JPRS ID: 10710 USSR REPORT ENGINEERING AND EQUIPMENT

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• Content Type: REPORTS

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JPRS,L/.10710
4 August1982
USSR Report
ENGINEERING AND EQUIPMENT
C F.O U O. 8[8 2)'FgI-S FOREIGN BROADCAST INFORMATION SERVICE
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JPRS L/1G?10
4 August 1982
USSR REPORI'
ENGINEERING AND EGIUIPMENT
(FOUO 5/8i;
CONTENTS
AERONAU'I'ICAL AND SPACE
Gas Generators of Rocket Systems 1
NUCLEAR ENERGY
Current State and Outlook for HTGR Research in USSR.............
3
Some Requirements for Nuclear-Chemical Facilities With
High-Temperature Reactors.....................................
12
Particulars of Layout and Constructian of Experimental
Industrial High-Temperature Gas-Cooled Reactor ModQl..........
18
Choosing Design Concept and Physical Features of HTGR Core
for Energy Facilities.........................................
22
Molten-Salt Reactor With Natural Convection of Fuel Mixture
and Open Gas-Turbine Air Cycle
28
Physical Features of HTGR With Circulating Fuel
36
Some Problems of Heat Exchange and Hydrodynamics in HTGR Core
Components (Survey)
43
Some Results of Experimental Research on HTGR Equipment
Components
50
Fabrication and Quality Control of Coated Fuel Particles,
Fuel Elements and Fuel Assemblies for HTGR's..................
61
Transposing Fuel Assemblies To Equalize Energy Distribution
and Improve Fuel Cycle in RBMK Reactors.......................
84
- a- [III - USSR - 21F S&T FOUO]
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NON-NUCLEAR ENERGY
Energy-Storing
Substances
and Their
Utilization.................
92
Reliability of
Electrical
Machinery
for Power Generation........
94
NAVIGATION AND GUIDANCE SYSTEMS
Navigation and Controlling Movement~ of Mechanical Syetems....... 102
~
Control System for Elastic Moving Objects 10
HIGH-ENERGY DEVICES, OPTICS AND PHOTOGRAPHY
Optical Devices for Measuring Surface Roughness 114
FLUID MECHANICS
Hydrodynamic Theory of Lubrication and Analysis of Plain
Bearings Operating Under Stationary Conditions.... 116
120
Increasing Heat Exchange Efficiency in Power Equipment..........
Applied Problems'in Hydromechanics.....94 ���4�9�009 129
riECHANICS OF SOLIDS
Oscillations of Kinematically-Driven Mechanical Systems
13E
Considering Energy Dissipation
Two-Dimensional Vibration Impact Systems: Dynamics and
Stability 138
- b -
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AERONAUTICAL AND SPACE
UDC 629.7.064.2
GAS GIIdgtATORS OF ROCKET SYSTENS
Moscow GAZOGIIJERATORY RAKETNYKH 5ISTEM in Russlan 1981 (signed to press 14 Aug 81)
PP 2-4
[Annotation and foreword from book "Gas Generators of Rocket Systems", by
A1'bert Alekseyevich Shishkov and Boris Vasil'yevich Rumyantsev, Izdatel'stvo
"Mas}iinostroyeniye", 1183 copies, 152 pages]
[Textl Annotation
This book gives a systematized description of the basic arrangements,chaxact eris-
t ics and spec3al features of the operating procasses in gas generators ubing
chemical fuels (liquid, solid and mixed) for use as power sources and gas jets
aboard aircraft and in ground systems of rocket equipment. Methods of experimental
finishing off of gas generaturs are briefly considered. This book is intended for engineers and designars in the area of rocket technology.
Foreword
Gas generators are widely used in rocket equipment. Their main units axe very
similax to the ma,in units of basic rocket engines; however, the operating processes
in gas generators have essential special features which must be taken into account
in designing and finishing them off.
Numerous patents and ma.gazine articles have appeaxed in recent yea,rs in connec-
tion with rocket equipment, and the expanded use of gas generators, which resulted
in investigations of gas generator devices [2]. Brief information on gas genera-
tors is available in manuals on the bases for designing rocket engines [2,
However, as a whole, published materials on gas generators are disconnected, frag-
mentary and methodologically inhomogeneous.
In this book the authors attempted to systematize the description of the axrange-
ments and special features of the operating processes in gas generators using
different fuels, based on the basic principlea of rocket engine theory.
The book contains five chapters. Chapter one describes the basic chaxacteristics
of gas generators and the fuel compositions used, and considers sepaxately the
methods for laboratory and test stand teata.
1
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Chapter two describes the special feattxes of gas flows in the gas generator,
the gas conduit and exhaust nozzles, as well as methods for calculating the gas
dynamic characteristics of gas generatora. Questions of filtering gas generation
products and of gas thermodynamic processes in the devices axe elucida.ted.
ohapter three reviews the special features of the devices on the basis of calcu-
lating one and two-component liquid gas generators, as well as gas generators
using fluidized (powdered) fuel.
Chapter four considers design arrangements, methods*for internal ba.llistic cal-
culations and vaxious possible methods for ragulating hard fuel gas generators
(especially, by front combustion chaxges), including multiplo connection gas
generators. The problem of transition processes during the change of decisive
parameters is solved.
The last chapter describes questions of developing vaxious combination gas genera-
tors using solid (with sepaxate components), quasihybrid and hydxid fuels in steam-
gas generators and gas generators of direct-flow rocket and rocket-tuxbine engines;
engineering methods axe given for calculating the bas;c caaracteristics of a num-
ber of gas generators.
Chapters one, four and i'ive were written together; chapter txo by A. A. Shishkov
and chapter three by B. V. Rumyantsev.
The authors express their dQep gratitude to A. P. Tishin for his valua,ble recom-
- mendations and for facilitE.ting the improvement of the manuscript in all its com-
ponents; to candidate of techrlical sciences M. Ye. Yevgen'yev for useful advice
in snlving problems in several sections. They will be grateful to readers who find
it possible to send their comments to IzdateZ'stvo "Mashinostroyeniye" to addresso
107076, Moscow, Stromynskiy per. 4.
COPYRIGHT: Izdatel'stvo "Mashinostroyeniye", 1981
2291
CSO: 1861/197
2
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NUCLEAR ENERGY
UDC 621.039
CURRENT STATE AND OUTLdOK FOR HTGR RESEARCH IN USSR
Moscow ATOMNO-VODORODNAYA ENERGETIKA I TERfIIdOLOGIYA in Russian No 2, 1979
(signed to press 8 Jun 79) pp 57-66
[Report TC-10913 at meeting of Technical Cammittee on HTGR's. IAEA, Vienna,
12-14 Dec 1977]
[Text] An examination is made of the ma3or advantages of
high-temperature gas-cooled thermal and fast reactors, along
with the feasibility of using them for electric power pro-
duction and generation of high-potential thermal energy.
A survey is given of theoretical and experimental research
on such reactors in the USSR.
Initial Assumptions
, Electric power production accounts for about 20% of the worldwide consumption
; of energy resources, while 80% of energy resources (petroleum, gas and coal)
are expended for industrial and household heating purposes, transportation,
in the chemical, metallurgical and otrer areas of industry.
Among world reserves of fossil fuel, only coal is far from being exhausted
(according to estinates, only 2-3X of the reserves will be used up by the
end of the century). However, transportation problems and the high cost of
electric power plants that use coal considerably reduce the competitiveness
of coal as compared with petroleum and gas, which has led in recent decades
to preferential use of these valuable chemical pxoducts for energy purposes,
to their increased cost, and in future will lead to earlier depletion of their
reserves as compared with coal. From this standpoint, nuclear power must
cover the needs of electric energy production, and at the same time be used
f.or producing process heat.
An important factor favoring development of nuclear power is the ecological
situation. Environmental pollution may become a serious limitation on the
road to further expenditure of fossil fuel, and especially coal, for power
production.
Naturally, nuclear power must undergo technical changes far successful intro-
duction in new fi.elds. District heating and production of low-potential heat
3
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can be successfully handled by light-water reactors. On the other hand, metal-
lurgy, the chemical industry and hydrogen production require the developm en t
of reactors with temperature level of 800-1000�C and higher. At the present
stage, high-temperature heliwn-cooled reactors (HTGR's) can be considered
the most efficient sources for combined production of electrical and high-
potential thermal energy.
Giith extensive development of nuclear power, scales may be limited by nuclear
fuel resources. The world reserves of inexpensive uranium commensurate in
respect to energy resources with petroleum reserves will have already been
exhausted by the beginning of the next century. The fuel problem can be solved
by breeder reactors that by expanded conversion can extend the capabilities
of uranium by dozens of times, putting into the cycle even the uranium dissolved
in sea water. Considering the actual characteristics of energy consumption
(i. e. the variable loading schedules, thE necessity of producing high-potential
heat and so forth), it is neceseary to set up an economically feasible two-
component nuclear power structure including thermal reactors (light-water
reactors and high-temperature plutonitun and thoriinn reactors) and breeders
with a short doubling time (4-6 years). Such a doubling time is easier to
achieve by fast helium reactors that have good physical and technological
characteristics. Helium breeders will be able to accumulate secondary plutonium
for their own development, as well as nuclear fuel for thermal reactors which
may comprise up to SOX in the nuclear power system. In this case, when involve-
ment of thorium is considered, an outlook is opened up for development of
nuclear power on a truly enormous scale (Fig. 1-3).
