CHAPTER 9

LARGE-SCALE HOMOGENEOUS REACTOR STUDIES*

9-1. INTRODUCTION

9-1.1 The status of large-scale technology. A large number of groups
in the national laboratories and in industry have prepared detailed designs
of full-scale homogeneous reactors because of the widespread interest in
these reactors and the generally accepted conclusion that they have long-
term potential for central-station power production and other applications.
These designs have, in some cases, been made to compare the economics of
power production in homogeneous reactors with other nuclear plants. In
other cases, the designs have served as the bases for actual construction
proposals. Unfortunately, none of the proposals has yet initiated the con-
struction of a reactor, for it is believed that the gap between the existing
technology of small plants and that necessary for a full-scale plant is too
great to bridge at the present time. Thus the construction of full-scale
plants must await further advances in technology which are expected to
be achieved in the development programs now under way. The extensive
studies of full-scale plants do, however, constitute a body of information
vital to the nuclear industry. Tt is hoped that the summaries of the large-
scale homogeneous reactors given in this chapter will serve as a guide to
those contemplating the building of a full-scale nuclear plant.

One of the major problems yet to be solved for a large-scale circulating-
fuel reactor is that of remotely repairing and /or replacing highly radioactive
equipment which fails during operation of the plant.

The various proposed solutions to this problem fall into two categories:

(1) Underwater maintenance, in which all equipment is installed in a
shield which can be filled with water after shutdown of the reactor so that
maintenance operations can be performed from above with special tools
and with visibility provided through the water.

(2) Dry maintenance, in which all operations are done by remote meth-
ods using special remotely operable tools and remote viewing methods
such as periscopes and wired television.

In either case, remote opening and closing of flanged joints or remote
cutting and rewelding of piping must be used to remove and replace equip-
ment. A solution of the problem of maintaining flanged joints in a leaktight
condition in large sizes has not been attempted, the largest pipe in use to

*By C. L. Segager, with contributions by R. . Chapman, W. R. Gall, J. A.
Lane, and R, C. Robertson, Oak Ridge National Laboratory.

466
9-1] INTRODUCTION 467

date being approximately 10 in. in diameter. Remote cutting and re-
welding equipment is still in the early stages of development.

The technology of solutions systems is in a more advanced stage of
development than that of slurry systems because of the design and opera-
tion of two homogeneous reactor experiments and the associated develop-
ment work. Some of the problems remaining to be solved for large-scale
solution reactors include the development of large-scale equipment such as
pumps, valves, feed pumps, and heat exchangers; radiation corrosion of
materials used in the reactor core; high-pressure recombination of hydrogen
and oxygen; and reduction of the number of vital components upon which
reactor operation depends. Instruments for measuring temperature in
high radiation fields and control of inventory and level are some of the
major instrumentation problems for which better solutions are needed.

The achievement of a suceessful aqueous homogeneous thorium breeder
requires a high-pressure thorium-oxide slurry system. Development work
has been under way for several years to determine the characteristics of
such a system and to develop ways of handling slurries. The technology is
not yet advanced to the point where a large-scale breeder reactor of this
tvpe can be built and operated. Slurry problems under study include
methods of production, circulation through pipes and vessels, storage and
resuspension, cvaporation, heat removal, flow distribution, particle size
degradation, internal recombination of deuterium and oxygen, general
information on erosion and corrosion effects, and effects of settling on
maintenance operations.

Extrapolation of small-scale technology to large-scale design presents
<everal problems of uncertain magnitude, especially in the design of equip-
ment for handling slurries of thorium oxide such as are specified for one-
or two-region breeder reactors. The problems of mamtenance of slurry
systems are essentially the same as for solutions, but are complicated by the
erosive nature of the slurry, its relatively high shear strength, and its tend-
ency to cake or settle in regions of low turbulence.

9-1.2 Summary of design studies. The design studies deseribed in this
chapter were made by the national laboratories of the Atomic Lnergy
Commission and by various industrial study groups for the purpose of
determining the technological and economic feasibility of aqueous homo-
geneous reactor systems as applied to central station power, research
reactors, and the production of plutonium. In general, the design criteria
used in the studies conform as closely as possible to known technology n
order to minimize the scope of new developments required to ensure the
suceess of the proposals. In all the studies, the importance of over-all
safety and reliability of the reactor complex and mdividual reactor com-
ponents has been emphasized. Also, considerable attention has been
devoted to the maintenance aspects of the desigus.
468 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHaP. O

The large-scale reactor designs described are grouped according to
the following categories:

(1) One-region solution reactors, typified by the Wolverine Reactor
Study, the Oak Ridge National Laboratory Homogeneous Research
Reactor, and the Aeronutronic* Advanced Engineering Test Reactor.

(2) One-region breeders and converters, such as the Pennsylvania Ad-
ranced Reactor reference design by Westinghouse Iflectric Corporation,
the Ilomogeneous Plutonium Producer Study by Argonne National
Laboratory, and the Dual Purpose Feasibility Study by Commonwealth
Idison,

(3) Two-region breeders, represented by the Nuclear Power Group studies,
the Babcock & Wilcox Breeder Reactor, and a sequence of conceptual
designs by the Oak Ridge National Laboratory Homogencous Reactor
Project.

92, GENERAL Prant Lavour axp DESIGN

9-2.1 Relation of plant layout to remote-maintenance methods. In lay-
ing out a homogeneous reactor plant, the designers, to achieve an optimum
arrangement, must simultaneously consider all aspects of the design,
including the requirements for remote maintenance, It is usual to start
with the high-pressure reactor system (the reactor vessel, circulating pump,
steam generator, and surge chamber and pressurizer), since there exists a
natural relationship between these items in elevation. The layout will
depend primarily on whether a one-region reactor or two-region reactor 1s
involved, since in the latter case special provision for removing the inner
core may be necessary. 1f it is feasible to construct the reactor vessel and
core tank as an integral all-welded unit, the layout of the system will be
considerably simplified. Otherwise, provisions will have to be built Into
the reactor vessel and the reactor system to remove the vessel and/or the
core tank.

Circulating pumps are vulnerable from the standpoint of long-term re-
liability, and extreme care must be given to their placement and anchorage
in the system layout. Installations and designs to date place the circulating
pumps in a position following the steam generator and the gas separator
and low in the cell in order to provide as low a temperature and least gas-
binding conditions as possible. These pumps, however, will operate at an
overpressure considerably in excess of saturation pressure, and if gas binding
does not prohibit, it may be desirable to place the circulating pumps at a
position more accessible for mamntenance.

The placement and design of the steam generators will be dictated to a
major degree by the maintenance philosophy adopted. One general

*Aeronutronic Systems, Inc., a subsidiary of Ford Motor Company.
9-2] GENERAL PLANT LAYOUT AND DESIGN 469

philosophy being considered uses many small steam generators in order to
permit easier removal and replacement when necessary. Another philoso-
phy considers the repair of the steam generators in siti, using remotely
manipulated tooling. One difficulty with this scheme will be the problem
of finding leaky tubes.

The steam generators are usually one of the bulkiest items of equipment
stalled in the plant and hence will largely determine the size of contain-
ment vessel and amount of shielding. Their location should be such that
some heat-removal capacity can be obtained by natural conveetion cir-
culation in the event of failure of the circulating pump.

In considering the layout of the surge chamber (which is normally also
the pressurizer) connecting piping must be as short as possible and the
diameter of the piping should be large for safe control of the reactor. If
a steam generator is used to provide high-pressure 1D-0 or H20O vapor, it
should be separated from the surge chamber, and preferably placed in a
location separate from the reactor compartment to facilitate maintenance.

9-2.2 Importance of specifications. To ensure that materials such as
type—-347 stainless steel and titanium and zirconium alloys meet the quali-
fications required for homogeneous systems, very rigid specifications cover-
mng strength, corrosion-resisting properties, impuct resistance, ete. must
be prepared.  To ensure leaktight integrity, specifications describing
acceptable weld joints and welding procedures are 1ssued. Such specifica-
trons will alzo describe the welder qualifications required. Since it is im-
perative that the main process piping system shall be absolutely ¢lean and
purged of any material which may poison the reactor or accelerate corro-
slon, cleaning procedures are a necessary part of the specifications.

0-2.3 Approach to an optimum piping system. The cost of the piping
system 1s one of the major items of expense, and its selection and arrange-
ment constitutes one of the major items of design., However, the pipe
diameters are generally determined on a maximum-velocity basis, deter-
mined by corrosion rates rather than from economie considerations. The
weight classification (i.e., pipe wall thickness) is sclected on the basis of
pressure, temperature, and corrosion rate for the proposed service life
of the reactor system using the appropriate design stresses from the ASMIZ
Code for the particular metal used. Other factors influencing piping layouts
are (a) provision for drainage, (b) provision for expansion, (¢) accessibility
and convenience of operation, (d) provision for support, and (e) the thick-
ness of insulation.

Long straight runs of high-temperature, high-pressure piping present
the main problem so far as expansion is concerned. Natural anchorages
should be noted, and at the same time, possible locations should be sought
170 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. 9

for special anchors needed to control expansion in accordance with the
design plan. The efficiency of the piping system layout depends largely
on the ability of the designer to visualize the over-all situation and to
select the best arrangement. The design of a piping system for minimum
holdup may be relegated to secondary importance compared with ease of
maintainability of the systems.

Piping joints. Piping joints for homogeneous reactor systems must be
capable of assembly and disassembly by remote methods and must have
essentially zero leakage. The first requirement implies some type of
mechanical joint such as used on the HRE-1 and HRE-2. The second
specification can only be guaranteed by an all-welded piping system, and
consequently an all-welded piping layout may be necessary for large-scale
homogeneous reactor systems. However, such a system requires a reliable
and easily manipulatable remote cutting and welding machine not yet
developed.

0-2.4 Shielding problems in a large-scale plant. Poor shield design can
lead to excessive cost and reduced accessibility for maintenance, Practical
shield designs are developed through the use of methods in the literature [1]
with particular attention to factors pertaining to the shield layout, such
as the arrangement of the piping and heat-exchanger system, materials se-
lection, radioactivity of the shutdown system, effect of radiation streaming
through openings, and the effect of the geometry of the radiation sources.

A number of proposed designs of large-scale homogeneous reactors use
a compartmentalized type of shield. This consists of a primary shield sur-
rounding the reactor pressure vessel to attenuate the neutron flux and re-
duce the radioactivity of auxiliary equipment, and a secondary shield sur-
rounding the coolant system. I'rom a shielding standpoint, the most
highly radioactive sources should be located near the center of the com-
partment, components of lower source strength should be arranged pro-
gressively outward, and equipment with little radioactivity should be
located to serve a dual purpose as shielding material where possible. High-
intensity sources containing primary coolant, which are poorly located
from a shielding standpoint, may be partially shadow-shielded. Equip-
ment requiring little or no maintenance and which can provide shielding
should be located around the outside of the secondary shield. Considerable
welght can be saved by contouring the secondary (coolant) shield (i.e.,
varying its thickness over the surface) to give closer conformance with the
specified permissible dose pattern. With respect to sample lines which
penetrate the coolant shield and contain radioactive materials of short
half-lives, the transport time from the primary coolant system to the out-
side of the shield should be made as long as practical to take advantage of
the decay of the coolant activity.
0-2] GENERAL PLANT LAYOUT AND DESIGN 471

9-2.5 Containment. Because of the possibility of release of highly
radioactive fuel solution from a homogencous reactor, such systems are
now belng designed to go within a containment vessel or to use doubly con-
tained piping.  The containment vessel must be designed to hold the
pre=sure resulting from expansion of the fluid and vapor contents of the
cquipment.  Such pressures may be of the order of 50 psi.  Although the
best shape of a containment vessel is either a eylinder or a sphere, such con-
frgurations present problems with respect to remote maintenance. To pre-
vent the penetration of the containment vessel by flying fragments which
may be released on failure of equipment, a blast shield can be placed around
the periphery of the containment vessel or relatively close to the equip-
ment.

