and the nozzle. Coolant could
enter as a liquid, vaporize as it is heated in passing through‘core, then dis-
sociate to atomic state in further heating. Outlet temp.: over 6000°F max. for
hydrogen. Porous structure designed to: permit close contact of éoolant gas
with fissionable material;‘and provide many short parallel paths for gas flow
through it, to allow large volume of gas to pass through in unit of time without
excessive pressure loss. The reactor could be divided into separate sections,

according to the temperature gradient in each section, with different materials

used in each section. At the hot (exit) end, a refractory material, such as

graphite, would be used. For the cooler sections:{(the inlet and the central
portions) less refractor& moderators with better nuclear characteristics would-
be used. Li7H or Bezc would be examples. Control: B or Cd control rod'may be
used,
Code: 0313 12 31714 44 5982 711 - 81x11 9XX 105

15 31715 45 ' 81X12

17 32115 46

33115
65

No. 24 ‘Solid Homogeneous Reactor Moderated with. Zirconium Hydride

Bendix Aviation Corporation’

- Reference: Unpublished reports, Bendix Corp., May 15, 1952. .

Originators: Staff members,.

© Status: - Preliminary design, :
Details: 'Thermal nefitrons, steady state, burner. Fuel-moderator: -U235 dis-
tributed homogeneouslywin‘Zer.- Coolant:--HZOL ‘Refleetor° H20 Fuel - moderator
compacts distributed in insulated spherical core, which also contains tubes for
coolant and for superheat. HZO (under 150 psi) reflector, 2 in. thick, around
core insulation; core 12.6 in. diam. Reactor shell 17.3 in. OD. Coolant enters
'a£'280°F, 625 psi, leaves at 600°F, 600 psi. Control: Cd absorbing material in
reflector. Power: 12 -kW. |

Coae: 0313 17 31101 44 5982 711 81X62 921 ° 105

No. 25 XMA-I Reactor

General Electric Co., Aircraft Nuclear Propulsion .Dept.

: Referenee: G.C. Leeth,  unpublished report,'General Electric Co., 1958.

.Orlglnator G.C.. Leeth.

Status: Program cancellea 1961 _

Details: Thermal neutrons, steady state, burner Fuel-moderator: U02.(highly
enrlched) in BeO yttrium hydride. Coolant: air. Core arrangement UOZ-BeO
and yttrlum hydrlde arranged homogeneously in concentric polygons. qulant
leaves at 1800- 2000° Moderator wall temperature: 1900°F; fuel-element temp-
erature:v‘2500°E. Control flat plate inserted radially betWeen yttrium
hydride siaBso |

Code: 0313 110 31714 44 5932 711 8132X  9XX 105
66

No. 26 Pratt & Whitney Aircraft Reactor-II (PWAR-II)

Pratt & Whitney Aircraft, Div. of United Aircraft Corp.

Reference: Unpublished report, Pratt & Whitney Aircraft, 1960.

Originators: J. Levengood, J. Dissler, and J. Kalinowski.

Status: Program cancelled, 1961.

Details: Thermal neutrons, steady state, burner. Fuel-moderator: 20% enriched
U in UO2 contained in BeO matrix. Coolant: Li. Reflector: Ber Fuel-moderator
in form of pins in hexagonal containers. Core; right circular cylinder contained
in Nb-Zr pressure vessel. BeO reflector on sides and ends of core. Control:

B4C contained in 8 rotating drums.. Power: 510 MW(t); similat reactor designed

to produce 200 MW(t).
Code: 0313 15 31106 43 5931 711 81441 921 105

No. 27 Hydrogen-Cooled, Solid Homogeneous Reactor for Rocket Propulsion

Thiokol Chemical Gorp.

Reference: Nucl. Sci. Eng., 7, pp. 377-386, April 1960.

Originators: J.J. Newgard and M.M. Leval.

Status: General design concept, 1960. .