N
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1 0 100 150 100 150
Power production
kWyr/hd
Years
Fig. l. Scales of development of
_ power production (zones of undeter-
mined use of energy resources are
shaded)
Fig. 2. Doubling time of nuclear
power capabilities
4
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Fig. 3. Uranium consumption and necessary numbers of breeders
Peculiarities and Advantages of HTGR's
HTGR's have some distinguishing features and advantages that make them the most
- promising for the nuclear power industry. Principal among these are:
1) high temperature, better thermal eff iciency, lower heat emissions to the
environment, lower consumption of cooling water;
2) high safety due to high negative reactivity, high heat capacity of the
graphite core, absence of phase transitions and chemica'L inertness of the
coolant, and the presence of a number of safety barriers (microsphere fuel
elements - el.ement cladding - prestressed concrete vessel - emergency enclosure
of the nuclear electric plant);
3) efficient fuel cycle (including Th and Pu) due to excellent neutron-physics
characteristics, high accumulation factor and breeding ratio of fuel;
4) reliability in operation, simplicity of servicing, lower specific activity
of the loop and leakage of radioactivity to the environment;
5) capability of high-power instdllations with lower capital investment, use
of a gas-turbine cycle, "dry" cooling towers, simultaneous production of elec-
tric power and high-potential thermal energy, use of nuclear electric plant to
cover peak loads.
HTGR Research Areas
Experimental facility with VGR-50 reactor of 50 MWe power.
Purpose: accumulation of experience in designing and building HTGTc's, working
out hetium technology, dynamics, safety, mass testing of fuel eler!2nts, test-
ing components of reactor control system, equipment components, etc.
The engineerir.g plan for the project has been worked up.
5
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2. Prototype facilities:
with VG-400 thermal reactor of 1000 MWt power for process energy purposes;
with BGR-300 fast reactor of 300 MWe power.
Purpose: accumulation of experience in making installations with HTGR's and
FGR's, developing equipment (gas blowers, sceam generators, heat exchangers)
studying problems of making prestressed concrete vessels, industrial methods
of using high-potential thermal energy, confirmation of feasibility of attain-
ing the required conversion properties in breeders.
Design work has been done on the rough-draft and planning stage. Coordinated
work is bPing done by scientific research and planning design agencies.
3. Industrial facilities with HTGR's and FGR's of more than three
million kWt po w er. Purpose: production of high-potential heat, regener-
ators, synthetic fuel, hydrogen for industry, transportation and household
use; for purposes of electric power production with the use of a direct cycle,
air cooling and utilization for covering peak loads.
Studies are being done on parameters and prospects for using large HTGR's
for the national economy.
State of Research on HTGR's
Theoretical and experimental research. 1. Studies are being done
on analyzing areas for most efficient use of HTGR's:
a) for electric power production (here other types of reactors may be com-
petitors), including with gas-turbine facilities and for covering peak loads;
b) to produce high-potential process heat (this field of application of HTGR's
is most promising, including in connection with the lack of competition at
the present time). Under consideration are the processes and sectors of in-
dustry that consume t}ie most 'energy, and also utilization for transportation
and domestic purposes:
metallurgy, where the use of HTGR's will reduce demands for coke and natural
gas, and when the technol.ogy has been successfully developed will enable tran-
sition to the process of direct reduction of iron;
chemical industry (production of ammonia, methanol, ete.);
gasification of coal;
}ieat supply to centralized consumers.
Partictilar attention is being given to hydrogen production by dissociating
water, since water is an unlimited source of the most ideal energy carrier,
which can be used in the power industry, metallurgy, chemistry, households,
transportation, etc.
6
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2. Physicotechnical studies are being done on different HTGR designs (physics
and thermophysics of reactors, dynamics, fuel cycles, various kinds of fuel
elements, safety, etc.).
3. Materials for HTGR's are being studied. Research is in progress on dif-
ferent kinds of graphite for fuel elements azd reflector, construction materi-
als for equipment (tubing, steam generators, heat exchangers, etc.), insulation
materials.
4. Work is under way on fuel element manufacturing technology (spherical
and prismatic versions) and making microsphere fuel elements.
5. Research is being done on equipment components in facilities and nuclear
- electric plants (prestressed concrete vessels, steam generators, heat exchan-
gers, etc.).
6. Research is in progress on helium coolant technology, yield of fission
products, monitoring instruments and helium cleaning system.
7. Facilities are being developed and made for studying patterns of movement
of fuel elements, rods of the control system, materials and the like, critical
stands (Astra, Grog) (Fig. 4), experimental helium loops (PG-100) (Fig. 5).
R.eactor ampule tests are being done on fuel elements and microspheres.
to sampling
system
6,1000 m 3
-0.03 kg/ M2 220'r.
20' 25'C r'-
~
B00

e
?00 C
M
~
t
~ 3800
) yo,60'c ~
=10 kW
A 0025~ r
kg/cm
4. Astra critical stand
8. Research is being done on heliun-cooled fast reactors that have an advan-
tage over other types of fast breeders in higher breeding gain and shorter
doubling time; e-cperiments are being done on critical assemblies (Korba).
P;anning and design work. 1. Development is in progress (engineering
design stage) on a two-loop experimental chemical process facility with HTGR
~ (Fig. 6).
7
'2BO�c
?90'C j
11 "-to evacuatio:i
=1.5k / 2 tem
pp- o makeup
220�C 25'c system
~ +
120'C `
T
40=
60'C
N=5 kW
n=9000 rpm
p=0.O1
~~kg/cm~
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Emer
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Channel
FOR' OFFICIAL USE ONLY
Intermediate
Fig. 5. Schematic diagram of PG-100
gas loop with channel for testing
spherical fuel elements
*expansion of 3IIIC not given
Fig. 6. Diagram of nuclear chemical processing facility:
1--irradiator; 2--reactor; 3--steam generator; 4--turbo-
generator
Power--50 M"We
Heli.um temperature--280/800�C
Pressure--40 kg/cm
Core dimensions--D/H = 2.8/4 m
Control rods:
in reflector and pylons--24
submerged--4-6
Fuel element--sphere = 60 mm)
Number of fuel elements:
in facility--260,000
in reactor--125,000
Fuel enrichment--21%
Burnup--100,000 MW�day/metric ton
Gamma power of radiation loop--300 kW
8
.
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Ne = 300 MW; tin= 300'C; T2 = 6.5(8.0) yr; t~iX= 800�C; Tcn= 0.5 (1.0) yr
max
tfuel � 2150%
' Fuel U02 + Pu02
p_: 20o kg/cm 1)..-180kg/cm
li, I(7,7; yr - 6.3(7,6) yr
N� - 25 MW
np
eHT3 = 0,7
5.9 (7,3) yr
t
`'Nf'iK-' M\I ~KOi1f-O
aKO;u=2 Mn~ 6,0 (7.4) yr 5.2 (6,3) yr
(
_ AH,.3-0,7 M AH ,-0.5
6 4 (7-7)yr I T .
Pu compo- bK~m=~
sition RgM (WER+RBMK) ' 5.6 (6.8) yr r 4,5 5MG yr
~ I I I
~
eHT,-o,7 M
eRr,3=o 7 M t_ AR6,-.0,4 N,
6, 1 (7.4) yr
~
Ne = 300 MW; dKOX = 0; T2 = 3.5(4.3) yr; AHT3 = 0.7 m;
Tnep= 0.5(1.0) yr; AR= 0.6 m; p= 200 kg/cm2; tin= 300�C
tout = 620�C; t~iX = 800�C; Nnp3 = 25 MW; tfuel m 2150�C;
Fuel U02 + Pu02
Fig. 7. Influence of BGR-300 parameters on fuel doubling time
Steam parameters--90 atm, 535�C Reactor shell--steel
Number of cooling loops--4 Proposed completion deadline--1985
9
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- p. - Ir,o kg/cm
t 620" C
out =f50' C F.4 (7,8) yr
Na,r -2> MW A~~'p' --:IS MW
6,1 (7.61 yr Ana.14 =3.81C$/Cm
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0
0
M
Fig. 8. Fast breeder reactor with helium cooling (integrated
version): 1--passage for extracting and installing cassettes;
2--passage for reloading mechanism; 3--gate; 4--prestressed
concrete vessel; 5--reloading mechanism; 6--steam generator;
7--breeding blanket; 8--core; 9--gas blawer; 10--control
system drive; 11--rotating device; 12--truck with telescopic .
hcisting crane
2. Development is in progress (rough-draft stage) on prototype facilities
(different versiotis) for producing process heat (up to 950�C) with power of
500-1000 MWt with HTGR using spherical and block fuel elements, including
the VG-400 technological power unit:
- 10
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M-_rmal power of reactor--1000-1100 MW
?lelium te:nperature--350/950�C
Helium pressure--50 kg/cm2
Core dimensions--D;H = 6.4/4/8 m
Control rods in reflector and core
Fuel element--sphere (0= 60 mm)
Number of fuel elements--800,000
Fuel enrici=ent--10%
Run--3-4 yr
Nunber of loops--4
Electric energy production--300-400 MW
Hydrogen production (thermochemistry,
methane conversion)--(20-25) 103
nm3/hr per loop
Vessel--prestressed concrete
Proposed deadline--1985-1990
3. Development is in progress (rough-draft stage) on a demonstration fast
helium reactor with power of 300 MWe (Fig. 7, 8):
Thermal power--800 MW
Helium temperature at reactor outlet--
600-850�C
Helium pressure--160 kg/cm2
Energy release rate--500 kW/liter
Fuel--U02-Pu02
Breeding ratio--1.6-1.7
Doubling time--6-8 yr
Second loop parameters--170 atm, 540�C
Vessel--prestressed concrete
Nuclear electric plant efficiency--38%
Deadline undetermined
4. Studies are being done on parameters of facilities proposed for industrial
introduction after 1990.
COPYRIGHT: Institut atomnoy energii im. I. V. Kurchatova, 1977
Atomizdat, 1979
6610
CSO: 8144/1052-A
11
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UDC 621.039
SOME REQUIREMENTS FOR NUCLEAR-CHEMICAL FACILITIES WITH HIGH-TEMPERATURE
REACTORS
Moscow ATOMNO-VODORODNAYA ENERGETIKA I TEKHNOLOGIYA in Russian No 2, 1979
(signed to press 8 Jun 79) pp 67-72
7
[Article by N. D. Zaichko, I. Ya. Yemel'yanov, A. M. Alekseyev, V. M.