9-2.6 Steam power cycles for homogeneous reactors.* In common with
other pressurized-water types of reactors, homogeneous reactors are handi-
capped by the high pressures required to prevent boiling in comparatively
low-temperature aqueous fluids.  Temperatures for steam generation in
homogeneous reactor power systems are limited to practical maximums of
500 to 600°F, and there are no significant opportunities for superheating
the steam with reactor heat. Separately fired superheating equipment,
using conventional fuels, may be expedient in some particular circum-
stances of plant size, load factors, and fuel costs but, in general, superheat-
ing by this means 1s not justified. Use of low-pressure saturated steam
limits the thermal efficiency obtainable in the heat-power cyele to maxi-
mum values in the range of 25 to 309%.

Homogeneous reactor systems circulate a hot reactor fluid to a steam
generator at essentially constant temperature; the temperature of the fluid
leaving the exchanger is varied with load by changing the temperature dif-
ference for heat transfer by controlling the pressure at which the water
boils in the steam generator. Since the steam pressure falls as the turbine
control valves open on increased electrical generator loads, the negative
temperature coeflicient for the reactivity causes the reactor power output
to be self-regulating to match the power demand on the plant.

The fuli-load steam pressure will be in the order of 450 to 60O psia, and
the near no-load pressure in excess of 1000 psia. The steam piping and
turbine casing must be dexigned for this maximum pressure rather than
the full-load pressure; design pressures of 1500 psia have been used for the
steam systems of the HRIE-1 and HRE-2.

Turbines designed for operation on saturated steam will cost more per
kilowatt of installed capacity than turbines designed for superheated steam
(see Chapter 10). The relatively low energy content, high specific volume
steam supplied to the saturated steam turbine throttle requires greater

 

*By R. C. Robertson.
472 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. 9

mass flow rates and flow areas for a given power output, adding to the cost
of the governor valves and the high-pressure stages. The low-pressure
stages, which are the most expensive, also must have more flow area, which
may require additional compounding, adding greatly to the cost. The
turbine efficiency will be somewhat less in a saturated steam turbine than
in one using superheated steam because of the greater amount of en-
trained moisture in the steam.

I'low delay tanks in the steam supply maing are considered necessary to
allow time for a stop valve to close in the event radioactivity is detected
in the steam flow from the heat exchangers. Heat losses in this equipment
tend to increase the moisture in the steam, and a separator may be re-
quired at the turbine inlet. The pressure loss in typical separators is about
5% of the inlet pressure and the leaving quality about 999, Some method
of moisture removal must be provided during the expansion process, either
internally in each of the low-pressure stages, or externally in one or more
separators located between turbine elements. Studies have indicated that
the optimum location for the first stage external moisture separator is at
109 of the throttle pressure. The presence of more moisture in the ex-
panding steam may require that the turbine be an 1800-rpm rather than a
3600-rpm machine.

As with steam power cyeles for other reactor types, an emergency by-
pass will probably be required to send the steam directly to the turbine
condenser in event of loss of turbine load. The condenser must be designed
to dispose of the full output energy of the reactor plant. The number of
stages of feedwater heating economically justified 1s probably limited to
three or four, since the temperature range in the cycle is not great.

Treatment of the water fed to the steam generators is a special problem
in that the water should be essentially free of chloride ions to reduce the
opportunity for stress-corrosion cracking in stainless steel parts of the
system, the water should be deaerated to control corrosion in the steam
system, it should be demineralized to reduce the radioactivity pickup of
the steam and in the heat exchanger blowdown, and additives may be
necessary to control the pH and to scavenge oxygen formed by radiolytic
decomposition of the water. Decomposition of these additives under radi-
ation poses problems not yet fully investigated.

Control of the water Ievel in the steam generator involves much the
same problems, due to steam bubbles that are experienced in conventional
hoilers, with the added complexity that the steam pressure increases as the
load on the plant decreases. Sizing of the ports in the feedwater regulating
valves must take this into consideration, and the boiler feed pumps must
be designed for the no-load, rather than the full-load head requirements.

Although some superheat can be obtained by recombining the decompo-
sition gases, it is doubtful if such a procedure is economical, owing to the
relatively small amount (59%,) of superheat obtained, and also because of
9-3] ONE-REGION U235 BURNER REACTORS 473

 

BT T T T T T T T T T

N
oo

26

Thermal Efficiency, %

24

 

 

 

 

350 400 450 500
Throttle Steam Temperature ,°F

F1c. 9-1. Effect of steam conditions on turbogenerator plant efficiency.

the desirability of minimizing gas production within the reactor. Unless
the superheat is more than 100°F, a saturated cycle with moisture separa-
tion may be equally as efficient and practical as a cycle using superheated
steam, provided that in either case the moisture in the turbine exhaust is
kept the same. It is also possible to superheat at the expense of throttle
pressure; while superheat normally is considered to increase the thermal
efficiency, this is not true if the inlet steam temperature is independent of
the amount of superheat. Also, the lower pressure associated with throttling
results in mereased turbine costs. Superheating by means of a conventional
plant does not appear economical.

In studies of homogeneous reactors, saturated steam cycles are assumed
in which 129, moisture is permitted in the last stages of the turbine.
Thermal efficiencies of such plants are shown in Fig. 9-1 as a function of
the steam temperature at the turbine throttle [2].

9-3. OneE-REecion U23% BurNER REACTORS

9-3.1 Foster-Wheeler Wolverine Design Study. In response to a re-
quest by the Atomic Inergy Commission for small-scale power demon-
stration reactors, the FFoster Wheeler Company proposed to construct an
aqueous solution reactor for the Wolverine Electric Cooperative in Hersey,
Michigan [3]. This proposal was rejected by the Atomic Energy Commis-
sion in October 1957 as a basis for negotiation due to increases in the esti-
mated cost of the plant (from $5.5 million to $14.4 million). The project
was canceled in May 1958 following a review of the design and estimated
costs. This review indicated that the cost of generating electricity would
be several times as great as that in Wolverine’s existing plant.

In December 1957 a group of engineers from the Oak Ridge National
Laboratory and Sargent and Lundy, with the help of Foster—-Wheeler, re-
designed the reactor on the basis of recent advances in homogeneous re-
actor technology and re-estimated its costs to be $10.7 million [4]. The
 

474 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. 9
Sample Valve, Reactor Press. Steam
rCondensate Tanks | Vessel #Drum
r_A Ty ! LT T e

 

 

 

 

 

TR T

"

 

 

 

 

    

 

 

 

 

 

3

 

 

 

 

 

 
 

 

 

 

 

P T N PR
P oy T
Condenser e A
i | ! ‘
™ .t
Cooling Pit | | E
o E

 

 

 

     
 
 
    
   

 

 

 

 

 

 

 

 

 

 

 

 

 

 

N N T =
Waste &  NOff Gas Plan Intermediate Heat

Decont. Cell Process Exchanger

Cell

E H
L Access Hatch 3;32 Control Room — —
i Waste-Decont. ,Off Gos & \<40 Ft. Dia. Sphere (I 1
/Ce“ Cell Press. S Busemem\\g | ‘ ‘

AT N A LA

~ fl ;[b A 'é - o A | Anal. > | Anal.}

: ; AR = . | £ Cell¥ | J L:
gl ey B3 o e
AL TE ol

a |
. Cooling Pit | o as
| | Heat Exchanger

b

 

 

 

 

Elevation-Section A-A

Fic. 9-2. Plan and sectional elevation of revised Wolverine reactor plant.

following section describes the revised reactor design. Iigure 9-2 shows a
plan and elevation sketch of the revised concept.

The fuel solution of highly enriched uranyl sulfate in heavy water is cir-
culated by a canned-motor pump located in the cold leg of the primary
loop and pressurized to prevent boiling and cavitation in the pump. The
steam generated in the heat exchanger is superheated in a gas-fired super-
heater, and the superheated steam drives conventional turbogenerating
equipment for the production of eclectricity.

The nuclear reactor plant is designed to permit initial operation at 5 Mw
with a single superheater-turbogenerator unit. By adding a second unit,
the capacity can be increased to 10 Mw. Doubling the electrical capacity
is thus accomplished without making any changes to the reactor other
than adjusting the operating temperatures and uranium concentration.

For 10 MwE operation, 31,000 kw of heat is generated in the reactor
under the following conditions: The hot fuel solution leaves the core at
300°C, is circulated through a heat exchanger, and returns to the reactor
at 260°C. The heat generated in the reactor is transferred to boiling water,
9-3] ONE-REGION UZ3? BURNER REACTORS

TABLE 9-1

DEesiaN DaTa ForR THE REVISED WOLVERINE PRIMARY SYSTEM
(10-MwkE OPERATION)

1. Core
Configuration Concentric outlet
Core diameter: inside thermal shields, ft 5
over-all, ft 6
Wall thickness, in. 3
Liquid volume, liters 2550

Power density, kw/liter

Core wall (inner thermal shield) 4
Average for system 6
Maximum 55
Initial fuel concentrations (critical at 300°C), m
1235 0.014
CuS0y 0.02
H2S04 0.02
Steady-state fuel concentrations, m
1235 0.030
Total U 0.034
CuS0y4 0.02
HaS04 0.025
N30y 0.017
2. Pump
T'uel flow rate, gpm at 260°C 2750
Head, ft 65
Approximate pumping power, hp 80
{assumes 509, over-all efficiency)
3. Heat exchanger
Shell diameter, in. 29
Tube diameter, in. 1/2
Tube waull thickness, in. 0.065
Number 1120
Approximate inside area wetted by fuel solution, ft? 4100
Steam tempcrature, °F 480
Log mean average temperature difference, °F 39
Over-all heat transfer coefficient, Btu/(hr)(ft?) 500
4. Pressurizer
Inside diameter, in. 56
Wall thickness, in. 3

Length of eylindrical portion 6 ft 9 in.
476 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHap. 9

TasLE 91 (Continued)

Concentric outlet

 

Volume of solution at low level, liters 150
Net gas volume (liquid at low level), liters 1400
Normal operating pressure, psia 1900
Normal operating temperature, °F 570
5. Piping
Nominal diameter, in. 10
Wall thickness, In. 1.125
Approximate total volume, liters 850
Maximum velocity, fps 17
6. Estimated power costs (10-MwE plant) Mulls/kwh
31-Mw reactor plant ($8,740,000)
Fuel burned 2.83
Fuel inventory @ 4% 0.67
(9000 kg D20 -+ 36.5 kg U235)
Fuel processing 2.46
Fuel preparation 0.62
D20 losses 0.30
Depreciation (@ 159 18.72
Operating costs 1.43
Maintenance costs 3.85
10-MwE superheater-turbine generator plant
($1,940,000)
Fuel {(oil) 0.69
Depreciation (@ 159 4 17
Operating costs 0.29
Maintenance costs 0.29
Total power costs 36.32

producing 116,000 Ib/hr of steam at 600 psia. For operation at 5§ MwkE,
the hot fuel solution would leave the reactor core at 276°C and return at
257°C, produeing 58,000 1b/hr of steam at 600 psi.

A pressurizer is connected to the outlet of the heat exchanger to pressur-
ize the system with oxygen to 1900 psia and to provide a location in the
primary system for the removal of fission-product and other noncon-
densable gases. The layout of the primary system is such as to permit heat
removal by natural circulation in case of pump failure.

A low-pressure system consisting of dump tanks, condenser, and con-
densate tanks Is incorporated to handle fluid discharged from the primary
loop and to furnish heavy water required to purge the canned-motor cir-
0-3] ONE-REGION U23° BURNER REACTORS 477

culating pump. Tacilities for adjusting fuel concentration and mamntaining
a continuous record of fuel inventory are also meluded.

Design date. Pertinent design information for the reactor systems and
components iy summarized in Table 91 and deseribed in the following
paragraphs.  Unless otherwise noted, all surfaces in contact with fuel solu-
tion are fabricated of type—347 stainless steel.

Equipment and system deseriptions.  Reactor vessel.  The single-region,
concentric-inlet and -outlet pressure vessel designed for 2500 psia 1n-
corporates two Inner concentric thermal shields to reduce gamma heating
effects in the outer pressure vessel. The thermal shields are constructed of
type—347 stainless steel and are 1 in. and 2 in. thick with inside diameters
of 5 ft 0 in., and 5 ft 5 in., respectively. Backflow through the vessel drain
line during normal operation provides some cooling of the outer thermal
shield.