Details: Thermal neutrons, steady state, burner. Fuel-moderator: uranium-
loaded graphite plates. Coolant:‘ hydrogen. Reflector: ' Be0. Core formed by
parallel plates of uranium-loaded graphite, which may.be clad or unclad. Plates
spaced by protrusions and are tied together in bundles, 2 in. by 2 in. by 6 in.,
by W or Mo wire. Reflector: 6 in. thick, BeO blocks across top of reactor and
along length of core. Axial holes: for coolant flow through reflector. Coolant
flows through top end section of reflector, through core (where temperature is
raised to 4200°F), and is exhausted to nozzle. Small amounts of gas pass through
reflector and through structural columns in core. Control: 18 curved control
plates in reflector. contain inside section of borated stainless steel, enclosed
in stainless steel; space between them for coolant gas. Plates supported by
tube. Rotating tube changes degree of neutron absorption. |

Code: 0313 12 31715 44  59%x2 711 81441 921 105
67

No. 28 Tory ITA-1- Reactor -

Lawrence ;Radiation Laboratory, Livermore .

References:  UCRL-6923; Nuclear News, 7, No. 7, pp. 8-11, July 1964. .

Originators: Staff members.

Status: Part of PLUTO program on nuclear ramjet engines; successful test of
reactor, May 1961; program suspended, 1964. ' | _
Details:  Intermediate neutrons, steady state, burner. Fuel-moderator: highly
enriched UO2 homdgeneously mified with Be0O to form hexagonal elements. Coolant:
air; exit-gas temperature, 1975°F. Reflector: grgphite. Core structure:
100,000 elements (4 in. iong, 0.297 in.-across'fl;ts, with 0.200 in. diameter
hole fof paésage of air) arranged ihn hexagofial bundles, about 5 in. across,
within BeQ structural elements to make cylindrical core of 32 in. diameter and
4 ft long. Air-cooled support tube in each corner of bundles. Around core are
shroud, Al pressure shell, and 2-ft-thick graphite reflector. Reflector contains
8 graphite cylinders with boron steel at outer edge of one quadrant of each.
Reflector is.water-cooled.’ Air coolant flows in through inlet diffuser, into
front of core, out core to exhaust. Control: cylinders rotate to control
reactor by increasing or decreasing'effective size of reflector; 4 rods, for
fast control, near inner wall of reflector. Power: 155 MW(t).
Code: 0213 15 31714 44 5932 711 82441 921 105

' ' 81111 A

No. 29 Tory IIC Reactor

Lawrence Radiation Laboratory, Livermore

Reference: .Nuclear News, 7, No. 7, pp. 8-11, Julf 1964.

Originators: Staff members.

Status: Part of PLUTO program on nuclear ramjet engines; successful test of
reactor, May 1964; program.suspended 1964. '
Details: Very similar to -Tory IIA-1 reactor. Several hundred'thousand'fuel-.

moderator elements make up core. Control rods within core. Operating tempera= :: .

“ture: > 2000°F. Power: 600 MW(t). .
. Code: 0213 15 31714 44 5932 711 8211 921 105

68

No. 30 SNAP Experimental Reactor (SER)

Atomics International, ‘a Division- of North American Aviation, Inc.

References: Nucleonics, 19, No. 4, pp. 73-76, April 1961; Proc. Third U.N. Conf.,

on Peaceful Uses of Atomic Energy, 15, pp. 164-175.

Originators: Staff members..

Status: Successful operation, 1959-1960; dismantled.

Details: Thermal neutrons, steady state, burner.  Fuel-moderator: - fully enriched
U235-in homogeneous mixture with Zer (7 w/o U235); 2.9 kg U235° Coolant: Nak;
inlet temp., l000°F; outlet, 1200°F. Reflector: axial, 1.5 in. Be; radial, 3 in.
Be. Core structure: 61 stainless-steel-clad fuel-moderator elements in tri-
angular pattern to make up hexagonal core 9 in. across corners and 8 in. across
flats. Rods 14 in. long and 1 in. in diameter. Core volume: ~Q.4 cu. ft. Core
vessel: 16 in. high, 9.5 in. diameter. Control: 3 Be drums rotate around core;
reflector thickness varied; 3 Be safety plates fall away from. reflector for scram.
Power: 50 kW(t).

Code: 0313 17 31204 44 5981 711 82448 921 105

No. 31 SNAP-2 Development Reactor (S2DR)

Atomics International

References: Proc. Third U.N. Conf. on Peaceful Uses of Atomic Energy, 15, pp. 164-
175; Nucleonics, 23, No. 6, pp. 44-47, June 1965.