Panchenkov, Yu. I. Koryakin, A. A. Orlov, E. K. Nazarov, V. A Chernyayev,
S. A. Mikhaylova, L. P. Dudakov and S. V. Radchenko]
[Text] The authors consider conditions of forming power-
process arrangements for prodi;r.ing hydrogen, amonia and
other goods based on direct conversion of heat from high-
temperature nuclear reactors to a technological process.
Based on these conditions, major requirements are formulated
for high-temperature reactors to serve industrial technology:
service life, time between repairs, radiation safety, etc.
Ref. 1 reported on the feasibility and major areas of introducing high-tempera-
ture reactors for making hydrogen, ammonia and other products. It was shown
that the process most ready for realization in respect to.receiving thermal
energy from high-temperature reactors is steam catalytic conversion of hydro-
carbons. Therefore let us consider some requirements for nuclear-technological
facilities that realize conversion processes.
As fossil fuel is displaced by nuclear fuel, there is a considerable reduction
of labor inputs, an increase in labor pr.oductivity, and a reduction in produc-
tion outlays and settling expenditures per unit of final product due to a
reduction of labor inputs on extracting and transporting fuel because of the
_ much higher "calorific value" of nuclear fuel over any kind of fossil fuel.
Nonetheless, along with the overall reduction in the level of labor expendi-
tures in the national economy, there may be some increase i.n the nitrogen
industry. Therefore careful analysis and detailed examination of this point
must be taken into consideration when studying the feasibility of introducing
nuclear reactors into industrial technology.
When high-temperature heat is produced by burning natural gas, an arrangement
with better technical-economic efficiency is two-stage endothermic conversion
of inethane, which is mainly the basis for ammonia production (Fig. 1).
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1 2
~
i ~ '
- SC
e
- ~
~u
-a-rt------e-~-a-,
--_-t-_~--~-~_ I
c t~ean ing ~

---j I
I
Y
3 etha onv onv
~atiMEA -1'~
. ,
Fig. 1. Classical ammonia production scheme: 1--natural gas
turbocompressor; 2--air turbocompressor; 3--nitrogen-air mix-
ture turbocompressor; 0 process gases;
air; natural gas (fuel); o-- feed water;
water vapor
Engineering, design and economic workups of plans for the first nuclear-chemical
complexes are based on ammonia and methanol production aggregates with capacity
of 2500-3000 metric tons per day. The choice uf this ammonia and methanol pro-
duction capacity has dictated a required power of the nuclear reactor instal-
lation of 550-600 MWt.
With r.espect to conditions of direct utilization of high-temperature heat
= from nuclear reactors in amanonia production, an arrangement with two-stage
methane conversion is most preferable, although it is possible to realize
a scheme with single-stage conversion of inethane and additional displacement
of fossil fuel even in the process channel. The advantages of the two-stage
. arrangement are as follows: maximiun capabilitias for replacing fuel gas with nuclear fuel;
minimum level of working temper.ature of the stage of the technological process
in whi,h heat from the liigh-temperature nuclear reactor is to be used--less
than 875�C;
the natural gas consumed in this arrangement is divided into two flows: fuel
~ (45-50%) burned to get high-temperature heat, and process gas, facilitating
cc>nditions of replacing the fuel natural gas with nuclear fuel;
steam catalytic endothermic conversion of inethane is accomplished in individual
tubular reactors with diameter of less than 150 mm with external supply of
high-temperature heat through a solid wall (rather than in an integrated work-
ing volume), which is more favorable for direct utilization of heat from high-
temperature nuclear reactors;
the tubular reactors used in present-day production have a guaranteed service
life of 100,000 hours, retain gas-tightness at a temperature of up to 950�C
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? 900-(1154=1500)'C BSQ'C 900-(1150T1300) 'C 950'CyonveYSion
conversiA\,~He or sol
0- (orsol products ' oolant) Pr~ducts
' Q,d coolan 0 30D f1CQ7=1100J'C SOO1C
500'C 6 y
BD0 -(1100=1100J CH4*H20 3501C
a CH4 +N20
9A0-(I150=1,i00J�C 850 ~
2 conversion c
~e (or soli products
~ coolant) 3 Fig. 2. Versions of utilizing
0 .~oo-~icna�r2oo1' S50'C
7 6 5 4 nuclear re:.;tor in ammonia pro-
~ 350�C duction: 1--reactor; 2--coolant
CN4+NZG loop; 3--converter; 4--heater for
b steam-gas mixture; 5--steam super-
heater; 6--steam generator; 7--
water heater
Structure of heat utilization (Gcal/hr) in energy scheme
of ammonia production facility with capacity of 3000- metric tons per day
Index
Conversion of steam-gas mixture
Heating steam-gas mixture
Steam generation
Steam superheating
Heating feed water
Heating steam-air mixture
Heating gas mixture
Heating fuel gas
Total power from heat source
Thermal power of reactor not
counting internal needs, MWt
With use of
Without nuclear reactors
using
nuclear only for for conversion and
reactors conversion generating steam
(initial at pressure of 110 atm
version)
Version 1 Version 2 Version 3
185.0
26.4
68.0
185.0
26.
68.0
185.0
26.4
68.0
185.0
26.4
238.5
113.1
113.1
113.1
-
57.4
57.
57.
-
12.0
12.0
12.0
12.0
10.6
10.6
10.6
10.6
0.2
0.2
-
-
472.7
472.7
472.5
472.5
-
215.0
525
525
N ote: Above the broken line the heat is supplied by the nuclear
reactor; below by burning natural gas
and pressure of 30-40 atm, and may serve as an engineering base for making new
tubular conversion reactors that take heat from the nuclear reactor coolant;
a considerable part of the produced hydrogen (theoretically up to SOy) is
formed from water rather than from methane (CHq+ 2H20= C02+ 02), which means
.
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that steam catalytic conversion of methane based on nuclear reactors can be
considered as a first step on the road to producing hydrogen from water.
The possible versions of using nuclear reactors, distribution of the thernnal
power of the reactor with respect to consumers in ammonia production based
on two-stage catalytic conversion of inethane, and the principal values of
technolog'Lcal parameters and working media are giver. in Fig. 2 and the table.
Version 1(see Fig. 2a) assumes utilization of heat from nuclear reactors
only in the high-temperatuce part of production--in the first stage of methane
conversion for heating reaction tubes.
Heating of the steam-gas mixture, the steam-air mixture, feed water, gas mix-
ture, superheating of steam and generation of saturated steam in an auxiliary
boiler are accomplished in this case by heating nacural gas. All equipment
remains unchanged other than the heat-utilizing fAcility, which is slightly
modified. Starting conditions and transient processes are unaltered. The
advantages of this version are: relative simplicity of construction of the
reactor unit that heats only one flow the steam-gas :aixture in the con-
version process, and most complete retention of the basic technc.logical equip-
ment for ammonia production (excepting the tubular furnace). Its disadvantage
is the small fraction (about 20%) of liberated natural gFs from the total
production requirement, and low reactor power.
Version 2(see Fig. 2b) almost totally obviates the use of natural gas as
a fuel. In this connection, the heat necessary for steam conversion of nethane,
heating the steam-gas mixture, feed water, and also for generating and super-
heating steam with pressure up to 110 atm is provided by a high-temperature
reactor. Version 3(see Fig. 2c) is distinguished by the fact that the reactor facility
is used only for steam conversion of inethane, heating the steam-gas mixture
and producing saturated steam at pressure up to 110 atm. In view of the com-
paratively small required total thermal power (500-600 MW), high-temperature
heat and energetic steam can be produced by a single reactor even for long-
range process arrangements. In practice, the area of a single chemical combine
accommodates several production facilities (ammonia, methanol, higher alco-
hols, etc.) with various technological chains in each one, and from arguments
of economy and standby capabilities it is advisable to arrange parallel con-
nections between the individual technological chains within each facility
and among facilities. This may make it possible to centralize production
of hydrogen-containing gas mixtures and energetic steam in separate specialized
reactor facilities.
The advantages of this arrangement are operating reliability of each technologi-
cal chain with lower expenditures on standby equipment and greater economy
and reliability of producing high-temperature heat and energetic steam on
specialized nuclear reactors.
Successful solution of the problem of hydrogenating large amounts of carbon
monoxide into methane with liberation of considerable amounts of heat may in
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future enable rational solution of problems of distributing the high-temperature
nuclear reactors over the area of the chemical,enterprise. Existing and newly
planned facilities in.the nitrogen industry must meet the following require-
ments i
prolonged accident-free operatioti over the.established work life (up to 30 years);
high level of utilization of installed power of process equipment (up to
8000 hr/yr);
high reliability of all machines and equipment incorporated in the techno-
logical chain;
ease of control, high degree of automation;
high level of labor productivity, maximum output from each worker;
minimum necessary consumption of raw materials and energy resources;
transportability, producibility and repairability of all equipment;
high economy of operation of entire facility.
Direct utilization of the heat of high-temperature nuclear reactors in the
energy-consuming industrial processes of ammonia and methanol production
also involves solution of some technological problems, chief among which are:
developing and producing reliable accident-free reactors with coolant tem-
perature of 900-1400�C at the core outlet and total working life of up to
30 years with yearly r.ontinuous-duty operation of up to 8000 hours;
development of reliable and efficient technical facilities for heat transfer
from the reactor core to the working volume of process equipment;
working out engineering measures to ensure protection of final goods and tech-
nical equipment from radioactive contamination;
.,olution of the problem of diffusion both from the core into the process chan-
, nel and in tte reverse direction.