Primary heat exchanger. The steam generator consists of a horizontal
U-shaped shell-and-tube heat exchanger with a separate steam drum.
These are interconnected with downcomers and rigers to provide natural
circulation of the boiling secondary water. Tuel solution is circulated on
the tube side of the heat exchanger, and the boiling secondary water 1s
circulated on the shell side. Feedwater is introduced into the liquid region
of the steam separating drum. All components in contact with secondary
water and steam are to be fabricated from conventional boiler steels.

Fuel circulating pump. A single, constant-speed, water-cooled, canned-
motor type pump is provided to maintain fuel circulation i the primary
loop. The rotating clements are removable through the top of the unit,
and may be removed without disturbing the piping connections to the
stuator casing or the pump volute. Regions of high fluid velocity in the
pump, including the impeller, are titanium or titanium-lined. A purge
flow of condensate is fed into the top end of the pump to reduce erosion and
corrosion of bearings, as well as to prolong the life of the motor windings
by reducing the radiation dose to the electrical installation. In the event
of pump faillure, the reactor will undergo a routine shutdown and the
fission-product decay heat will be removed by natural circulation through
the steam generator.

Pressurizer. A small sidestream of fuel solution is continuously directed
into the pressurizer, where it spills through a distribution header and drips
down through an oxygen gas space to the liquid reservoir in the bottom of
the vessel. The pressurizer liquid return line is connected to the suction side
of the primary-loop circulating pump. Oxygen is added batchwise to the
pressurizer to keep the fuel saturated at all times to prevent precipitation
of uranium. As fission-product gases accumulate in the pressurizer, they
are vented to the off-gas system, also in a batchwise operation.

Fuel makeup pump. Two diaphragm-type high-head pumps (one for
478 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. 9

standby) rated at 3 gpm at a pressure of 1900 psi are provided to add ura-
nium to the fuel solution and to fill” the primary system with fluid on
startup.

Dump tanks. The dump tanks, 48 ft long and 28 in. 1D, are designed to
remain suberitical while holding the entire contents of the primary system.
An evaporator section underneath cach of the vessels is provided to con-
centrate the fuel when necessary, and to aid in mixing the contents of the
tank.

Containment. The primary coolant system is enclosed in a 40-ft-diameter
spherical carbon-steel vessel, lined with 2 ft of concrete, mterconnected
with a 12-ft-diameter by 50-ft-long stainless-clad vessel housing the dump
tanks. The liner serves the dual functions of missile protection and struc-
tural support to withstand the loading of the external conerete. An addi-
tional 1/8-in. stainless steel liner 1s furmished to permit decontamination
of the primary cell, Sinee these vessels provide a net containment volume
of approximately 31,000 ft3, the vaporization and releasc of the reactor
contents results In a maximum pressure of approximately 105 psi.  Ac-
cordingly, the primary-cell containment vessel wull thickness 1s 15/16 n.
and the dump-tank containment vessel wall thickness 1s 9/16 in. A spray
system 1s incorporated in the design to quickly reduce the pressure within
the containment vessel by condensing the water vapor present.,

A bolted hatch is provided i the top head of the vessel to allow aceess
and removal of equipment for maintenance. A bolted manway is also
provided to permit entrance into the containment vessel without removing
the larger auxiliary hatch. In the event of o major maintenance program,
however, the top closure would be cut and removed for free access to the
primary cell.

Biological shielding. The plant biological shielding is indicated on the
general arrangement drawing (IMg. 9-2). The shielding for the primary
system, including the reactor core, consists of a 2-ft thickness of ordinary
concrete lining the inside of the primary-cell containment vessel and a
minimum of 7 ft of concrete surrounding the outside of the vessel, cooled
by a series of cooling-water coils located in the 2-ft-thick liner. The top of
the primary vessel is shielded with 6 ft of removable blocks of barytes
aggregate concrete (average density of approximately 220 ib/ft?) located
beneath the removable portion of the containment vessel.

A 2-ft-thick water-cooled heavy aggregate thermal shield i1s placed
around the reactor vessel to reduce the radiation level to approximately
that of the remainder of the primary system. The primary coolant pump
access pit, located inside the containment vessel, is constructed of 3% ft of
barytes aggregate concrete to permit pump removal after the primary cell
has been filled with water and the system drained and partially decon-
taminated. During periods of normal operation, the temperature of the
9-3] ONE-REGION U2%° BURNER REACTORS 479

concrete walls and floor of the pit is maintained at 150°F by cooling-water
coils.

Around each of the analytical and chemical processing cells there will be
a minimum of 4 {t of ordinary concrete with a maintenance gallery between
these facilities for access to, and operation of, the cells. Tlach of the two
analytical cells will be provided with thick glass windows adequate for
shiclding. The dump-tank cell will be shielded by a 5-ft thickness of
concrete.

Remote maintenance. Both dry and underwater removal methods are
proposed for remote maintenance of radioactive components in this system,
following practices similar to those developed for HRIG-2. All the equip-
ment cells are provided with stainless-steel liners to permit the cells to he
filled with ordinary water during maintenance operations. I'or removal of
the large components it is neeessary to move the container vessel cover
through the west end of the building to & temporury storage area. After
the primary vessel cover and top shield are removed, the system com-
ponents are accessible by crane and operations are performed with specially
designed long-handled tools.

9-3.2 Aqueous Homogeneous Research Reactor—feasibility study. A
preliminary investigation of the feasibility of an aqueous homogencous
rescarch reactor (HRR) for producing a thermal flux of 5 X 10'° neu-
trons ' (em?)(sec) was completed by the Oak Ridge National Laboratory in
the spring of 1957 [5]. The design considered is illustrative of a homogene-
ous reactor capable of producing high neutron fluxes for research and power
for the production of electricity. It consists of a 500-Mw (thermal) single-
region reactor with 8% enriched uranium as the fuel in the form of uranyl
sulfate (10 g of total uranium per kilogram of D»0) with sufficient copper
sulfate added to recombine 1009 of the radiolytic gases produced and
excess sulfuric acid to stabilize the copper sulfate, uranyl sulfate, and
corrosion-product nickel.

The system operates at solution temperatures of 225 to 275°C, and a
total system pressure of 1400 psia. Under these conditions a maximum
thermal neutron flux of 6.5 X 10'> neutrons/(ecm?)(sec) is achieved in a
10-ft-diameter stainless-steel-lined carbon-steel sphere.  Approximate
power densities are 2 kw/liter at the core wall, 35 kw/liter average, and
110 kw/liter maximum. After correcting for the effect of experiments, a
maximum thermal flux of about 3 X 10'° neutrons/(cm?)(sec) and a fast
neutron flux of about 5 X 10 neutrons/(ecm?)(sce) are available,

To minimize corrosion of equipment and piping in the external circuit,
all flow velocities are held to values below the critical velocities. Istimated
corrosion rates are 70 to 80 mpy for the Zircaloy-2 experimental thimbles
and about 10 mpy for the stainless-steel liner of the reactor vessel (based
on a maximurm flow velocity of 3 fps).
480 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. 9

TaBLE 9-2

HRR SteaM-GENERATOR SPECIFICATIONS
(OnE UniT)

 

Reactor fluids, forced cireulation (tube side)

Inlet temperature, °F 527
Outlet temperature, °F 437
Flow rate, Ib/hr 2,730,000
Pressure, psia 1400
Velocity through tubing, fps 10
Steam, natural recirculation (shell side)
Generation temperature, °F 417
Pressure, psia 300
Generation rate, 1b/hr 351,600
Heat load, Btu/hr 284,300,000
Heat load, Mw 83.3
Steam generator
Number of 3/8-in. 18 BWG tubes 3280
Effective length of tubing, ft 25.9
Heat-transfer surface, ft2 8330
Shell internal diameter, in. 38%
Shell thickness, in. 1%
Tube-sheet thickness, in, 5

Steam drum

Internal diameter, in. 36
Length, ft 16
Wall thickness, in. 13
Height above gencrator, ft 15

 

 

 

 

Fission- and corrosion-product solids, produced at a rate of approxi-
mately 20 Ib/day under normal reactor operating conditions, are con-
centrated into 750 liters of fuel solution by means of hydroclones with self-
contained underflow pots and removed from the reactor to limit the buildup
of fission and corrosion produets. This solution is subsequently treated for
recovery of uranium and D»0.

The temperature coefficient of reactivity at 250°C is approximately
—2.5 X 1073/°C and at 20°C is approximately —9 X 1074/°C, which, in
combination with fuel-concentration control, is adequate for operation
without control rods.

Reactor vessel. The 10-ft-1D spherical pressure vessel is designed according
to the ASME Unfired Pressure Vessel Code, with consideration given to
9-3]

ONE-REGION UZ3% BURNER REACTORS

481

TaBLE 9-3

Key DEsIGN PARAMETERS

 

Reactor type

Fuel type
Amount of U235
Uranium concentration
Total uranium
U235
CuS0y4 to recombine 1009} of gas

Maximum nickel concentration
Fuel-solution temperature
Minimum (inlet to reactor vessel)
Maximum (outlet of reactor vessel)
Average (system)
Fuel system pressure
Neutron flux (experimental)
Maximum thermal
Maximum fast in 1-in. diameter
cvlindrical converter
Power density
Maximum (at reactor center)
Average
AMinimum (at thermal shield)
Total heat generated
Reactor-vessel key specifications
Inside diameter
Vessel material

Total volume

Net fluid volume (approximate)
Experimental facilities

Horizontal

Vertical

Maximum inside diameter

Material

Minimum wall thickness

Maximum wall thickness

 

H 2804 to stabilize uranium and copper |
- 0.01m

!

 

| UO2SO4 -— Dzo -+ CU.SO4 + I‘IzSOz;

Single-region, ecirculating-fuel, homo-
genecus

45.8 ke

10 g/liter at 250°C
0.8 g/liter at 250°C
0.02m
0.02m

225°C
275°C
250°C
1400 psi

3-4 X 1015 n/(em?)(see)
4 X 10 to 1 X 10" n{em?)(sec)

110 kw /liter
34 kw/liter

2 kw/liter i
500 Mw

10 ft

Carbon steel clad with type-347
stainless steel

14,800 liters

12,000 liters

6

1

6 1.
Zircaloy-2
3/4 1.
!in,

 

 

continued
482 LARGE-SCALE HOMOGENEOQOUS REACTOR STUDIES [cHaP. 9

TapLe 9-3 (Continued)

 

External system
Material Type-347 stainless steel, HRDP speci-

fications

Fluid volume (external system only) | 34,000 liters

Allowable velocities

225°C 10-15 fps
250°C 25-35 fps
275°C 30-40 fps
Reactor control Negative temperature coefficient of
reactivity, changes in concentration
of fuel
Heat dissipation Generation of approximately 125 Mw

of electrical power

 

 

 

 

the special problems introduced by the heating of the shell from radiation
absorption and by the neccessity of penetrating the shell for insertion of
experimental thimbles. The proposed vessel is fabricated of a carbon-steel
base material with a type-347 stainless steel cladding on all surfaces
exposed to fuel solution.

The fuel solution enters the vessel through two 24-in. nozzles, sized for a
fluid velocity of 10 to 15 fps, flows upward through the vessel, and exits
through two 18-in. nozzle connectors in the top, sized for a fluid velocity
of 30 to 40 fps. A diffuser screen, serving also as part of the thermal shield,
is placed at the entrance to the reactor vessel.

A stainless steel blast shield is placed around the reactor vessel to con-
tain fragments of the vessel in the event of a brittle failure, and cooling coils
are wrapped around the blast shield to control the pressure-vessel tem-
perature.

Heat exchangers (steam generators). Six heat exchangers of 83.3 Mw
capacity each are required to dissipate the 500 Mw of heat generated in the
reactor. The design of these consists of a lower vaporizing shell connected
to a steam drum at a suitable elevation to promote natural eirculation by
means of risers and downcomers welded to the shells. Specifications are
summarized in Table 9-2.

Pressurizer.  The pressurizer surge chamber, constructed of 24-n.
schedule-100 pipe provides the necessary 1500 liters of surge volume.
Steam is provided in a small high-pressure steam generator physically
separated from the pressurizer surge chamber. Space limitations and ac-
cessibility problems make this separation desirable.
9-3] ONE-REGION U23% BURNER REACTORS 483

Biclogical Shield Outline

Ph Heat Exchangers  Crane Rail
Cell For Remote '

Operated Tools /

     
   
 
     
    
    

Concrete Wall
Biological Shield

Thermal Shield

   

Circulating
Pump

  

Transfer Carriage

 
  

‘ressurizer .