Originators: Staff members.

Status: Test operation; pfedecess&r‘to cénceiled.SNAP-Z Reéctor.

Details: Very similar to SER, but with differences in dimensions, etc. 10 w/o U235
in fuel-moderator element; 37 fuel-moderator elefien;s, Clédding: Hastelioy—No*
Core loading: 4.3 kg u235f Core vol.: ~Q24 cu, ft.. Control: 2 Be drums; 2 Be
safety plates. Power: 50 kW(t).

Code: 0313 17 31204 44 5981 711 82448 921 105

69

No. 32 SNAP-2 Reactor

Atomics International

Reference: Nucleonics, 19, No. ‘4, pp. 73-76, ‘April 196l.

Originators: Staff members.

Status: Design intended as predecessor for SNAP-8; discontinued, 1963.
Details: Same as SNAP-8 (see below), but with power of ‘3 kW(e). “Meréury S
Rankiné conversion cycle. ' ‘

Code: 0313 - 17 - "~31204 44 - 5911 711 82448 - 921 105 -

No. 33 SNAP-8 Reactor
Atomics International
References: Proc. Third U.N. Conf. on Peaceful Uses of Atomic Energy, 15,

pp. 164-175; Nucleonics, 19, No. &4, pp. 73-76, April 1961; 23, No. 6, pp. 44-
47, June'1965.

Originators: ;Staff members, _ _

Status: Test operation of nuclear prototype continuing in 1965. ,
Details: Very similar to SER. 10 w/o U235 in fuel-moderator element. 6.56 kg
U235° 211 fuel-moderator elements. Cla&ding: Hastelloy-N. Coolantviniep

" temperature: -1100°F; outlet, 1300°F. No .axial reflector; 3 in. Be for radial,
6 control'drums,-'Power:: 660 kW(t); 35-70 kW(e), With mercury Rankine :cyéle.

Code: 0313 . 17 31204 | 44 5911 711 . 82448 921 . 105
70

No. 34 SNAP-10 Reactor

Atomics International

References: NAA-SR-3473; personal communication, R.;A, Du Val.

Originators: Staff members.

Status: Preliminary design, 1959; terminated 1960. A

Details: Thermal neutrons, steady state, burner. Fuel-moderator: U-Zrfix,

containing 3.2 kg U235, in flat plates, 7 in. diam., 3/4 in. thick, stacked

to form cylindrical core. Be plates, 1/8 in. thick, alternatinglwith,fuelf:

moderator plates in stacks to promote radial heat transfer. Cooling by

direct conduction of heat from core to thermoelectric converter devices on

face of reactor; cold-junction temperature:; 780°F; hot-junction temperature:

1100°F. Reflector: 2% in. Be on sides and ends. Reactor 12 in, diam., 18 in.

high. Control: strdng negative temperature coefficient; for startup or shutf

down, two halves of reactor Brought together or separated. Power: 12 kW(t);

0.3 kW(e). ‘ ‘

Code: 0313 17 2122 44 5981 711 83169 921 109
I 84677

No. 351'SNAP—10A Reactor

Atomics Interhational

References: Proc. Third U.N. Conf. on Peaceful Uses of Atomic Energy, 15,

pp. 164-175; Nucleonics, 23, No. 6, pp. 44-47, June 1965; personal communi-
cation, R.A. Du Val. -

Originators: Staff members.

Status: One reactor orbited in spacecraft April 3, 1965; another operated

in ground test for 16,000 hours from Jan. 1965 to March 1966.

Details: Based on SNAP-2 and SNAP-8.. Thermal neutrons, steady state, burner.
Fuel-modefator: 10 w/o fully enriched U235 in homogeneous mixture with ZrHX.
Coolant: NaK. Reflector: Be. 37 fuel elements (l% in. diam.) in triangular
array form hexagonal core, with Be plates in spaces between core angles and'cylinQ
drical core vessel (9 in..diam.). Coolant flows axially in spaces between fuel
elements., Inlet;tempg, 900°F; outlet, 1010°F. External reflector of Be. Control:
rotating four sections of reflector varies neutron leakage for startup control; -
burnable poisons; negative temperature coefficient for long-term control. Power:
34 kW(t), 0.5 kW(e). Thermoelectrfc'direct powef conversion system,

* Code: 0313 17 31204 44 5981 711 82448 921 105

' ' 84677
.around central island containing D

71

No. 36 Hydride Moderated Boiling Reactor (SNAP-4).
Atomics International
’ )
References: Unpublished report, Feb. 18, 1963; SNAP Fact Sheet, USAEC, June 15,
1962, .