Even ncw when operating facilities for amnonia production with a capacity
of 1360 metric tons per day, considerable difficulties arise in th~l matter
of training operating personnel with appropriate skills. The introduction
of nuclear-chemical complexes requires implementation of additional steps
in this direction, since the working conditions for service personnel in the
process nart of the facility will evidently be on a par with those of nuclear
elec*iic plants; requirements will change with respect to the makeup by
specialties and the training of service personnel.
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Conclusions
1. The current state and engineering prediction of development of high-tempera-
ture reactor equipment and nuclear fuel technology, the start that has been
made on research and development, and analysis of the outlook for making a
nuclear-chemical f acility for ammonia and methanol production based on methane
conversion allow us to count on organizing an experimental industrial plant
of this type in the next 10-15 years.
2. The major prob lems in making and using hi gh- temperature nuclear reactors
in industrial processes of inethane conversion are in the area of transferring
~ the high-temperature heat from the core to the working volume for carrying
out the technological process with appropriate observance of conditions of
protecting products, service personnel and the enviroment from radiation at
the required level of rr_liability and redundancy of the nuclear power source.
REFERENCE
1. Dollezhal', N. A., in: "Voprosy atomnoy nauki i tekhniki. Seriya:
Atomno-vodorodnaya energetika" [Ptoblems of Nuclear Science and Engi-
neering: Series on Atomic Hydrogen Power], Preprint No 2, I. V.
Kurchatov Nuclear Power Institute [IAE im. I. V. Kurchatova], 1977, p 5.
COPYRIGHT: Institut atomnoy energii im. I. V. Kurchatova, 1977
Atomizdat, 1979
6610
CSO: 8144/1052-A
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~
"a
hOR OEFICIAL USE ONLY
UDC 621.039
PARTICULARS OF LAYOUT AND CONSTRUCTION OF EXPERIMENTAL INDUSTRIAL HIGH-
TIIMPERATURE GAS-COOI.ED REACTOR MODEL
Moscow ATOMNO-VODORODNAYA ENERGETIKA I TEKHNOLOGIYA in Russian No 2, 1979
(signed to press 8 Jun 79) pp 73-77
[Article by F. M. Mitenkov, Yu. N. Koshkin, U. B. Samoylov and Ye. V. Komarov]
[Text] An examination is made of the layout and construc-
tion features of the reactor, and also the problems to be
solved on an experimental industrial facility. The proposed
plan allows development of the facility by stages with dif-
ferent temperature levels.
High-temperature helium-cooled reactors are a new field in nuclear power.
Their distinguishing feature is the feasibility in principle of getting heat
with a high temperature up to 1000�C or more. Such a temperature potential
cannot be attained in other power reactors currently known.
Possible ways of utilizing high-potential heat have been extensively studied
in the USSR and elsewhere. Research shows that raising the temperature of
the heat generated in a reactor to 750-800�C enables utilization of modern
turbines with high steam parameters (ts = 530-580�C). It is evidently inad-
visab le to further increase the temperature for a steam-turbine cycle.
There is a much better outlook for using the HTGR in a gas-turbine cycle,
and also as a source of thermal energy for technological processes in various
sectors of the national economy in which 70-80% of all generated energy is
consumed as heat, particularly in the most energy-intensive processes of the
chemical and metallurgical industry. Analysis has shown that to replace fossil
fuel with nuclear fuel in these processes the coolant temperature must be
950�C or more. Such a temperature is attainable in the HTGR, opening up ex-
tensive possibilities for using the reactor in this field.
The use of high-temperature heliwn-cooled reactors in a high-energy process
facility for producing thermal energy is looked upon as the major area for
reactor utilization. Design developments have revealed that combined production
of electric power and high-potential heat is most advisable from the econrnaic
:ztandpoint, enabling effective uCilization of generated heat with fairly high
e:rficiency.
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Development of commercial reactor installations for combined production of
high-potential thermal energy and electric power necessitates a large voliune
of scientif ic research and experimental design work, enabling resolution of
some engineering proUlems relating to the production, conveyance and utilization
of heat with very high temperature, assimilating heliwn technology, working
out new kinds of equipment and new materials.
Considering the complexity of the problem, it seems necessary to make an ex-
perimental industrial reactor installation. The purpose of such a facility
is to check and confirm all major engineering decisions, and to develop the
principal equipment and control systems under conditions of industrial oper-
ation. The development of an experimental industrial facility will open up
the way for producing commercial models of the facility and using them on
a wide scale. In order that experience in developing, making and using the model might sub-
sequently to the maximum extent become a basis for making commercial instal-
lations, it is necessary first of a11 to make a correct and sound choice of
the direction of p].anning and the initial technical parameters, i. e. to work
out the optimum technical requirements for an experimental industrial model
of the high-temperature reactor with consideration of its ultimate purpose.
The technical requirements for an experimental industrial model stem from
the jobs it is to handle:
1) checking and working out the layout of the facility, including the process
loop;
2) checking designs of the principal kinds of equipment and systems, and re-
fining them from operational results, when such units are clearly to a great
extent unique (gas blower, steam generator, drives in the reactor control
system, heat exchanger, cleaning system, monitoring system, etc.);
3) operational. check of structural components and technology of reinforced
concrete vessel;
4) checking new heat-resistant structural materials under conditions of pro-
longed operation, atc.
The necessity for prelimirary solution of the enumerated problems precludes
the use of a low-power facility for this purpose. A reactor with thermal
power of 1000 MW can be recommended for the experimental industrial facility.
In this case the ma3or components of the installation (core, steam generators,
gas blowers, heat exchangers) will have been prototyped on a sufficient scale
fur future commercia] facilities.
Wtien working up the design of an experimental industrial facility, considerable
attention must be given to optimizing the layout of the facility with cunsider-
ation of ensuring potential capabilities of getting information when developing
new commercial energy-intensive technological complexes.
As an experimental industrial model we can consider a facility intended for
generating high-temperature heat that would be utilized for producing electric
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encrgy in a nuclear puwer plant. Realization of such a project would enable
us to work out principal solutions of the plan layout, as well as the equipment
in the facility, including the reinforced concrete vessel, to incorporate
into operation the fundamentally new loop with helium coolant, and to elaborate
the working conditions of the facility. In this case it would be much easier
to solve the probleTn of choosing structural materials since steam parameters
(ts = 530-550�1~) can be ensured at a temperature of 750�C in the first circuit.
However, this.version would lack capabilities for checking the plan of the
technological complex using high-potential heat and developing equipment at
a temperature necessary for the commercial technological complex.
;Che second way is to develop a direct prototype of the energy complex in full
scale, and hence with working temperature needed for supporting the given
process cycle (t= 950�C). This assumes development of new heat-resistant
materials, thereby pushing back the real deadline for making the experimental
industrial facility. Furthermore, it should be taken into consideration that
startup and alignment on this facility will undoubtedly take a long time,
and bringing the temperature up to working level will be gradual. Most of
the time on startup and alignment is taken up by the reactor installation
proper since it is the most complex and important part.
However, the disadvantages of the second version can in large measure Ue elimi-
nated if the design of the experimental industrial model allows development
of the prototype by stages. For this purpose, the facility diagrammed in
Fig. 1 can be suggested as an experimental industrial model of the installa-
tion. The heat produced in the energy unit can be used to generate electric
power in a turbogenerator, and also for producing hydrogen in a chemical
process coiplex.
i5
~
I

Fig. 1. Schematic diagram of experimental
industrial nuclear power-producing and pro-
cess facility (NPPF): 1--reactor; 2--high-
temperature intermediate heat exchanger; 3--
steam generator; 4--main gas blower; 5--chemi-
cal process circuit; 6--intermediate circuit;
9 7--gas blower for intermediate circuit; 8--
turbine; 9--generator; 10--condenser; 11--
condensate pump; 12--low-pressure water heater;
13--deaerator; 14--feed pump; 15--high-pressure
water heater
Recilization of the chemical technological process of hydrogen production re-
qtiires a first-loop coolant temperature of 900-950�C at the reactor outlet.
High-p:irameter steam is generated in the steam generator at temperature of
750�C at the inlet to the first loop. Such a facility can be manuf actured
and developed in three stages.
On the first stage the facility can be worked out on a temperature level up
to 750�C with generation of electric energy in a steam-turbine cycle (Fig. 2)
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Fig. 2. Sc'nematic diagram of NPPF for
first and second stages of operation:
1--reactor; 2--high-temperature inter-
7 mediate heat exchanger loop; 3--steam
generator; 4--main gas blower; 5--bypass;
6--turbine; 7--generator; 8--condenser;
9--condensate pump; 10--low-pressure
water heater; 11--deaerator; 12--feed
pump; 13--high-pressure water heater
without completing development and manufacture of the heat exchanger, devel-
oping high-temperature fuel or installing the equipment of the chemical complex,
by installing bypass pipes in place of the heat exchanger. Maximum reactor
power in this mode is 70% of Wnom� The flowrate of coolant in the first loop,
and all paramsters of the steam generator and gas blower will be nomina"l.
On the second stage, with the same makeup of equipment as on the first, the
temperature at the outlet of the core can be raised to 950�C, the former tem-
perature level being maintained in the steam generators by diluting the hot
coolant coming from the core with cool gas fed from the pressure side of the
gas blower through bypass 5(see Fig. 2).
On the third stage after making and installing the intermediate heat exchanger
and all chemical process equipment, as well as completing startup and alignment
work on the reactor facility, the entire facility shown in Fig. I can be de-
veloped and brought up to nominal power.
The plan for an installation with combined production of haat and electric
energy is the most optimtnn solution in choosing an experimental industrial
model of a facility with high-temperature helium reactor using thermal neutrons.
This model provides an excellent prototype for the most promising power-
producing and process f acilities, and the design of the equipment and layout
of the instllation allow the necessary multistage manufacture and operation
of the model.