N

    
  
  
    
 
 
 
   

1N

Maintenance Building

  

Reactor
i

Surge
Chamber
Experimental Thimbles

8 0 B 1624 32

\ Scale in Feet

« -

Fic. 9-3. Homogeneous Research Reactor layout plan view.

System design. Two 17,650-gpm pumps mounted on the outlet pipes of
the heat exchanger circulate the reactor solution around the primary cir-
cuit. Saturated steam at 300 psia is generated at a rate of 2.11 X 10° lb/hr
and is used to generate 133,000 kw of gross electrical power at a cycle
efficiency of 26.5%. A net power generation of 125,000 kw will be delivered
at the station bus bars, approximately 6% being required for station
auxiliaries. Feedwater, consisting of D20 from the condensate tank, is
supplied to the steam generator through an economizer by means of a
0.5-gpm feedwater pump. The reactor does not contain a letdown system
for separating and recombining radiolytic gases, since 100% internal re-
combination will be achieved by means of internal copper catalyst. Key
design parameters are summarized in Table 9-3.

Conceptual layouts of reactor complex. Preliminary conceptual layouts
showing the relation of the items pertaining to the nuclear reactor com-
ponents are given by Figs. 9-3 and 9-4.

Figure 9-3 is a plan view of the reactor complex, indicating the general
relation of the reactor pressure vessel and its auxiliaries to the heat ex-
changers and circulating pumps. Shielded cubicles around the reactor
provide a means for handling the experimental thimbles. The outer
diameter of the containment vessel around the cubicles is approximately
60 ft. Approximately 6 ft of high-density concrete is placed around the
reactor area, with an additional 3 ft around the periphery of the contain-
ment vessel.
484 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. §

 

 
 
 
 

_#Crane
L Le-Shielded Control Cab For

3 } "’ | —
: . Clrculcnng P Crane Operator
L s - umw, gzj;

 

 

 
 
  
 

Removable Plugs

1 smge s A
Ch:razer éfflfl"filprasunzer < 11 FAl

i|[Reacto

{ Concrete
Wall

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Transter Carriage

 

 

   

 

 

 

 

 

 

 

 

 

 

 

8 0 8 16 24 32

Scale in Feet

Fic. 9-4. Homogeneous Research Reactor layout sectional elevation.

A sectional elevation of the reactor complex is shown in Fig. 9-4. The
arrangement of the heat exchangers relative to the reactor vessel is such
that natural circulation through the system will be promoted in the event
of pump failure. Since the centerline of the reactor vessel is located at 30
to 36 in. above the operating-floor level for convenience in experimentation,
the containment vessels for the heat exchangers and circulating pumps
are above ground. The containment vessel for the reactor is a vertical
cylindrical tank. Two separate horizontally mounted containment vessels,
each 60 ft in diameter, house the heat-exchanger equipment. The dump
tanks are directly below the heat-exchanger containment vessels. Means
for limited access to those portions of the dump-tank system which will
require periodic maintenance, such as dump valves, is provided.

Unique design features. Five horizontal in-pile thimbles spaced equally
on the midplane of the reactor opening into cubicles, and one vertical
nozzle, opening from the top of the reactor, are included in the design.
Figure 9-5 shows the location of the thimbles relative to the containment
vessel and cubicles, and the shield arrangement. As shown by Fig. 9-5,
piping to the heat exchangers passes through one of the hot-cell working
areas. Consequently, this area is not usable for experiments, but contains
the pressurizer and other items which must be adjacent to the reactor but
removed far enough from the reactor cell to permit maintenance.

Maintenance concept. Both dry and underwater removal methods have
been mvestigated for the HRR; however, both schemes present difficult
design problems. Dry-maintenance philosophy, chosen on a somewhat
arbitrary basis, has been followed in the layouts presented herein.

Maintenance of equipment in the reactor compartment is expected to
0-3] ONE-REGION U235 BURNER REACTORS 485

Experiment Thimble

Thermal Shield
Biological Shield

      
  

Blast Shield

Reactor

 
 

‘/ Equipment
y Cubicle

Experimental
Facilities Cell

Fic. 9-5. Plan view of Homogeneous Research Reactor, showing pressure vessel
shielding and cells for remote handling of experiments.

be largely confined to the reactor auxiliary equipment and to the experi-
mental thimbles and equipment. The reactor vessel itself is designed for
the life of the system with thickness for the pressure-vessel wall and corro-
sion liner selected accordingly on the basis of existing corrosion data.

The handling equipment in the reactor containment vessel consists of a
revolving-type crane with a shiclded cab for the operator and provision
for remote operation from outside the shielded area using commerecial,
remotely operated television cameras. Access from above to any part of
the area is thus possible and all flanges and pipe disconnects are faced
upward to facilitate removal.

A horizontal traveling crane, also with a shielded control cab and remote-
operation control, is provided in each of the containment vessels for
removal of the heat-exchanger equipment. Flanges connecting the circu-
lating pumps and heat exchangers with the main piping are faced hori-
zontally in these installations. All flanges are grouped at one end of the
area and bolts are removed by means of remotely manipulated tools from a
shielded cell. The heat exchangers, mounted on wheeled dollies guided by
486 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [caaPr. 9

tracks, can be moved horizoutally along the track and onto another track
section which can move transversely. Trom this section, the heat ex-
changer is moved through a large air-lock type of door at the end of the
containment vessel to the maintenance area. The circulating pumps are
designed so that the pump impeller and motor windings may be removed
vertically without removing the pump casing.

During any part of the maintenance procedure, the system is shut down
and drained and the piping and equipment decontaminated as thoroughly
as possible. Shutoff valves of a size and type suitable for the piping of the
HRR have not been developed.

9-3.3 The Advanced Engineering Test Reactor. A study was completed
in March 1957 by Acronutronic Systems, Ine., to select a reactor system
for an advanced engineering test reactor (AETR), with seven major loop
facilities providing a thermal-neutron flux >2 x 101% neutrons/ (em?) (sec)
[6]. To obtain the required flux level while keeping the power density
low, only heavy water-moderated reactors were considered. Comparisons
of two heterogeneous and one homogeneous type, and comparison of single
and multiple reactor installations, led to the conclusion that a single homo-
geneous reactor provides the greatest flexibility and is the most economical
system for research at high neutron fluxes. A description of the homogene-
ous AISTR reference design by the Aeronutronic group is given below:

Description of reactor. The 500-Mw reactor consists of a large core
operating at moderate temperature and pressure and containing a D20
solution of 109, enriched uranyl sulfate (10 g total U/liter). The reactor
design, which was based upon the design and operational experience of
the HRE-1 and HRE-2 and upon a design study for a homogencous
research reactor by ORNL, features continuous fission-product removal
and fuel addition to maintamn the total contained excess reactivity at an
essentially constant level. In the center, or loop region, the unperturbed
thermal-neutron flux is approximately 6 X 10'® neutrons/(cm?)(sec).

The reactor vessel is a spherical, stainless steel contalner with an internal
diameter of 8 ft and a wall thickness of 3/4 in., contained in a eylindrical
pressure vessel with balanced pressures inside and out. The design is such
that the test loop and coolant circuit tubes emerging through the lid of the
pressure vessel can be disconnected, the packing glands at the bottom of the
pressure vessel removed, and the entire reactor core vessel can be lifted
out of the main container. The thin walls of the core vessel give it a low
gross weight, enabling 1t to be lifted conveniently.

The cylindrieal pressure vessel, 10 ft in diameter, 123 ft high, and 3 in.
thick, is constructed of carbon steel to the specifications of the unfired
pressure vessel code for an internal working pressure of 500 psia.

Operating parameters of the AETR are summarized in Table 9-4.
9-4] ONE-REGION BREEDERS AND CONVERTERS 487

TaBLE 9-4

Key Desiey Paravierers (ALETR)

 

Type Thermal, homogeneous
Total heat power 500 Mw
Fuel Aqueous solution of UOgS04 in D20
Fuel content
Core 80 kg uranium, enriched to 109

6.5-8.5 kg U285

7500 liters fuel solution
System 375 kg uranium

37.5 kg U235

37,500 liters fuel solution

Fuel temperature: Inlet 98°C
Outlet 153°C
System pressure 500 psia

Flux (no test loops) : Maximum thermal | 6 X 10' n/(cm?)(sec)
Power density distribution (with no

loops): Maximum (center) 220 kw/liter
Average 70 kw/liter
Minimum (wall) 4 kw/liter

 

 

 

 

9-4. O~NE-IREcIoN BREEDERS AND CONVERTERS

9-4.1 The Pennsylvania Advanced Reactor U?33-thorium oxide refer-
ence design. The Pennsylvania Power and Light Company and the
Westinghouse Electric Corporation joined forces in November 1954 to
survey various reactor types for power generation. The results of the survey
indicated the potential of the aqueous homogeneous reactor to be exceed-
ingly encouraging and led to the formal establishment of the Pennsylvania
Advanced Reactor Project in August 1955 to study the technical and eco-
nomic feasibility of a large aqueous homogeneous reactor plant for central
service application having an electrical output of at least 150,000 k.

Two reactor plant reference designs were completed, and preliminary
equipment layouts and cost estimates of these two plants were prepared
[7,8]. In the first design it was proposed to use overhead dry maintenance
with the equipment housed in a vertical cylinder 124 ft in diameter and
175 {t long. By incorporating shutoff valves in the system, any one of the
four main coolant loops could be isolated in case of an equipment failure to
permit the remainder of the plant to continue operation. At a convenient
time, the plant would be shut down and the defective item removed and
replaced with remote equipment such as heavy-duty manipulators, special
488 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [crAP, ©

Personnel Access

Shielded Equipment Cab
Corridor /

| iEH:-:w Steel Shell

LTJ ’E Personnel Access
~ . Y Corridor Door
b T

 
 
  

 

Primary Pump

 

 

 
    
  
   
 
  

 

—

Surge Tan :
TN

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

!
Ifeactor,b\ Auxiliary
Pressurizer I Equipment Cells
( Gasfld\i 4 i 7
Separator = . :f
vl Iy )
l A4 \A!ir Lock
I Access to
S'ream—_fl Maintenance
Generator Building

 

 

dry

B

 

 

 

 

 

 

 

| Personnel Access
- Corridor

 

 

Personnel Access S h
Corridor pAis

Steam Line

Shielding Window

 

 

 

 

F1g. 9-6. Plan view of Pennsylvania Advanced Reactor Reference Design No.
1A (courtesy of Westinghouse Electric Corp).

jigs and fixtures, and television viewing equipment lowered into the com-
partment. HHowever, it was concluded that such a scheme would be ex-
tremely expensive. Therefore, a new design (Reference Design 1A) was
prepared based on the specifications embodied in the following recom-
mendations:

(1) Elimination of stop valves in each loop and abandonment of the idea
of partial plant operation.

(2) Compartmentalization of equipment depending on type and level of
radioactivity.

(3) Use of semidirect maintenance techniques wherever possible.

(4) Modification of the vapor container design to permit personnel
access in limited areas during plant operation.

(5) Increased emphasis on design of components to minimize difficulty
of maintenance.

Iigures 9~6 and 9-7 show a plan and cross-sectional elevation of Refer-
ence Design 1A. In this design, a mixed-oxide slurry of a concentration of
about 260 g/kg of D20, corresponding to a solids concentration of approxi-
mately 39, by volume, is circulated through the reactor vessel releasing
550,000 kw of thermal power, which in turn yields 150,000 kwE. Leaving
the reactor vessel, the slurry branches into four parallel identical loops.
94] ONE-REGION BREEDERS AND CONVERTERS 489

 

 

 

 

 

  

 

 

 

 

 

 

100 Ten Crane
Surge Tank
Reactor :
N SN % Steel Shell
Shielded Equipment Cab Bl ~ a7
\ Steam To
Shielded Crane Cab _'-' ANy T e T\ ¢~ Turbine
! Tn) o XN
. ot 2] Steam Generator
Primary Pump v Aa
N Gas Separator
4. Personnel
o s Access
- e Corridor
. i
> s >
- f Ly -
L VL - |
7 A %
T ak . i'
3
ks 7.7
ae AT =
T 1 = F= = ; W ’ R
N < — = = = | \
=5

3 =
Tk .. » 5 eI
o g b e L aka M

Frc. 9-7. Cross section through main loops of proposed Pennsylvania Advanced
Reactor (courtesy of Westinghouse Electric Corp.).