Originators: 'Staff members.

Status: Development incorporated into later work. _
Details: Thermal neutrons, steady state, burner. Fuel-moderator: highly en-
riched U alloyed with hydrided zirconium in form of rods clad with noncorrosive

alloy. Coolant: boiling H,0 at 1200 psi. Bundles of fuel-moderator rods make -

2 :
up vertical hexagonal core. Water enters bottom of reactor, is heated to boiling,.
and passes out of top to go through steam separator to turbine in direct-cycle
operation. Control: rotating drums in reflector around core. Power: 1-4 MW(e).

Code: 0313 17 32101 44 5981 711 82448 921 105

No. 37 KIWI-A Reactor

LASL ~

References: Nucleonics,19, No. 4, pp. 77-79, April 1961; Astronautics and
Aeronautics, 3, No. 6, pp. 42-46, June 1965; LADC-5261.

Originators: Staff members.

Status: Tested, 1959-60; ipcorporated‘in later designs.

Details: Thermal neutrons, steady state, burner. Fuel-moderator: graphite

plates in which U23502 is incorporated. Coolant: H2° Reflector:. graphite.

Core arrangément:- fuel plates, 8 in. long, 5-8 in. wide, and % in. thick, with
longitudinal ribs on one surface. Plates stacked in cylindrical boxes, with
ribs maintaining passage for flow of gés. Graphite disc, with coolant passages
acts as neutroh reflector at core.inlet side. Fuel elements in annular zone
2O and control rods. Surrounding fuel zone
is graphite flow-separator cylinder, graphite outer reflector, and Al double-

walled pressure vessel, which is water-cooled. Pressurized H, goes into reactor .

2
via a plenum at upper end of reflector, passes through holes in reflector down
into core inlet plenum, passes up between fuel plates, enters core-exhaust plenum,

and leaves through the nozzle. Control: vertical scram and shim rods in central

- island. Exit-gas temperature: 3200°R. Power: 70 MW(t).

Code: 0313 12 31715 44 5931 711 8l11X = 923 - 105
72

No. 38 KIWI-B Reactor

LASL
¥

References: Aslronautics and Aeronautics,'g; No. 6, pp. 42-46, June 1965;

LADC-5490; IEE Trans. in Nucl. Sci., NS-12, No. 1, pp. 160-168, Feb. 1965.

Originators: Staff members.

Status: Several designs tested,
Details: Similar to Kiwi-A. Fuel: minute beads of uranium carbide coated

with pyrolytic carbon. Coolant: 1iqfiid H Control: drums in reflector.

2° ,
‘Code: 0313 = 12 31115 44 5931 711 8l441 °~ 923 105

———————

No. 39 NRX-A Reactor /

Westinghouse Astronuclear Laboratory, Westinghouse Electric Co.

Reference: Unpublished report, Nov. 1963.

Originators: Staff members; design developed from LASL Kiwi-B concepts.

Status: Engineering design, 1963; development continuing.

Details: Thermal neutrons, steady state, bufner. Fuel: enriched U. Moderator:
graphite, Coolant: H2. Reflector:‘ inner, graphite; ofiter, Be. Fuel-- ‘
elements: enriched UC2 coated with pyrolytic carbon and uniformly dispersed in
graphite. NbC coating on coolant passages and on parts of external surface. - -
Fuel elements grouped to form clusters. Pressure vessel around outer reflector.
Shield between pressure vessel and propellant tank. Control: 12 cylindrical

Be control drums within outer reflector rotate poison strip (20% B10 in Al)

toward or away from core; shim rods in outer reflector; shim rods in some un-

fueled elements in core--moderator shims of graphite and poison shims of graphite

containing dispersed Ta carbide.
Code: 0313 12 '31715f‘ 44 5942 711 81441 921 105
81111 |

5t
10.

11.

12.

‘Gas-cooled Breeder Reactors, AEC-tr-3620, Oct. 20, 1958.

-73

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