COPYRIGHT: Institut atomnoy energii im. I. V. Kurchatova, 1977
Atomizdat, 1979
6610
CSO: 8144/1.052-A
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UDC 621.039
CHOOSING DESIGN CONCEPT AND PHYSICAL FEATURES OF HTGR CORE FOR ENERGY FACILITIES
Moscow ATOMNO-VODORODNAYA ENERGETIKA I TEKHNOLOGIYA in Russian No 2, 1979
(signed ti press $ Jun 79) pp 78-84
[Article by G. P. Goroshkin, A. S. Kaminskiy, V. D. Kolganov, Ye. M. Kuz'min,
M. D. Segal' and V. P. Smetannikov]
[Text] The authors consider the design of a high-temperature
channel reactor with spherical fuel elements with thermal
power of 540 MW with helium temperature of 950�C at the
outlet of the core. Physical and hydraulic profiling of
the core reduces the maximtun temperature of the fuel elements.
, The advantages of such a reactor over other designs are
indicated for use as part of a power-producing facility
_ with combined supply of energy to chemical, metallurgical
and other energy-intensive facilities.
Ttao directions of development of HTGR's are known in world practice: with
a stationary core and with moving fuel elements in the core.
The first direction is characterized by the use of large graphite blocks in
the form of hexagonal prisms in the core (reactors of the HTR type) with ninner-
- ous openings for accommodating the fuel and passage of the cooling gas. This
' type of core is stationary, and recharging requires.the use of loading machines
that operate relatively rarely when replacing individual depleted blocks of
the core or replacing the core in its entirety.
In the second direction, the core consists of a chargE of spherical fuel ele-
ments that contain both fuel and moderator. In a reactor with such a core
(type AVR and THTR) fresh elements are loaded and depleted elements are unloaded
continuously during operation at power for a prolonged period.
In our view, disadvantages of such directions in reactor construction are:
a) for the first type of reactor: shutdown of the reactor during reloading
of core blocks; variation of energy release distributions during reactor oper-
ation; comparatively long time for reloading core; high labor-intensiveness
and technological complexity of fabricating the graphite core blocks; consid-
erable thermal sCresses that arise in graphite blocks under high heat loads;
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2
1-
4
5.
b
Fig. 1. Agat reactor facility: 1--loading macl-iine; 2--
intermediate heat exchanger; 3--actuating mechanism of reac-
tor control system; 4--prestressed concrete reactor vessel;
5--convers?on furnace; 6--core
b) fcr the second type of reactor: nonuniformity of spherical fuel element
movement in the core; no capability for precise profiling af energy release
with respect to core radius; necessity for accommodating the rods of the reac-
tor control system in the charge of spherical fuel elements in the core in
graphite pylons specially provided in the body o� the core; possible fluctua-
tions of porosity in the charge of spherical fuel elements of the core over
the entire period of reactor operation; graphite moderator in the fuel ele-
ments, leading to additional expenditures when reprocessing depleted elements.
For industrial facilities (e. g. chemical, metallurgical and other sectors
of the national economy), uninterrupted operation of the reactor throughout
the technological production cycle is of decisive importance. This condition
is met to a greater extent by reactors with continuous fuel recharging during
oPeration at Power. Reactors with spherical fuel el.ements can be put into
this category. The Agat reactor (Fig. 1), designed tu produce high-potential.
licat fur the needs of chemical production, represents the first attempt to
develop a facility that would not have the disadvantages of the above-mentioned
reactors while retaining their positive features. The following design features
have been incorporated into the facil.ity: channel type core; regular geometry
of core channels and control rods; partial separation of moderator and fuel;
use of principle of one-time passage of fuel elements through the core; con-
tinuous reloading of spherical elements during reactor operation at power;
clpability for rearranging physic3l channels.
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The core of the Agat reactor is made up of graphite componerits with the excep-
tion of the upper spacing plate and the lowPr support plate. The fuel ele-
ments are r.iicrospheres in a graphite matrix enclosed in a graphite cladding
of spherical Shape 60 mm in diameter. The moc+erator of the core is a set of
vertically arranged prisms in which the fuel
elements move (Fig. 2). The prisms are.held
in the lower support plate by metal sleeves
that are hinged to the plate. Concomitant
movement of fuel elements and coolant is neces-
sary to maximize the' coolant temperature at
the outlet from the reactor at the permissible
temperature in the center of the fuel element
with the highest heat release rate (for example
~ we can have a coolant temperature of 1200�C
at the reactor outlet when the temperature
~ inside the fuel element is 1300-1350�C).
The core is surrounded by a graphite reflector
made of hexagonal prisms. Above it is an
Fig. 2. Core fragment upper heat shield made of graphite blocks.
The upper face of the core is conical; this
is necessary so that the depleted fuel elements can roll down past the boun-
daries of the core. The thickness of the side reflector in the radial direc-
tion averages 1200 mm, and that of the upper reflector in the axial direction
averages 1000 mm.
The reactor control system consists of 61 rods; these move in the same channels
as the fuel elements and have a working section with absorber 5 m long. The
drive mechanisms are situated on the cover of the reactor vessel.
Ttie equipment is configured in the following way.
The prestressed reactor vessel is a monolithic block with recesses'to accom-
modate the major equipment of the facility, emergency a�tercooling equipment,
and the channels of the reactor circuit (see Fig. 1). In the central part
of the vessel is a cylindrical space for accommodating the core, reflectors,
heat insulation of the vessel and lower support plate for the core. Above
the central cavity is a passage to accommodate the reactor cover. In the
lower part of the vessel under the core are seven vertical passages for in-
stalling the loading mechanisms for the spherical fuel elements. To accom-
modate the gas blowers af the reactor loop and the emergency aftercooling
system, horizontal passages are provided in the reactor vessel in which the
blowers are placed toget}ier with their drives. A hermetically sealed carbon
steel facing (liner) fills the inside of the reactar vessel. To ensure the
necessary temperature conditions for operation of the vessel, the liner is
protected by heat insulation with a gas layer a so-called gas wall and
tubes of a water-cooling system are buried in the concrete of the vessel at
a certain distance from the liner. The cylindrical cavity of the housing
is divided into two sections by the lower support plate, which has a passage
for the fuel elements under each channel of the core, and openings for passage
of the coolant that serves the moderator and reflector. A device for feeding
fresh fuel elements into the channel is placed in each channel opening of
the plate.
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Seven reloading mechanisms provide continuous reloading of fuel elements during
reactor operation. The spherical elements are loaded into each channel, each
serving its own part of the core. When one of these mechanisms fails, the
reactor is capable of operation at somewhat reduced power (about 15% below
nominal) right up to shutdown for routine maintenance.
The design with lower placement of reloading mechanisms is chosen for the
fotlowing reasons: the supporting structures of the core are situated in
the zone of "cool" coolant; operation of the reloading mechanisms is facili-
tated by the absence of control rods under the core; the force from the pres-
sure differential in the core is directed opposite to the force of gravity
of the fuel elements and side reflector; coolant circulation in the reactor
circuit coincides with the direction of motion of the coolant with natural
circulation in the reactor in case the gas blowers stop; the principle of
one-time passaqe of fuel elements through the core requires concomitant move-
ment of fuel elements and coolant.
The propased design has typical physical features of the HTGR, among which
we note the following: 1) use of graphite, a weak absorber of neutrons, as
the moderator, which ensures neutron economy, improves the breeding proper-
ties of the core and reduces the charge of uranium compared with other types
of thermal reactors; 2) bur.nup is much higher than in other reactors: about
105 MW�days per metric ton; 3) the negative temperature coefficient of reac-
tivity and large heat capacity of the core ensure a high degree of safety.
At the same time, the proposed design successfully combines the advantages
anci avoids the disadvantages of the spherical and prismatic forms of HTGR
fuel elements.
Regularity of the passage of fuel elements through the core in the proposed
ch:innel reactor enables the use of a var.iety of effective methods of profiling
the field of energy release. The necessary profiling through the core body
can be achieved by: varying the fuel. enrichment with 235U in profiling zones;
c.hanging the rate of passage of the fuel elements in profiling zones; varying
the density of the graphite moderator.
Temperature fields in the core body can also be equalized by hydraulic pro-
filing.
II, kW - -
~
20 - - -
15 L 0 50 100 150 200 250 R, cM
Fig. 3. Channel power distribution with
respect to core radius
ln the proposed reactor, three-zone profiling of the core is provided by using
two types of fuel elements differing wi.th respect to 235U enrichment and moving
25
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at different velocities. Enrichment of fuel elements in the central (I) and
peripheral (III) zones is 6.5%, and in the intermediate zone (II) 10%.
This reduces the coe�ficient of nonuniformity of channel power along the radius
of the reactor to 1.05 (Fig. 3).
The proposed design extends capabilitiss for controlling the process of bring-
ing the core up to steady-state operation. This is done by providing ccntrol-
lable placement of absorbing elements specially introduced into the channels,
and corresponding configuration of the control rods.
The proposed channel reactor design facilitates physical monitoring of core
parameters at the necessary nunber of points to get reliable information,
enabling on-the-spot correction of the coeff icient of nonuniformity of energy
release, raising the specific and thermal load on the fuel, and increasing
the average burnup. The configuration of the control rods in the channels
can also be optimized, which raises their effectiveness. In this way, the
design has advantages over the AVR, where the control rods are situated around
the periphery of the core in individual pylons.
f, .C
1250
1200 t c
1150
1100
QK' [f050
rel s
0,95 950
0,90 900
pgg 850
0
Q
0,5 1,0 1,5 2,0 2,5 R,A:
Fig. 4. Distribution of rela-
tive heat release Qk, gas tem-
perature t~ and temperature in
the center of the fuel core tc
along reactor radius R
_Z
NK
0,8
0,7~
0,6
0,5
0,4
0,3
0,2
0,1
Fig. 5. Distribution of rela-
tive heat release QZ, gas tem-
perature tg, cladding temperature
tcl and temperature in the center
of the fuel core tc over the height
of the core Z /Hk
Proy;rammed motion of the fuel elements in each channel assumes stable dis-
tribution of energy release heightwise of the core throughout a run, and con-
comitant motion of fuel elements and coolant maximizes the gas temperature
at the core outlet at the permissible fuel temperature. Thermohydrauiic cal-
culations for steady-state operation iiave shown that the temperature distri-
bution can be kept fairly uniform through the core body (Fig. 4, 5):
Stiidies have shown that nonuniformity of the coefficient of heat transfer
over the surface of a sphere {rom maximum to minimum is 2.2-3.0, and as a
result, heightwise the maximur.i fuel temperature may be 10�C higher.