Each loop containsg a circulating pump, a gas separator, and a steam gen-
erator. The system is pressurized with 2000 psia steam generated in a DO
steam generator connected to a surge chamber mounted in close-coupled
position to the reactor vessel. The major portion of radiolytic gases is
recombined internally; the remainder (~109;) is left unrecombined in
order to purge the system of xenon and other gaseous fission products.
These gases are removed from the main stream by a pipeline gas separator
to a catalytic-type recombiner. The recombined heavy water is used to
wash the primary-pump bearings and as makeup water to the steam
pressurizer.

A small bleed stream is concentrated in the slurry letdown system and
delivered to a chemical processing plant where the uranium and thorium
are recovered by a thorex solvent extraction process. The chemical plant
1s designed for a small throughput and low over-all decontamination factors.
Although the rates of flow to the auxiliary systems are small compared
with the 18,000,000 lb/hr rate of circulation in the primary system, these
auxiliary systems contribute the major part of the complexity of the plant
and a large fraction of its cost.

The reactor plant layout shown in Figs. 9-6 and 9-7 consists essentially
of a horizontal steel cylinder 125 ft in diameter and 132 ft long with 7-ft-
thick biological shielding walls completely separate from the vapor con-
490 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. 9

 

 

 

 

 

  
    
 
 
  
  

 

T f—"k:?’fi
. A .
{LY—:* s erepane i Primary Loop
bt ER A 41‘ Compartment
i .
; < —
1
— .
= U ! Mechanical Arm
q
R i
PR
L. o1 Y0 7 G —
I 7% Retractable Rail r')-’-i‘w /1
---------- P |
Work PO ST R i \
Area SRR
Heav *—— Master-Slave
Cor;cr;e K Manipulator
Purge
L Water

   
  
 

_Shielding  Vent Line

Window

 

 

 

 

 

Control Cooling
Console . O Water
For o
Mechanical A\
Arm — TG
‘L': S
—Closed Position 2300 V. Supply
e

Primary Pump

 

<")/ \/ Discharge
N }\.,.///
Suction

Fig. 9-8. Primary circulating pump maintenance, Pennsylvania Advanced
Reactor (courtesy of Westinghouse Electric Corp.).

tainer. The reactor vessel is shielded separately; however, the four primary
coolant loops are contained in one large compartment with no shielding
between the separate loops. Auxiliary equipment is contained in separate
compartments, the equipment being segregated according to the type and
level of radioactivity after shutdown. All four of the primary coolant loops
are designed with polar symmetry to permit any component to be used as a
replacement part in any of the four loops, and any special equipment re-
quired to be equally adaptable to all four loops. In addition, like pieces of
equipment have been grouped to permit the use of relatively permanent
maintenance facilities designed into that particular area. Personnel
access corridors are provided to permit limited access to certain areas inside
of the vapor container during full power operation of the reactor.

Dry-maintenance operations are accomplished primarily through the
use of a 100-ton, shielded-cab crane which traverses the length of the reactor
container. Since the cab can be occupied during operation, the crane
serves as a remote tool for handling heavy shield blocks and removing and
replacing equipment. The design is based on an all-welded piping system
and removal of any item requires a remote cutting and welding machine
not yet developed.
9-4] ONE-REGION BREEDERS AND CONVERTERS 491

/ Steam Qutlet

 

   
    

 

 

 

 

 

 

 

 

 

 

 

7

| _———Steam Separator

 

 

—~t—— Drain

 

Liquid Levef
Connections

 

 

 

 

 

Feedwater Nozzle

 

 

U Tubes

 

 

 

 

 

 

 

Shell Drain

Slurry Qutlet
Slurry Inlet

 

 

 

 

 

 

 

 

 

 

Fig. 9-9. Steam generator for Pennsylvania Advanced Reactor (courtesy of
Westinghouse Electric Corp.).

Because of the vulnerability of the circulating pumps and steam gener-
ators, special modifications are provided to permit these items to be re-
paired in place. A maintenance facility for repair of the primary circulating
pumps is shown in Fig. 9-8. This consists of two mechanical master-slave
manipulators inserted through the shielding wall adjacent to the pump,
and a mechanical arm which may be placed on two retractable rails eanti-
levered from the shielding wall. Visibility is obtained by a glass shielding
492 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. 9

      
 
    
 
 
 
 
    
   

Steam
Generator

Periscope

Tube Bundle

Shielding Window

Leveling Jack

Periscope

Manipulator

Fig. 9-10. Facility for remote maintenance of Pennsylvania Advanced Reactor
steam generator (courtesy of Westinghouse Electric Corp.).

window located bencath the manipulators. The window is designed to be
an cffective shield only during plant shutdown, and will be covered by
iron shutters during plant operation to provide neutron and thermal
shielding. A second shielding wall is located behind the work area to make
up for the thin wall at this point.

The pump is provided with flanged joints with bolts and all other con-
nections at the top for easy accessibility. The low-pressure cooling water
connections are easily disconnected with the manipulators. The high-
pressure purge line and vent lines, however, must be disconnected through
flanges or by cutting and rewelding. The large flange bolts on the pump
are provided with centrally drilled holes into which electric resistance
heaters can be inserted with the mechanical master-slave manipulator.
The heated bolts are easily loosened with a power-driven wrench held by
the mechanical arm and removed with the master-slave manipulator.
A lifting fixture is then lowered from the overhead crane and attached to
9-4] ONE-REGION BREEDERS AND CONVERTERS 493

the pump flange and pump internals, which are then pulled from the pump
volute easing. The pump is reinstalled in reverse order.

The steam generator shown in Fig. 9-9 uses inverted vertical U-—tubes
and has an integral steam separator. The unit is 6 ft in diameter and
has an over-all length of about 40 ft. Because of the physical size and cost,
it is not considered practical to use the spare-part replacement philosophy
for this component. Instead, the design of the steam generator and the
over-all plant layout is such that remote maintenance in place is possible
without requiring a prohibitively long shutdown of the plant.

Figurc 9-10 illustrates the proposed semiremote method for locating and
repairing a leaky boiler tube.

The facility consists of a manipulator unit mounted on a horizontal
rack which drives the unit through the shielding wall into access holes in
the steam generator head. The manipulator is used to carry and position
a detector for locating a leaky tube and the necessary tools for plugging
and welding the tube. The faulty tube is prepared for welding by a spe-
cially designed grinding machine positioned and supported by the manipu-
lator. The grinder will automatically shape the tube for a plug and the
seal weld which will be made with an automatic welder. This equipment
can be moved from one cell to another as needed; thus all four steam
gencrator tube sheets can be maintained by semiremote methods.

0-4.2 Large-scale aqueous plutonium-power reactors. Studies of the
feasibility and cconomics of producing plutonium in homogencous reactors
fueled with slightly enriched uranium as UO2S804 in D»0 were carried out
by the Oak Ridge National Laboratory [9-11], by the Argonne National
Laboratory [12-13], and by others [14-15]. The studies were all based
on one-region converters constructed of stainless steel utilizing spherical
pressure vessels ranging in size from 15 to 24 ft in diameter. The design
and operating characteristics of typical reactors considered in the studies
are summarized in Table 9-5.

The general conclusion reached was that aqueous homogeneous reactors
are potentially very low-cost plutonium producers; however, considerable
development work remains before large-scale reactors can be constructed.
The major problem is due to the corrosiveness of the relatively concentrated
uranyl sulfate solutions used in such reactors, which requires that all the
equipment in contact with high-temperature fuel be made of titanium, or
carbon-steel lined, or clad with titanium. The development of suitably
strong titanium alloys, bonding methods, or satisfactory steel-titanium
joints has not yet proceeded sufficiently to consider the construction of
full-seale plutonium producers. Alternate approaches, such as the addition
of LS04 to reduce the corrosiveness of stainless steel by the fuel solution
(see Chap. 5), show promise but also require further development.
TABLE 9-5

CHARACTERISTICS OF LARGE-ScaALE AQUEroUus PrutoNiuM PRODUCERS

 

Source of Data

 

ORNL-CF-

ORNI~1096

 

 

 

ORNIL-835 | ORNL-1096 52-8—7 Rev. Data ANT.-4891 CEPS-1101

Date Oct. 1950 Dec. 1951 Aug. 1952 1955 Dec. 1952 May 1952
Power, Mw (thermal) 1000 2000 1028 2000 1028 1064
Net electric output, Mw 230 470 228 435 211 211
Reactor core diameter, ft 24 15 15 15 15 12
Pressure vessel thickness, in. 5.3 4.5 7 - 7 -
Fuel concentration, g U/liter 115 250 250 250 250 290
Initial fuel enrichment, 9, U233 0.80 1.05 1.075 1.08 1.12 1.2
Fuel inventory, metric tons 22 30 35 35 33 53
D20 inventory, metric tons 208 130 105 150 155 160
Plutonium production rate, g/MwD 0.97 1.05 1.09 1.05 1.12 1.0
Liquid inlet temperature, °C 208 206 200 200 210 208
Liquid exit temperature, °C 250 250 250 250 250 250
System pressure, psi 1000 1000 1000 1000 1000 1000
Steam conditions, °F/psia 480* /200 380,200 385/210 380,200 370/175 382/200
Capital cost data ($ millions)

Reactor plant 37 50 36 45 50 49

Turbogenerator plant 15 37 31 60 47 39

Total 52 87 67 105 97 88
Unit costs, $/kwhki 226 185 294 232 460 415

 

 

 

 

 

 

 

 

*Quperheated 100°F

Po¥

HIVOS-HDUVL

SHICNLS HOLOVHY SA0HNADONOH

6 "dVHO]
9-4] ONE-REGION BREEDERS AND CONVERTERS 495

9—4.3 QOak Ridge National Laboratory one-region power reactor studies.
Preliminary designs of intermediate and large-scale one-region reactors
have been carried out at the Oak Ridge National Laboratory for the pur-
pose of establishing the desirability, relative to two-region reactors, of such
plants for producing power [16,17]. A description of the design of a typical
large-scale plant with a capacity of approximately 316 net Mw of electricity
follows.

The uranium-plutonium or thorium-uranium fuel is pumped at 130,000
gpm through a 15- to 20-ft-diameter core, where the temperature is in-
creased from 213 to 250°C. Slurry leaving the core flows through four large
gas separators, where D and O3 are separated and diluted with helium, O,
and D20 vapor, and then to eight 160-Mw heat exchangers. The slurry is
cooled in the exchangers and returned to the reactor by eight 16,000 gpm,
canned-motor pumps.

Gas and entrained liquid from the separators pass through four parallel
circuits into high-pressure storage tanks, where the entrained liquid is
removed to be returned to the reactor. The D2 and Oz are recombined on a
platinized alumina catalyst and cooled in 17 Mw, tubular heat exchangers
which condense the 76 gpm of excess D20. The cooled gases are recirculated
to the gas separators, and the condensate returns to the fuel through the
rotor cavities of the pumps, the demisters, and the high-pressure storage
tanks.

The slurry fuel is expected to contain 100 to 300 g/liter of uranium as
either oxide or phosphate, and thorium as either oxide or hydroxide sus-
pended in D»0. Estimates of gas generation rates have been based on the
use of UQy3 platelet particles 1 micron thick and approximately 1 to 5
microns on a side. The (p,o value was taken as 1.3 molecules of D20
disintegrated per 100 ev of energy dissipated in the slurry, postulating
that 809 of the fission fragments escape from the oxide particles. It is
possible that much lower G-values will be obtained in representative experi-
ments and that the size of the gas system can thereby be reduced con-
siderably.

A 15-ft-diameter sphere operated at 1000 psi and 250°C requires a

s-in.-thick wall to keep the combined pressure and thermal stress

below 15,000 psi. Carbon steel, clad with stainless steel, is specified as the
material of construction for the vessel. The thermal shield may be stainless
steel or stainless-clad carbon steel, depending on which would be the less
costly. The weight of the vessel and thermal shield 1s 150 tons, while 75
tons of slurry containing 200 g U/liter are required to fill the vessel.

The estimated cost of the 316-Mw plant was $14-19 million for the
reactor portion and $44 million for the power plant scetion, which cor-
responds to a unit cost of $185-200/kwkE.
496 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cmaP. 9

TABLE 9-6

OprERATING ConpITioNs—180 Mw ErecTrICAL PranT

 

 

 

 

 

 

 

 

 

 

Core system Blanket system
General
Thermal power, Mw 360 280
Fluid U02804-D20 sol. ThO2-D20 disp.
Concentration, g/liter U233 1.80 8.00
Th232 — 1000.00
Primary system—pressure, psia 1800 1800
Reactor inlet temperature, °C 258 258
Reactor outlet temperature, °C 300 300
System volume, liters 28,760 41,785
Maximum fluid veloeity, fps 33.6 28.7
Loop head loss, psig 58 80
Fuel
Total fuel in system, kg, U233 51.8 334.3
Th232 — 41,785
Fuel burnup, g/day,
U233 447 332
Th232 (consumption) — 828
Fuel removed
grams U?233/day 210 498
kilograms thorium/day — 62.0
liters/day 117 62.0
Primary circulating pumps
Number 3 3
Capacity, gpm 11,300 9600
Differential pressure, psi 58 80
Estimated efficiency, % 60 60
Horsepower 650 750
(Continued)

9-5. Two-REGION BREEDERS

9-5.1 Nuclear Power Group aqueous homogeneous reactor. A study of
power stations ranging in size from 94 to 1080 megawatts of net electrical
generating capacity was carried out by the Nuclear Power Group [18].
The plants considered utilized a two-region Th-U233 reactor. While
several plants of different electrical capacities were studied, emphasis was
9-5] TWO-REGION BREEDERS 497

TaBLE 9-6 (Continued)

 

 

 

 

Core system Blanket system
Steam generators
Number of units ' 3 3
Surface sq. ft./unit 14,800 12,650
Feedwater inlet temperature, °F 405 405
Steam temperature, °F 480 480
Steam pressure, psia 566 566
Thermal capacity/unit, Mw 119 93
Gas condenser
Type: Horizontal, straight-tube,
single-pass, shell-and-tube ex-
changers with internal elimi-
nator
Number of units 1 1
Surface area, ft2 800 800
Feedwater inlet temperature, °F 405 405
Steam temperature, °F 480 480
Steam pressure, psia 566 566
Thermal capacity/unit, Mw 3.5 3.5

 

 

 

 

 

directed toward a plant having a net electrical capacity of 180 MwE.
Pertinent operating conditions of this plant are listed in Table 9-6.

The reactor consists of a 6-ft-diameter spherical core surrounded by a
2-ft-thick blanket enclosed in an 11 ft 4 in. ID stainless-clad carbon steel
pressure vessel with a wall thickness of approximately 6 in. The pressure
vessel has a bolted head to permit removal of the concentric-flow core tank
if necessary. The fuel solution enters the core through a 24-in. inner pipe
and exits through an annulus of equivalent area between two concentric
pipes forming the inlet and outlet connections for the core tank. One
mechanical joint is required to attach the zirconium core tank to the
stainless steel outlet pipe. The slurry enters the blanket through a 24-in.
connection in the bottom of the pressure vessel and exits through three
14-in. connections located near the top of the vessel.

Thermal shield. The 4-in.-thick thermal shield to protect the pressure
vessel from excessive radiation is provided in the form of two 2-in.-thick
stainless steel plates. A 2-in. space is maintained between these plates and
between the thermal shield and the pressure vessel. Sufficient flow of the
498 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHap. O

slurry is maintained between and around the shield segments to ensure
proper cooling.

Vessel closure. The bolted head closure utilizes two Flexitallic gaskets
(asbestos encased in stainless steel), having a low-pressure leakoff between
gaskets. The internal diameter of the closure is slightly greater than 6 ft,
to allow for the core tank removal. Similar bolted joints are provided in
the inlet and outlet piping connections to the core as well as in the core
tank dump line.

Steam generators. Six steam generator units each consisting of two heat
exchangers connected to a common steam drum are required, three for the
core system and three for the blanket heat removal. By using U-bend
tubes in the exchangers, the need for an expansion joint in the shell or in
the connection to a floating tube sheet is eliminated. This adds reliability
to the unit, since any expansion joint subject to even infrequent work is a
potential and likely source of trouble.

Utilizing the compartmentalized concept in the heat exchangers offers
added reliability, ease of fabrication, and a means by which maintenance
of the units becomes practical. The individual “bottles,” consisting of 19
U-tubes attached to their tube sheets in the eccentric pipe reducers by
rolling and welding, can be fabricated and tested as units before installation
in the exchanger. The drilling of the “bottle” tube sheets presents prac-
tically no difficulty because they are only 52 in. in diameter. Similarly, the
drilling of the exchanger head for insertion of the 2-in. inlet and outlet
pipes to the “bottles” presents no unusual fabrication problems.

Although the goal is theoretically “leakproof” heat exchangers, pro-
visions are incorporated for maintenance. This has been done in the
compartmentalized concept. Should a leak occur, it is practical to seal off
the “bottle” in which the leak occeurs by plugging the 2-in. inlet and outlet
pipe connections, rather than remove an entire heat exchanger.

Primary circulating pumps. The hermetically-sealed-motor, centrifugal
pumps required to recirculate the core and blanket fluids are vertical, with
the main impeller mounted on the lower end of a shaft on which also 1s
mounted the motor rotor. The motor rotor and bearing chamber are
separated from the impeller and volute by means of a labyrinth seal.
D50 from the high-pressure condensate tank is injected into the bearing
chamber and continuously flushed through the labyrinth, thus minimizing
corrosion on the rotor and bearing parts. Both the radial and thrust
bearings are of the fluid piston type. The drive motors are induction type,
suitable for 3-phase 60-cycle 4160-volt power supply. They have 1m-
pervious liners in the stator bore for hermetic sealing, and an outer housing
totally enclosing the stator as a sccond safeguard against loss of system
fluid. The motor stator windings are cooled by a liquid passing through
tubular conductors installed in the stator. All material in contact with the
9-5] TWO-REGION BREEDERS 499

core solution and DO is stainless stecl, except for the impeller, labyrinth
inserts, impeller nut, and wear rings, which are titanium.

By unbolting the top flange, the entire pump mechanism can be re-
moved, leaving only the high-pressure pump casing in the pipeline, thus
facilitating maintenance.

Shielding and containment. The reactor plant is housed in a 175-t-
diameter stecl sphere. Design pressure is 40 psia, which requires a nominal
plate thickness of 3/4 in. The sphere is buried to a depth of 50 ft, allowing
the reactor vessel to be located below grade for natural ground shielding.

Radioactive components are enclosed by a barytes concrete structure
which serves both as a biological and as a blast shield. The top of this
housing is 35 ft above grade elevation. The side walls are 5 ft thick, and
the top shield is 6 {t thick except for an 8-ft-thick section directly over the
reactor vessel. Compartment walls are provided within the housing to
facilitate flooding of individual component sections. The floor and side
walls of each of the compartments are lined with 1/8-in. stainless steel
plate to permit decontamination. Stepped plugs are provided 1n the top
shield to permit access to the components. The portion of the shielding
around the reactor vessel, which is below grade, is 4 ft thick. A slight
negative pressure is maintained within the container by continuously dis-
charging a small quantity of air to a stack for dispersal. The quantity of
air removed is regulated to control the ambient temperature in the com-
ponent compartments.

Cost analysts

This study indicates that a generating station with a net thermal effi-
ciency of 28.197, might be constructed for approximately $240.00/kw and
$200.00,'kw at the 180-Mw and 1080-Mw electrical levels, respectively.
These values result in capital expenses of approximately 4.2 and 3.86
mills/kwh.

9-5.2 Single-fluid two-region aqueous homogeneous reactor power
plant. The feasibility of a 150,000-kw (electrical) aqueous homogeneous
nuclear power plant has been investigated by a joint study team of the
Nuelear Power Group and The Babcock & Wilcox Company [19]. In this
concept, the reactor is a single-fluid two-region design in which the fuel
solution circulates through the thoria pellet blanket as the coolant. Com-
ponents and plant arrangement have been designed to provide maximum
overhead accessibility for maintenance. All components in contact with
reactor fuel at high pressure are themselves enclosed in close-fitting high-
pressure containment envelopes.

General description and operation of plant. The reactor generates 620-psia
steam at the rate of 2.13 X 10° Ib/hr.
500 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHap. 9

The reactor system is contained in a building 196 ft long, 131 ft wide,
and 50 ft high. The equipment is located in a group of gastight cells meas-
uring 196 ft X 131 ft over-all. These cells are equipped with pressure-tight
concrete lids to facilitate overhead maintenance of the various system
components and minimize the radiation shielding required above the floor
and outside of the building. All components in the reactor plant have been
designed in accordance with this overhead maintenance philosophy.

The basic systems comprising the reactor plant are: (1) A primary sys-
tem, containing the reactor, the main coolant loops, the boiler heat ex-
changers, the pressurizer, the surge tank, and the standby cooler; (2) the
letdown system; (3) the fuel handling and storage system; (4) the off-gas
system; and (5) the auxiliary systems containing leak-detection and fuel-
sampling facilities.

The blanket consists of 14 eylindrical assemblies arranged around the
periphery of the corc region. These assemblies contain thorium-oxide
pellet beds which are cooled hy fuel flowing from a ring header below the
reactor vessel. The fuel follows a zigzag path through the pellets and
leaves the assemblies through top outlets, flows through the core region,
and out the bottom of the vessel. By this means, the usual core-tank cor-
rosion and replacement problems and slurry handling problems are mini-
mized. By means of devices located at the tops of the tubes extending out
of the reactor the assemblies are periodically rotated to minimize absorp-
tion of neutrons by protactinium and equalize the buildup of U2?% in the
thorium. Replacement of the assemblies is possible through a smaller clo-
sure than would be required {or a two-region reactor with a single core tank.

The reactor vessel is surrounded by a high-pressure containment vessel
which forms part of the containment system described below. Over-all
height of the reactor 1s 28 ft 6 in. and the outer diameter of the containment
shell is 12 ft 1§ in.

The boiler heat exchangers are designed so that by removing the head
there is direct access for plugging tubes or for removing entire tube bundles
if necessary. These exchangers are a once-through type designed to evapo-
rate 959, of the feedwater flow at full load. The reactor fuel flows counter-
currently through the shell side of the exchangers. I'eedwater enters the
baffled heads and passes through the U-tubes where the steam is generated.
The steam-water mixture then leaves the exchangers and flows through a
cyclone separator and scrubber to the turbine. These components, with
their containment, are 34 ft 24 in. high and 4 ft 34 in. OD.

All piping and components holding high-pressure reactor fuel are con-
tained in a close-fitting pressurized envelope capable of withstanding the
total system pressure. These components and piping are further contained
in pressure-tight concrete cells which are vented through rupture disks to a
low-pressure gas holder, as shown in Fig. 9-11. This holder has a liquid-
9-5] TWO-REGION BREEDERS 501

Low Pressure
Heat Exchanger Reactor Gas Holder

 

  
  
   
  

 

 

L ! ]
V 7+

—_

N

\

"
%
P

% ‘

T

T

)
1

 

T,

&

 

 

-

f

f

 

 

   
 

 

 

 

 

 

D 2

e
<

 

 

 

 

 

Relief Pipe

Fig. 9-11. Schematic illustration of containment system (courtesy of the Nu-
clear Power Group and the Babcock & Wilcox Co.).

sealed roof which moves up and down in a manner similar to the movement
of a conventional gas holder section. The low-pressure components do not
have a close-fitting high-pressure envelope but are contained in pressure-
tight cells and are vented to the low-pressure gas holders in a manner similar
to the high-pressure components.

This type of containment permits operation of components for their full
service life, reduces or eliminates missile formation and fuel losses, reduces
primary system working stresses, and allows equipment arrangement
giving maximum access for maintenance.

Table 9-7 summarizes the characteristies of the proposed plant.