26
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The feasibilitv of hydraulic profiling was checked for the selected version
of physical profiling to determine the effectiveness of equalizing temperature
fields along the radius of the core. For a channel diameter of 74 mm in physi-
cal profiling zone II and 75 nnn in the other zones, the gas temperature at
the outlet from the channel cores can be equalized and maximwn fuel temperature
can be reduced by 20�C compared wi.th the hydraulically unprofiled version.
When the channe.l_ diameter is 72 mm in physical profiling zone II, a reduction
in maximu~.n fuel temperature by 50�C is achieved, but nonuniformity of gas
temperature at the core outlet is increased to 130�C as compared with the
110�C nonuniformity for the unprofiled version. Of course, hydraulic profiling
somewhat increases the hydraulic drag of the core: for the unprofiled version,
hydraulic drag is 0.39 kgf/cm2; with channel diameter of 74 mm, drag is
0.41 kgf/cm2, and for 72 mm 0.45 kgf/cm2.
, Based on thermohydraulic calculation, we can conclude that physical profiling
enables attainment of the necessary equalization of temperature fields, while
hydraulic profiling in the given case is less effective and requires at least
two sizes of channels in the core.
As a result of design analysis, neutron-physics and thermohydraulic calcula-
tions of the reactor, the following characteristics are obtained:
Thermal power of reactor, MW
538
Dimensions of core, m:
6
diameter
5
height
Number of channels
3481
Fuel element diameter, mm
60
ChargP of uranium, kg
6570
Enrichment,
6
5
in profiling zones I and III
.
in profiling zone II
10
Run, days:
for fuel elements of profiling zones I and III
860
for fuel elements of profiling zone II
800
Reactor coolant
heliian
Coolant flowrate in reactor circuit, kg/s
160
Coolant temperature, �C:
at inlet to reactor
306
at outler from reactor
950
Coolant pressure in reactor circuit, kgf/cm2
40
In summary, we can conclude that engineering and design cal.culations have dem-
onstrated the feasibility of developing a high-temperature gas-cooled reactor
combining the advantages of cores of channel and microsphere types. Such
a reactor can be used to produce high-potential heat in the chemical, metal-
lurgical and other energy-intensive sectors of the national economy.
COPYRIGHT: Institut atomnoy energii im. I. V. Kurchatova, 1977
Atomizdat, 1979
6610
CSO: 8144/1052-A
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UDC 621.039
MOLTEN-SALT REACTOR WITH NATURAL CONVECTION OF FUEL MIXTURE AND OPEN GAS-
TURBINE AIR CYCLE
Moscow ATOMNO-VODORODNAYA ENERGETIKA I TEKHNOLOGIYA in Russian No 2, 1979
(signed to press 8 Jun 79) pp 85-93
[Article by V. A. Legasov, I. G. Belousov, N. K. Yerokhin and A. S. Doronin]
[Text] The authors consider some problems of using a high-
temperature molten-salt reactor with natural convection
of the fuel mixture in the primary circuit. Radiation pro-
vides the thermal coupling between the primary circuit and
the energy (or process) circuit. Reactor heat can be used
at a temperature near maximum. Combining a reactor of this
type with a gas-turbine facility operating on an open cycle
gives efficient conversion of heat to electricity.
N,tclear reactors with molten salt fuel mixture are in many respects an instruc-
tive phenomenon in nuclear power. lnterest i.L these reactors arose in tiie
1950's in the United States in connection with development of a nuclear air-
craft [Ref. 1]. An experimental reactor with pawer of 8 MW (MSRE) [Ref. 2]
was operated at Oak Ridge National Laboratory from 1966 to the spring of 1968.
During this period, 70,000 MWh of electricity was generated, and extensive
experimental material was acctuuulated on many aspects of scientific and design
developments of molten-salt reactors, which was the basis of project MSBR
a one-fluid breeder reactor with thermal power of 2250 MW [Ref. 3, 41.
Project MSBR was shelved, although the anticipated properties of a nuclear
power plant of this type opened up unique prospects both from the standpoint
of utilizing (Ref. 51 and reprocessing [Ref. 6, 71 nuclear fuel, and from
the standpoint of capital expenditures [Ref. 41. There was no more really
serious work on the problem of the molten-salt reactor, and poorly advised
attempts to revive interest in the idea have led to considerable discreditation.
There are two reasons why it was quite natural to stop work on pro,ject MSBR.
First of all, the power industry was not yet feeling any scarcity of nuclear
fuel, ar.d with the comparatively low fuel component in the cost of nuclear
electricity, extra efforts to develop technology for processing salt fuel
seem economically superfluous even in an ideal fuel cycle like that of the
MSBR. In the second place, the coolant temperature attained in the primary
28
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circuit (700�C) is lower than in the HTGR, and further upward movement on
the temperature scale is impeded by the particulars of interaction of salt
with structural materials based on metals. On the other hand, the low capital
component of the MSBR is obtained by calculation and cannot be taken as a
conclusive argument.
The situation is considerably altered if inetal-based structural materials
are eliminated from the molten-salt energy circuit. An isothermal high-
temperature molten-salt reactor design (VTRS) has been proposed in which the
only metallic structural materials in the salt circuit are in the pump group.
The purpose of our research is to examine the peculiarities of a VTRS in which
the only structural matarial in contact with the molten-salt fuel mixture
is isotropic pyrolytically precipitated graphite. The thermophysical layout
of a power plant with such a reactor is exceptionally simple, and the attainable
level of the coolant temperature in the primary loop, and also the thermodynamic
quality [Ref. 81 of the reactor may be anomalously high. Heat from the core is
transferred to a radiant heat exchanger by natural convection of the fuel mix-
ture in graphite coaxial fuel elements. The resultant efficiency of a nuclear
power plant with VTRS may reach 50-60%, and the thermodynamic quality of the
reactor may be 0.95-0.98. Thus the way is opened
up for considerable improvement. of high-temperature
nuclear heat sources for future pawer 3nd process
applications.
~
~
I
.
I
I
~
A temperature difference between the inner and outer
colwnns of liquid (Fig. 1) is obtained by external
heat removal in a radiant heat exchanger from the
upper part of the fuel element, and volumetric nu-
clear heating of the lower part. Pararaeters of
flow of the molten salt in the circuit of the fuel
element are: Gr - (gd2 l.pl10) 0_%l, �Dl(D + d) - 10" = 1010;
Re = p::,d; Et - 5- 103 5- 10';
Nu = ad,'}. - 50 100;
St = Nu/(f',e�Pr) - 10-3 = 10-2,
where Gr, Re, Nu, St are the Grashof, Reynolds,
nusselt and Stanton n,unbers. Conventional symbols
are used to denote the parameters. The law of cir-
culation is approximately described by the dime:-
sionless equation [Ref. 9]
I D-d 3 D-}-d 1�75 Gr = 2,5Re~ I 4 > ( d >
~
Fig. 1. Design of co-
axial fuel element with
natural convection of
salt fuel mixture (LiF
--0.75, ThF4--0.24, UF4
--0.01)
D- d D= - d= Reo.2s
~2)
L+ I c1= )21 0,316 '
We can see from equations (1) and (2) that a chanoe
in temperature head Atl by a factor of nearly 100
is possible between the salt and the outer wal.l of
29
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the graphite channel in the vicinity of the radiative heat exchanger without
changing the turbulent flow state (Blasius interval). In other words, at
a fixed average temperature of the salt fuel mixture the external load can
be reduced by two orders of magnitude while maintaining stable circulation.
,
6 ~
3 ~o ro
2
0~
B00 1000 /ZAD 1fQ0 T, K
1
0 10 20 30 H, kcal/mole
Fig. 3. Example of process consumption of heat from the
nuclear reactor (conversion of coal to carbon monoxide)
-
t, . C ~
f000I
5,00
a~lp 100 ar,
~oe
0,5 47 1a
Fig. 2. Typical diagram of
temperature distribution in
the salt circuit, the outer
jacket of a fuel element,
the heat exchanger tubes and
air heated therein
to
64Ne
kP - exP~ R j* 28,2)
_
0 10 "LU JU 9U Pv uv /mole
Fig. 4. t-H diagram of heat consumption on the first stage
of a two-stage water thermolysis cycle
The nuclear reactor is made up of a series of fuel elements. The temperature
differential lengthwise of the fuel element cladding (Fig. 2) in the zone of
30
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,'C -
a. ~p p kp - exp ~ R . f*2~
100 arm
10 i 2INe
000
10 10
la
a5 d 1
S00
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heat removal is about 100-150�C. Consequently, all high-temperature reactor
heat can be transferred to the heat exchanger at a temperature little differer.Lt
from uaximun. This explains the high thermodynamic quality of the VTRS as
a heat source. Fig. 3 and 4 show examples of possible process consumption
of nuclear heat of comparatively high quality. For example, when coal is
converted to carbon monoxide (Fig. 3) according to the Boudoir reaction
2C0 2Ht0 2COt } 4H:; 1 (4)
AGz 9,55 + 1,9 T1300 kcal/mold
with subsequent exothermal generation of carbon dioxide
COt C � 2C0; (
AG, = 41,3 - 12, 6 T1300kca1/mo1~_
(3)
It can be seen that carrying out reaction (3) at pressure of 10 atm involves
compensation of the reaction energy in the temperature range of 750-1050�C.
Fig. 4 shows the t-H diagram of heat consumption in the temperature range
of 820-1000�C associated with the first stage of the two-stage cycle of water
thermolysis:
1: 2M11rO:, . 11n -MII_% ; ?(.n. 1 L;
;1C., . 65,5 - 22.1 T/300kca1/mo14 LsO;
ll: Mn.,% 2C0_ ; 2,%WC0., I- 1,20.;
~1G4 8, I-- 12,6 T/300kca1/mole )
~G~
Obviously hig:i-efficiency thermolysis cycles can be realized only in the case
where an external heat source is capable of compensating the reaction energy
in a narrow range of temperatures close to maximimm. An example of a cycle
of water dissociation with low-quality heat from the source is the sulfuric
acid cycle:
i: ii:sol so, + }izo;
11: SO3 SOs + i/2Oz;
111: 2Hz0 + SOs FI'SO4 -f-'i=
11

0 25 50 75 ,H,. kcal/mole
Fig. 5. t-H diagram of'heat consumption on two stages of
sulfuric acid dissociation: I--S03-S02+;102i II--H2SO4-*
S03 + H20
31
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(7)
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The t-H diagram of the first two stages is shown on Fig. 5. The reaction
energy is compensated in a temperature range of 370-900�C, and in this case
we cannot count on high energy efficiency of the cycle corresponding to the
maximum temperature of the heat source. It is known [Ref. 10] that the limit-,
ing thermodynamZc efficiency of a cycle of dissociation of a substance inCo
components does not depend on the use of either exceptionally high-temperature
heat or electricity obtained from the Carnot-cycle machine. The limiting
efficiency of the dissociation cycle,is uniquely determined by the nature
of the substance being decomposed and the extremum temperatures of the heat
source and drain. In this sense the process utilization of high-temperature
nuclear heat in the thermolysis cycle does not automatically give any advan-
tages over the traditional method of generating electric energy and subsequent
electrolysis (or plasma-chemical reaction [Ref. 11]). High-productivity elec-
trolyzers or plasma-chemical reactors combined with a good electric power
plant may be preferable to thermolysis process facilities. Therefore the
method described below for converting high-temperature and high-quality nu-
clear heat to electricity by using a gas-turbine installation with open air
cycle may be taken as a component in development of one of the important ele-
ments of process utilization of nuclear energy.
I -----1 r
6 ~ti=-50�C
0, g8atm I
P�
~ ~
0,5~576 atm ;
0,4
0,7 0,8 0,9 bz 0 S
a b
--`l-
~A-7 _2
5/ 3
y
c
Fig. 6. Simple thermodynamic cyzle (b) of gas turbine power
plant with open air cycle (rlt = i.95; nc = 0. 9; e= 0.8; tg =
1273 K; SI:- n 8;); dependence of efficiency on total hydrau-
r=i
lic losses in the air channel and air temperature at the
compressor inlet (a) and schematic diagram of realization
of this cycle based on VTRS heat (c): . 1--reactor; 2--heat
exchanger; 3--turbine; 4--regenerator; 5--compressor
A comparison of temperature curves for salt coolant and air in a radiative
heat exchanger (Fig. 2) shows the possibility of a further appreciable increase
in efficiency of utilizing high-temperature heat of the VTRS. However, some
open gas-turbine cycles even on the given stage of optimization enable us
to get quite high characteristics (Fig. 6). The air temperature at the inlet
to the turbine is taken as 1000�C. The degree of regeneration of 0.8 is near
optimum; the total relative hydraulic losses through the channel 1- dE are
equal to -0.14. The resultant efficiency of the power plant is 0.44 for air
32
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temperature of +20�C at the inlet to the compressor, and 0.53 for air tempera-
ture of -50�C. We should also take note of the low maximum pressure in the
air circuit for temperatures of +20 and -50�C: about 6 and 8 atm respectively.
This factor is quite significant for getting a reliable design of a high-
temperature radiant heat exchanger. It is proposed that grade KhN45Yu steel
be used as the construction material. The wall thickness of the hot tubes
of the heat exchanger is 3 mm. At a temperature of 1100�C and pressure dif-
ferential of 1 atm, the long-term (10,000 hr) strength reserve is 2.
T Tg
p=6,5-18,5at
TK-540K
4
T in
' 0,5 0,6 0,7 0,8 0,9 8r 0 S
a b
~ tin`'50�C
0,6 p=13,4 atm
,
0,5 p~9,45
94 tin20�C
n;
J
c
Fig. 7. Characteristics of open gas turbine cycle with
one intermediate heater (conditions and notation same as
on Fig. 6)
More complicated cycles enable us to improve the efficiency of the power
plant (Fig. 7). The schematic of the power plant is a little more complex,
but there is an appreciable gain in eff iciency. For example at ambient air
temperatures of +20 and -50�C the efficiency of a gas-turbine unit with two-
stage heating is 0.48 and 0.56 respectively. The next step in improving the
'1
-
07
'
tins-50�C
p=23 atm\
0,6
patm
0,5
0,4 1-1
0,5 0,6
r
P=17: 27 atr
T=\0
0,7 0,8 0,9 6F
a b
6
c
?
Fig. 8. Characteristics of open gas turbine cycle with
intermediate cooling of air in the compressor (conditions
and symbols the same as on Fig. 6; b--intermediate cooler)
cycle involves adding an intermediate stage of air cooling in the compressor
(Fig. 8). The efficiency of the facility increases to 0.52 and 0.60 for air
temperature at the inlet ri the gas turbine unit of +20 and -50�C respectively.
Optimum pressure in the air channel is 18.5 and 23 atm. Fig. 9 shows a clear
comparison of efficiency of complex gas-turbine cycl.es for different air tem-
peratures at the inlet to the compressor. As these data imply, it makes prac-
tical sense to develop complex cycles although pressure increase is a con-
straint.
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p-23at
p=11,y atm
1,6 3
9,5 3
1 p1,98at
7~4 2 j ,
~3
- '
~
p-1gfat
p=Batm r ~
~ p=576. a
' 0,5 0,6 0,7 0,8 0,9 6Z 45 0)6 47 0,8 0,9 6Z
a b c
Fig. 9. Comparison of efficiency of complex gas-turbine
cycles at different air temperatures at the inlet to the
compressor: a--nt = 0.95; nc= 0.9; e= 0.8; titl=-50�C;
Tg = 1273 K; b--nt = 0. 95; nc= 0. 9; E= 0. 8; tin = 20�C;
Tg = 1273 K
We must emphasize a number of f.actors that are organically related to direct
process utilization of VTRS's, or to their use in combination with a gas tur-
bine power plant.
1. The attainable level of working temperatures of structural components
based on graphite-salt opens up wide vistas for conquest of the h3gh-tempera-
ture region.
2. The use of natural circulation of the fuel mixture in the primary circuit
- obviates the need for developing a high-temperature pwnp group, and enables
transition to nuclear sources of high-temperature, high-quality heat.
3. The fuel cycle of molten-salt reactor systems is among the most promising
both from the standpoint of complete utilization of fissionable materials
(uranium, thorium, plutonium) in accordance with the sequence
u_
1
and from the standpoint of reprocessing and disposal of fission products.
4. Absence of a water-cooling energy cycle makes nuclear power plants of
the proposed type independent of a source of cooling water and more ecological
compared with conventional plants.
5. Thc attainable efficiencies in the VTRS give access to new possibilities
in technology of converting heat to electricity.
6. The fact that industry is prepared to produce air gas-turbine plants of
the required class gives important economic advantages to development of
34
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high-temperature nuclear pawer in the direction of the VTRS. Research orgzni-
zations, design offices, metallurgists and machine builders have accumulated
adequate experience for f ormulating and solving the problem of making gas-
turbine power plants of the required class.
REFERENCES
1. Bettis, E. S., NUCL. SCI. ENGNG, Vol 2, No 6, 1957, pp 804-825.
2. Haubenreich, P. N., Engel, I. R., NUCL. APPL. TECHN., Vol 18, 1970, p 118.
3. Tosenthal, M. W., Kasten, P. R., Briggs, R. B., NUCL. APPL. TECHN., Vol 8,
1970, p 170.
4. Bettis, E. S., Robertson, R. C., NUCL. APPL. TECHN., Vol 8, 1970, p 190.
5. Engel, I. R., Kerr, H. T., Allen, E. I., TRANS. ANS., No 22, 1975,
pp 705-706
6. Grimes, W. R., NUCL. APPL. TECHN., Vol 8, No 2, 1970, pp 137-153.
7. Kashcheyev, I. N., Zolotarev, A. B., "Pirokhimicheskiye metody regeneratsii
metallicheskogo i solevogo yadernogo topliva (obzor patentov i
nauchno-tekhnicheskoy literatury 1956-1972 gg.)" [Pyrochemical Methods
of Regenerating Metallic and Salt Nuclear Fuel (Survey of Patents and
Scientific-Technical Literature for 1956-1972)], Moscow, Gosudarstvennoye
ob"yedinennoye nauchno-tekhnicheskoye izdatel'stvo, 1973.
8. Belousov, I. G., in: "Voprosy atomnoy nauki i tekhniki. Seriya: Atomno-
vodorodnaya energetika" [Problems of Nuclear Science and Engineering:
Series on Atomic Hydrogen Power Engineering], Preprint, I. V. Kurchatov
Institute-of Nuclear Power [IAE imeni I. V. Kurchatova], No 2, 1977, p 152.
9. Belousov, I. G., "Th.rmal Physics of Fuel Elements With Natural Circulation
of Molten-Salt Fue' ',xture" in: "Voprosy ato�,nnoy nauki i tekhniki. Seriya:
Atomno-vodorodnaya entrgetika", Preprint, I. V. Kurchatov Institute of
Nuclear Power [IAE imeni I. V. KurchatovaJ, No 1(4), 1973, p 201.