Reactor. The general characteristics of the reactor are illustrated by
Fig. 9-12, which shows the annular arrangement of the Zircaloy—2 blanket
assemblies around the core region. The thorium-oxide pellets within these
assemblies are cooled by the reactor fuel solution, which is pumped up
through the packed beds from the supply header. To reduce the pressure
drop across the pebble bed, the solution is introduced through a tapered
perforated pipe the same length as the assemblies, flows into the bed and
by means of baffles is directed back to the center outlet pipe, which is con-
centric with the inlet. _

The vessel has ellipsoidal heads, is 9 ft 11 in. ID, and has a cylindrical
shell length of 10 ft 6 in. The upper head contains a 3-ft 33-in.-diameter
502 LARGE-SCALL HOMOGENEOUS REACTOR STUDIES [cHAP. 9

TABLE 9-7

Dusiay Darta rorR THE SINGLE-FLUID Two-RicioNn REACTOR

 

Over-all plant performance
Thermal power developed in reactor, Mw 520
Gross electrical power, Mw 158
Net electrical power, Mw 150
Station efficiency, 97, 28 5
General reactor data
Fuel solution U0:804-D-0
Operating pressure, psia 1500
Fuel inlet temperature, °F 514
Fuel outlet temperature, °F h72
Area of stainless steel (in contact with fuel solution), ft2 90,000
Total volume of primary system, ft? 2,300
Area of Zircaloy-2 (in contact with fuel solution), ft? 2,240
Core
Fuel flow rate, 1b/hr 24 .9 x 106
Velocity, fps 6.3
Volume of core solution, liters 67,000
Letdown rate, gpm 100
Thorex cycle time, days 115
Hydroclone cyele time, days 1
Hydroclone underflow removal rate, liters/day 583
Blanket
Assembly diameter, 1n. 18
Fertile material ThO: pellets
Thorium loading, kg 17,850
Thorium irradiation eycle, days 744
Thorium processing rate, kg/day 23
Proceésing rate of mags-233 elements, g/day 350

 

 

 

 

flanged opening to permit removal of the thorium assemblies. The closure
is a double-gasketed bolted cover having a monitoring or buffer seal con-
nection to the annulus between the gaskets to detect leakage of the fuel
solution. The lower pressure vessel head is penetrated by fourteen 7-in.
openings through which fuel flows upward into the blanket assemblies from
the toroidal supply header. This head has a 3-ft-diameter fuet outlet. The
thermal shielding consists of alternate layers of stainless steel and fuel
solution. The total shielding thickness is 8 in. over the eylindrical portion
and 12 in. at the head ends of the reactor pressure vessel.

Boiler heat exchangers. The boiler heat exchanger is a U-tube, vertical,
9-3] TWO-REGION BREEDERS 503

T ||’|l| Double-Gasketed
' fl' Bolted Closures

     
     

Thorium
Oxide

Rotation Shaft
Pellets

1 >
| S Q ABIunt?t

Containment
Vessel

  
 
 

Aj Reactor
j Vessel

Core Region

 

Thermal
Shielding

‘ "1 Fuel Solution

g Supply
S Header

 

 

 

 

 

 

 

 

 

Fuel Solution

inlet
- 34——-—

Fie. 9-12. Single-fluid two-region aqueous reactor (courtesy of the Nuclear
Power Group and the Babcock & Wilcox Co.).

forced-circulation design in which reactor fuel flows on the shell side and
boiling light water on the tube side.

By placing the reactor fuel on the shell side, the tube sheet acts as its own
shield and is subjected to less intense nuclear radiation, minimizing gamma
heating and thermal-stress problems. Such a design permits the use of
thermal shields which would also serve to protect the tube sheet from
thermal shock due to sudden variations of fuel temperatures. In the event
of a failure of tubes or tube sheet connections it is necessary to remove only
the faulty tube bundle and leave the exchanger shell and flanged con-
nections intact. This is accomplished by removing the bolted head and
tube sheet brace and cutting the seal ring weld at the periphery of the
tube sheet. The bundle is then lifted out of the shell by the overhead crane.
d04 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP. O

Reactor building. 'The building which houses the reactor plant will be
airtight and will serve as a containment for radioactive gases which may
be released during maintenance. The building air will be monitored and
filtered and will be vented to the exhaust stack.

Space 18 provided outside the reactor plant for an off-gas building stack,
gas and vapor holders, gas handling building, hot laboratory and shops,
waste handling building, and chemical processing buildings. The hot shops
and chemical processing building are located as shown to permit mutual
access to a crane bay which extends from the reactor building between the
two buildings. Both "hot” components and blanket assemblies are trans-
ported from the reactor building to the far end of the bay by a low, U-frame
traveling crane. At the end of the bay they are transferred to an overhead
crane running perpendicular to the bay, and transported to either of the
two buildings. With this arrangement, hot materials may be transferred
entirely underwater, thereby eliminating the need for bulky shielding and
mobile cooling systems.

Maintenance considerations. A study of the problem of maintenance of
a large-scale homogeneous reactor indicated the following. It appears im-
possible to accomplish some repair operations remotely in place and under
20 ft of water. Experience to date tends to indicate that the repair of
radioactive equipment may be so difficult that it will be uneconomical to
repair anything except such small components as valves and pumps. The
larger defective components must be removed from the system and a re-
placement installed. The repairs, if possible, can then be made in a “‘hot
machine shop™ after the system is back in operation.

Removal of components from the cells will require shielding, such as
lead easks. Further study is necessary to determine the optimum means
of performing this operation.

Extensive use of jigs and fixtures in performing maintenance work will
be necessary for rapid and safe work. All the jigs and fixtures should be
designed and constructed before the plant is put into operation. In many
ases it will be advantageous to use the jigs during initial construction to
be certain that they will function properly.

The estimated annual maintenance cost for a plant of the size considered
is approximately $3,300,000, which includes the capital investment of
maintenance equipment. This amounts to about 4 mills/kwh at 60%
capacity factor or ~3 mills/kwh at 809 capacity factor. The 150,000-kw
nuclear power plant deseribed 18 estimated to cost $375.00/kw  or
$56,400,000, excluding $5,000,000 to $10,000,000 for research and de-
velopment.

0-5.3 Oak Ridge National Laboratory two-region reactor studies. In-
termediate-scale homogeneous reactor. In October 1952 design studies were
9-5] TWO-REGION BREEDERS 505

Chemical Plant Reactor Entrance Turbine Plant Transformer

. Containers Hcfch#
= F

l@" =
= vl

    

 
  
  
 

  

T1G. 9-13. Artist’s concept of Thorium Breeder Reactor Power Station.

started for a two-region multipurpose intermediate-scale homogeneous re-
actor as an alternate to the single-region reactors previously studied
[20-22]. Several suggested core vessel arrangements for two-region con-
verters were presented. In all designs the core shape approximates a
4-ft-diameter sphere, and a central thimble is incorporated to permit startup
and shutdown with a full core containing the operating concentration of
fuel. The major interest in the design of this type of reactor is the possi-
bility of converting thorium into U2#¥ in the blanket region of the reactor.

The principal system parameters on which the design of the two-region
intermediate-scale homogeneous reactor is based are presented in Table 9-8.

Large-scale conceptual designs. Design work on the two-region inter-
mediate-scale homogeneous reactor continued through the fall of 1953, with
emphasis being placed on design studies of components and reactor layouts
for an optimum design. In the meantime, conceptual designs of large-scale
two-region reactors described below were carried out as a basis of feasibility
studies.

The first design involved a 1350-Mw (heat) power plant containing three
reactors [17]. Each of these operated at 450 Mw to produce a net of
105 Mw of electricity. The design of this plant, which is reviewed in the
following paragraphs, is representative of the technology as of Septem-
ber 1953.
206

LARGE-SCALE HOMOGENEQUS REACTOR STUDIES

TABLE 9-8

[cHAP. 9

DEsIGN PAraMETERS oF A Two-REGION
INTERMEDIATE Scante HomoGENEoUs REAacToOr

 

 

 

 

Core system Blanket system

Power level, Mw 48 9.6

Fluid [U02504-D20 sol ThO-D30 slurry
Concentration, g U/liter 4.8 1000
System pressure, psia 1000 1000
System temperature, °C 250 250
Vessel diameter, ft 4 8
Maximum fluid velocity, fps 22.3 12.3
Pumping requirements, gpm 5000 1000
Steam pressure, psia 215 215
Steam temperature, °F 388 388
Steam generated, 1b/sec 38.7 8.0

 

 

 

 

In the proposed arrangement, three large cells are provided for the re-
actors and associated high-pressure equipment and a fourth is provided
for the dump tanks and low-pressure equipment. Iach reactor cell is di-
vided into compartments for the reactor, heat exchanger, pumps, and gas-
circulating systems. The low-pressure equipment cell contains compart-
ments for dump tanks, feed equipment, heat and fission-product removal,
D20 recovery, and the limited amount of chemical processing that can be
included in the reactor circulating system. Radiation from the cells is re-
duced to tolerable levels under operating conditions by concrete shielding.
The individual compartments have sufficient shielding to permit limited
access when equipment is being replaced.

Each reactor consists of a 6-ft-diameter spherical core, operated at a
power of 320 Mw (100 kw/liter), surrounded by a 2-ft-thick blanket which
1s operated at o power of 130 Mw (11 kw/liter). Under equilibrium con-
ditions, a solution containing 1.30 g of U233, U234 U235 17236 (45 uranyl
sulfate dissolved in D20) is circulated through the core at a rate of
30,000 gpm under a pressure of 1000 psia. Iluid enters the core at
213°C and leaves at 250°C. Decomposition of the D20 moderator by
fission fragments yields 240 efm of gas containing 28 mole 9 D>, 14 mole 9},
(2, and 58 mole 9 D0,

Liquid leaving the core divides into two parallel ecircuits, each at
15,000 gpm, which lead into centrifugal gas separators. There the ex-
plosive mixture of deuterium and oxygen is separated from the liquid and
9-5] TWO-REGION BREEDERS 07

diluted below the explosive limit with a recirculated gas stream which con-
tains oxyveen, helium, and D20, The gas-free liquid circulates through heat
exchangers and is returned to the core by canned-motor circulating pumps.
Steam is produced in the exchangers at 215 psig and 388°17.

The gas streams from the separators are joined and flow Into o high-
pressure storage tank accompanied by about 500 gpm of entrained liuid.
After the entrainment is removed In mist separators for return to the
liquid system, the Dy and Oz are recombined when the gus passes uto a
atalyst bed containing platinized alumina pellets. Ieat liberated in the
recombiner increases the temperature of the gas from 250 to 464°C. The
hot guses are cooled to 250°C in a gas condenser which has a capacity of
20 Mw and condenses D20 at a rate of about 89 gpm. Some of the D20
is used to wash the mist separators and to purge the pump bearings and
rotor cavity; the remainder is either returned to the system through the
high-pressure storage tank or held in condensate storage tanks during
periods when the concentration of reactor solution is being adjusted. The
oax= 1= recireulated to the gas separators by an oxygen blower.

Similar gas- and liquid-recireulating systems are used to remove heat
from the blanket, which consists of a thorium-oxide slurry m D20 contain-
ing 500 to 1000 g Th/liter. The slurry is recireulated by means of a
12.400—gpm canned-motor pump through a gas separator and through a
130-Aw heut exchanger.

The reactor is pressurized with a mixture of helium and oxygen which
i~ udmitted as required. It is expected that most of the fission-product
guses will be retained in the high-pressure gas-circulating systems with
only whatever small, daily letdown ig required to adjust the pressures.
Caleulations for @ similar system indicate that enough Xe!#5 will be trans-
ferred into the gas stream to reduce the xenon poisoning in the reactor by
a factor of 5 to 10.

The two-region thorium breeder reactor. A later design study was com-
pleted in the full of 1954 by the Reactor Lixperimental Engineering Divi-
sjon of the Ouk Ridge Nutionul Laboratory [23] for the purpose of de-
lineating the technical and economic problems which would determine the
ultimate feasibility of an aqueous homogeneous reactor for producing cen-
tral station power.

The concept of the reactor chosen for study was based essentially on
nuclear considerations and consists of a spherical two-region reactor with
dimensions limited by economic considerations.

Table 9-9 presents the principal reactor characteristics for the pre-
liminary design of a 300-Mw station.

The power plant complex, consisting of the reactor plant, the turbo-
generator plant, the chemical processing plant, the cooling system, and
part of the electrical distribution system, is showw n Fig. Y-13.
28

LARGE-SCALE HOMOGENEQUS REACTOR STUDILS

TABLE 9-9

[cHAP.

REscTor CHARACTERISTICS FOR PRELIMINARY
Desiegy or A 300-Mw STAaTION

Electrical capability (each of three reactors), 100 Mw
Gross station efficiency, 27.49,
Net station ethiciency, 26.09
Operating pressure, 2000 psia

 

Core

Blanket

 

 

 

Material of construetion

Wall thickness, in.
Thermal shield thickness, in.
Pipe connections
Inside diameter, ft
Blanket thickness, in.
Volume, liters
Operating temperature, °C
Average
Inlet
Outlet
System volume, liters
Fluid composition, g/liter D20O*
17233
U234
'U'235
U236
Thorium
Inventory, kg
Ds0O
U235 4 7233
Thorium
Flux at core wall, n/{cm?2)(sec)
Power density at core wull,
kw/liter
Mean power density in reactor,
kw/liter
Power density in external system,
kw/ liter
Reactor power, Mw (heat)
Circulation rate, gpm

 

Zircaloy-2
0.5

Concentric
5

1855

275
250
300
9740
U02504-D,0-CuS0y
1.88
1.90
0.26
3.00

10,900
2.1
1.10 x 1015
70.0
193
61

313
24,000

 

2097, stainless-steel
clad carbon steel
5.0
4.0
Straight-through
103
27
11,600

280
245
315
13,800
UO 3T}IO 2*]_) 20
3.00
0.17
0.01
0.00
1000

12,000

42.9
14,300

1 x 10

_‘-l
<

~1 Tt
o Dt

 

*At operating conditions.

 
9-5] TWO-REGION BREEDERS 509

The reactor plant consists of a space 80 ft () in. wide, 300 ft 0 in. long,
and 15 ft 6 m. above ground level. At one end of the structure 1s located a
storage pool for items {reshly removed from the shield. A gantry crane
services the reactors and its runway extends a distance heyond the shield
for access to the pool and to provide lay-down space for a reactor-shield
tank dome. Three evhindrical shield tanks ure provided to contain each
reactor and components.

In the event of a line rupture or equipment fuilure resulting in gross
leakage of reactor fluids from the reactor system, no radioactive material
will be released. This is accomplished by placing the reactor and com-
ponents in a ecylindrieal tank, 66 ft 0 m. diameter X 116 ft 0 in. high,
apable of withstanding 50 psig.  Light feet of conerete are poured around
this tank to a height of 45 {1 6 in. above ground level for biological shielding.

Ax shown m Ilig. 9-13, each of the container buildings has an access
hatch.  Items such as circulating pumps, pressurizer heater elements,
evaporators, ete,, for which the probability of maintenance i1s high, are
grouped on one side of the shield and more or less under this hatch for
servicing.  No specific procedure for repairing or replacing equipment was
developed; however, consideration was given to both wet and dry main-
tenance methods,

Homuogerneous reactor cxperiment No. 3.* In 1957, conceptual design studies
of HRI-3, a two-region homogeneous breeder reactor fueled with 17233
and thorium, were initiated at the Ouk Ridge National Laboratory [24].
This reactor, operating at 60 Mw of heat to produce approximately 19 Mw
of electrical energy, will be designed to provide operational and technical
data and to demonstrate the technical feasibility of an intermediate-scale
aqueous homogeneous power breeder. The power plant will be a completely
integrated fucility incorporating (1) the nuclear reactor complex, (2) the
electrical generating plant, and (3) the nuclear fuel recycle processing
plant. Prelimimary design criteria are given in Table 9-10.

As presently concelved, the reactor will bhe of the two-region type with
a heavy-water uranyl-sulfate solution being circulated through the inner
(core) region, where 50 Nw of heat are produced. A thorium-oxide slurry
will be circulated through the outer pressure-retaining (blanket) region,
where 10 Mw of heat are produced at equilibrium conditions. The fuel
solution and slurry will be circulated through separate steam generators by
the use of canned-motor pumps. The fuel heat exchanger will provide
195,700 1b/hr of saturated steam at 450 psia (456°F) at 50-Mw core power,
and the slurry heat exchanger will provide 39,100 1b/hr of saturated steam
at 450 psia (150°F) at 10-Mw blanket power. The blanket and core regions,
operating at 1500 psia and 275°C and 280°C average temperatures, re-

*By R. H. Chapman.
510 LARGE-SCALE HOMOGENLOUS REACTOR STUDIES [crap. 9

TaBLE 9-10
HRE-3 Drsiax CRITERIA

 

 

 

Type Two-region breeder

Core U02804 4 CuSO4+ DeSO4 in DO

Blanket ThO2 + U0z 4+ MoOgz in D20

Core eritical concentration at 50 Mw and | 4.8 g U233/kg D50
equilibrium 045 g U /kg DO

Blanket concentration at 10 Mw and | 1000 g Th/kg D0
equilibrium 4.02 g U233 /kg D10

Average core temperature, °C 280

Average blanket temperature, °C 275

Average core power density for 50 Mw,
kw /liter 52.6

Average blanket power density for 10 Mw,
kw /liter 1.04

Listimated breeding ratio 1.05 (minimum)

Saturated steam pressure, psia 450

Gross clectrical power, Mw i ~19

 

 

spectively, are interconnected in the vapor region. Inasmuch as oxygen
is consumed by mechanisms of corrosion and must be added continuously,
1t 1s currently favored as the pressurizing medium to provide the over-
pressure necessary to prevent boiling and bubble formation. Sufficient
homogeneous catalysts will be provided in the solution and slurry to re-
combine all the radiolytic gases formed in the cireulating system during
operation. In this manner it will be unnecessary to operate with continuous
letdown of slurry and/or solution. Purge water for use in the high-pressure
eirculating systems will be produced by condensing a portion of the steam
contained 1n the vapor volume of the pressurizer. Advantage is taken of
the beta and gamma decay energy to maintain the pressurizers at a slightly
higher temperature than the remaining portion of the system.

The electrical generating plant will be essentially of conventional
design. The 20-Mw turbine will operate at 1800 rpm on 450 psia saturated
steam with moisture separation equipment provided. The condensing water
recquirements are 31,300 gpm, assuming the water enters at 70°F and leaves
at 80°F. The generating voltage of 13.8 kv is sufficient to permit direct
connection to an existing distribution system.

The fuel reprocessing for HRE-3 will consist of concentrating insoluble
fisston and corrosion products in the underflow pots of hydroclone sepa-
rators, recovery of uranium from the hydroclone underflow by TUO, pre-
cipitation, and recovery of D20 by evaporation. The slurry processing
operation will consist of D20 recovery by evaporation and packaging the
9 5] TWO-REGION BREEDERS il

irradiated ThO, for shipment to the existing Oak Ridge National Labo-
ratory Thorex Pilot Plant, where the uranium and thorium will be scpa-
rated and reclaimed. The thorium will appear from the Thorex process as
a thorium nitrate solution and will be converted to ThO., before being
returned to the blanket of the reactor.

The maintenance philosophy of the reactor complex has not vet been
established, However, it can be said that the reactor system will be designed
<0 that all components and equipment will be capable of being removed
and replaced, but with varying degrees of difficulty. The choice of under-
water, dry, or combination thercof, maintenance techniques has not been
made. The major components of HRE-3 are considered to be in the range
of sizes which might be used in a large-scale Thorium Breeder Power Plant.
Design, development, fabrication, and operational and reliability data are
expected to be gained from HRE-3, in addition to maintenance techniques
for large-scale aqueous homogencous reactors. A very preliminary cost
study indicates a cost of about $29,000,000 for the reactor complex, the
clectrical generating plant, and the fuel reprocessing plant.
512 LARGE-SCALE HOMOGENEOUS REACTOR STUDIES [cHAP, 9

REFERENCES

1. Tueopore RockweLL LI, Reactor Shielding Destgn Manual, 1st ed. New
York: D. Van Nostrand Company, Inc., 1956.

2. W. F. Tavyror, TBR Plant Turbogenerator System Study, USAEC Report
CF-56-7-127, Oak Ridge National Laboratory, June 1956.

3. FosTtEr-WuErLER CorroraTiON, Wolverine Electric Cooperative Proposal,
Feb. 1, 1956, Report FW-56-004; Sargent and Lundy Fstimate No. 3800-2,
November 1957.

4. M. I. Luxpin and R. Vany WinkLe, Conceptual Design and Evaluation
Study of 10,000 KW IE Aqueous Homogeneous Nuclear Power Plant, USAEC
Report CF-57-12-8, Oak Ridge National Laboratory, Dec. 11, 1957,

5. P. R. Kasten ¢t al., Aqueous Homogeneous Research Reactor—Feastbility
Study, USAEC Report ORNL-2256, Oak Ridge National Laboratory, Apr. 10,
1957.

6. FForp Motor CompraNy, A Selection Study for an Advanced Engineering Test
Reactor, Document No. U-047, Aeronutronic Systems, Inc., Glendale, Calif.,
Mar. 29, 1957.

7. W. E. Joanson et al., The P.AR. Homogeneous Reactor Project, Mech.
Eng. 79, 242-245 (1957).

8. WestTiNgHoUuskt ELectric CorrPoraTioN, 1956-1957. Unpublished.

9. J. A. Lane et al., Oak Ridge National Laboratory, 1950. Unpublished,

10. J. A. Laxrk et al., Oak Ridge National Laboratory, 1951, Unpublished.

11. R. H. BawL et al.,, Oak Ridge National Laboratory, 1952. Unpublished.

12, J. J. Karz et al., Argonne National Labhoratory, 1952, Unpublished.

13. L. E. LiNk et al., Argonne National Laboratory, 1952, Unpublished.

14. H. A. Onrerex and 1) J. Mavron, Idaho Operations Office, 1952, Un-
published.

15. CommonwraLTH EpisoN Company anp PuBLic ServicE COMPANY OF
NortaeERN Ivuizors, .U Report on the [easibility of Power (eneration Using
Nuelear finergy, 1952, {Unpublished.

16. W. E. Tuomrson (Comp.), Homogeneous Reactor Project Quarterly Progress
Report for the Period Ending Mar. 15, 1952, USAEC Report ORNL-1280, Oak
Ridge National Laboratory, 1352.

17. R. B. Brices et al., Aqgueous Homogeneous Reactors for Producing Central-
station Power, USAEC Report ORNL-1642(Del)), Oak Ridge National Labora-
tory, 1954,

18. H. G. Carsox and L. H. Laxpruym (Eds.), Preliminary Design and Cost
Estimate for the Production of Central-station Power from an Aqueous Homo-
geneous Reactor Utilizing Thoriuwm-Uranium-233, USALEC Report NPG-112,
Commonwealth Edison Company (Nuclear Power Group), Feb. 1, 1935.

19. CommonweaLTy EbpisoNn Cosmeany, Single-fluid Thwo-region  Aqueous
Homogeneous Reactor Power Plant: Conceptual Design and Feasibility Study,
USAEC Report NPG-171. July 1957.

20. W. I. Trowrsox (Comp.), ITomogeneous Reactor Project Quarterly Progress
Report for the Period Ending July 1, 1552, USATLC Report ORNL-1318, Oak
Ridge National Laboratory, 1952.
REFERENCES 513

21. W. E. Taompson (Comp.), Homogeneous Reactor Project Quarterly Progress
Report for the Period Ending Oct. 1, 1952, USAEC Report ORNI-1424(Del.),
Oak Ridge National Laboratory, 1953.

22, W. E. Tnompson (Comp.), Homogeneous Reactor Project Quarterly Progress
Report for the Pertod Fnding Jan. 1, 1953, USAEC Report ORNL-1478(Del.),
Oak Ridge National Laboratory, 1953.

23. Oak Ridge National Laboratory, 1954. Unpublished.

24. J. C. Bougrer et al., Preliminary H RI-3 Design Data ( Revised to 11-15-57),
USAEC Report CF-57-11-74, Oak Ridge National Laboratory, Nov. 29, 1957.