10. Belousov, I. G., in: "Voprosy atomnoy nauki i tekhniki. Seriya: Atomno-
vodorodnaya energetika", Preprint, I. V. Kurchatov Institute of Nuclear
Power [IAE imeni I. V. Kurchatova], No 1, 1976, p 65.
11. Belousov, I. G., Legasov, V. A., Rusanov, V. D., in: "Voprosy atomnoy
nauki i tekhniki. Seriya: Atomno-vodorodnaya energetika", Preprint,
I. V. Kurchatov Institute uf Nuclear Power [IAE imeni I. V. Kurchatova],
No 2, 1977, p 158.
COPYRIGHT: Institut atomnoy energii im. I. V. Kurchatova, 1977
Atomizdat, 1979
6610
CSO: 8144/10; 2-A
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UDC 621.039
PHYSICAL FEATURES OF HTGR WITH CIRCULATING FUEL
Moscow ATOMNO-VODORODNAYA ENERGETIKA I TEKHNOLOGIYA in Russian No 2, 1979
(signed to press 8 Jun 79) pp 94-100
[Article by N. N. Ponomarev-Stepnoy, A.
Grebennik, V. Ye. Demin, V. S. Malkov,
Tsurikov and L. A. BogatovaJ
N. Protsenko, Ye. S. Glushkov, V. N.
L. K. Malkova, 0. N. Smirnov, D. F.
[Text] An investigation is made of the characteristics of
a high-temperature gas-cooled thernal reactor with graphite
moderator for combined utilization of high-temperature heat
and gamma radiation of spherical fuel elements circulating
in the system made up of the reactor and irradiator. A
curve is given for the way that the power of gamana radiation
of the irradiator depends on the multiplicity of fuel circu-
lation in the system. An examination is made of the particu-
lars of fuel burnup effects compensated by absorbing elements
that circulate concomitantly in che system made up of the
reactor and irradiator.
The main tendency in the deve.lopment of nuclear power at the present time
is expansion of limits of application not only in electric power production,
but also for producing high-temperature heat, energy supply to the metallurgi-
cal industry and production of reducing agents for metallurgy, power and heat
supply to many sectors of the chemical industry, stimulation of chemical pro-
cesses, etc. [Ref. 1-31. Research has shown [Ref. 3] that these problems
can best be solved by using high-temperature reactors with helium coolant.
In particular there is a certain interest in the use of a nuclear reactor
as a source of radiation for radiation-chemical processes. In this connection,
use is made of gamma radiation of fissuion products in the radiation circuits
as fuel is circulated in the reactor-irradiator system [Ref. 5, 6].
A diagram of fuel circulation is shown on Fig. 1. The reactor is the high-
temperature VGR-50 with graphite moderator, helium coolant and spherical graph-
ite fuel elements based on microspheres with multilayer coating [Ref. 41. The
reactor permits combined use of high-temperature tieat that is removed by the
helium coolant from the packed spherical fuel elements, and transportation
to the irradiator of gamma-emitting fisGiun products in the makeup of the
irradiated fuel elements due to their circulation. It is desirable that the
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I
Fig. 1. Diagram of fuel cir-
culation in facility with com-
bined use of energy and gamma
radiation: 1--reactor; 2--
irradiator
time of delivery of the irradiated
fuel and its stay in the irradiator
be short (a few hours) so as to use
the radiation of short-lived isotopes
of fission products.
Fig. 2. Diagram of nuclear
reactor: 1--core (packing of
spherical fuel elements and
absorbers); 2--gap; 3--top end
reflector (graphite); 4--control
rods; 5--radial reflector (graphite);
6--bottom end reflector (graphite);
7--pylons (graphite pro3ections)
for control rods
A physical diagram of the nuclear reactor is shown in F'ig. 2. The radial
graphite reflector forms a cylindrical cavity that tapers into a cone in the .
lower part. In the bottom end reflector is an opening for unloading the con-
tents of the core. The reactor is covered by the top end graphite reflector.
The inner cavity is filled with spherical fuel elements and absorbers. Pro-
visions are made for changing the ntnnber of absorbers in the circuit during
circulation to compensate for fuel burnup during a run, the amount of ab-
sorbers not exceeding 15% of the amount of fuel elements. Between the top
end reflector and the spherical packing of the core is a gap that can be varied
over a wide range during reactor operation. The nominal size of the gap is
about 0.5 m. The main parameters of the reactor are'as follows:
Reactor pozaer, MW 140
Coolant helium
Helium pressure, atm 40
Helium temperature,
input/output,�C 270/800
Core dimensions,
D/H, cm 280/450
Spherical fuel element
diameter, mm 60
Content of U in one
fuel. element, g 2-5
Fuel enrichment, % 10-30
Power of gamma radia-
tion in irradiator, kW 400
To study the influence that multiplicity of fuel circulation has on the power
of gamma radiation in the irradiator, an analysis was made of experimental
data on power and the spectral makeup of fission products [Ref. 7-10]. As
a result of the analysis, an approximation formula is recommended for the
time dependence of power of gamma radiation of fission products after fission:
I'(t) = 1.5t-1'2+ 3.4t-1'4 MeV/(s�fission),
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where t> 103 s. The change in spectral makeup of gamma radiation is illus-
trated by the data of Table 1[Ref. 10]
TABLE 1
Spectral makeup of gamma radiation of 235U fission products
at different times after �ission, relative units
-
E
nerpv
ranQe,
MeV
Time after
fission, s
-
o,i-o,s I
o,s-i,o
I i-z
1 :-3 I
a-a
1 a-s
900
0,125
0,266
0,381
0,172
0,040
0,016
7,2�10.1
0,088
0,345
0,372
0,169
0,018
0,008
1,8�103
0,070
0,342
0,407
0,160
0,014
0,007
3,6�10'
0,103
0,376
0,410
0,098
0,008
0,007
8,6�10'
0,196
0,515
0,259
0,018
0,005
0,007
2,6�105
0,249
0,523
0,203
0,015
0,005
0,008
Theoretical studies h
tion of an irradiator
Pi
P'
P
ve shown strong dependence of the power of gamma radia-
on multiplicity of circulation in the system. Fig. 3
shows the way that the power of gamma radia-
tion in the irradiator depends on the multi-
plicity of fuel circulation. It can be seen
that an increase in the muitiplicity of fuel
circulation can raise the power of gamma
radiation by a f actor of approximately 10
as compared with the case without fuel circu-
lation. A further increase in the power
of gamma radiation is limited by the hold
of the fuel following the irradiator that
is necessary for de-excitation of delayed
neutrons (tdel = 10-20 min). The high core
temperature leads to a strong temperature
effect of reactivity. The change in effective
Fig. 3. Dependence of power breeding ratio as temperature increases is
of gamma radiation in the due to a ninnber of factors:
irradiator on multiplicity
of fuel circulation (P Doppler broadening of resonant levels of
power of gamma radiation; 238U as fuel temperature is increased;
Pp--reactor power; N--mul-
tiplicity of circulation (run a change in the spectrinn of low-energy neu-
of 1-2 years) trons, whic:i leads to a reduction in the
yield of secondary neutrons per absorption
in the fuel, a change in the ratio of absorption of the neutrons in the fuel
and in other elements of the reactor, and an increase in the square of the
diffusion path of thermal neutrons;
the temperature change in the density of reactor materials and the dimensions
of its components, which primarily affects neutron leakage.
Tahen studying reactor dynamics, it is important to break down the temperature
effect of the reactor into individual components, which we took as follows:
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the temperature effect of the fuel; this is the fastest-acting effect and
is due mainly to Dopper broadening of resonant levels of 238U;
temperature effect of the moderator associated with effects of thermalization
of slow neutrons, change in density of the moderator in the core and change
- of dimensions;
temperature effect of the reflector associated with the change in the spectrum
of low-energy neutrons and the dimensions and density of the reflector.
ek
Fig. 4. Components of reactor
temperature effect: 1--reflec-
tor effect; 2--fuel effect; 3--
moderator effect
dk ~
'k -
0,07 ~ -
0 10 20 CN?" kg/m3
Fig. 5. Influence that water in
the core has on reactivity (CH2o
is the amount of water per m3 of
the core with submerged (1) and
extracted (2) compensating control
rods)
The overall temperature effect is negative (Fig. 4), and the reduction in
the eff.ective breeding ratio as temperature increases may reach -10Y.
Fluctuation of the level of sphere stacking in the core influences reactivity
as a result of change in core height and the shooting effect in the cavity
between the top end reflector and the core. Dependence of reactivity on the
relative width of the gap between the top end reflector and the core stacking
(Fig. S) near the nominal stacking level has a rather flat slope, which is
due to the large height of the core compared with its diameter.
As the reactor operates, certain changes in the density of spherical packing
can be observed. In this connection, an estimate was made of the effect that
a change in the porosity of spherical packing of the core has on the effective
breeding ratio; this effect is characterized by the quantity dk/de= -0.33.
The high por.osity of the core makes such a reactor quite sensitive to hydrogen-
containing substances in the core (water, water vapor, etc.). The following
factors may influence the breeding ratio:
an increase in the moderating power of the core, leading to an increase in
the probability of avoiding resonant absorption of neutrons by 23BU (positive
effect) ;
a reduction in the length of migration of neutrons in the reactor, leading
to a reduction cf neutron leakage from the reactor (positive effect);
Absorption of neutrons in hydrogen (negative effect).
A typical maximum can be observed on the curve for reactivity as a function
of the content of water (water vapor) in the core (Fig. 6).
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_
Fig. 6. Reactivity as a function
of gap between top end reflector
and spherical stacking of the core
(d--gap; HQ--distance between top
and bottom end reflectors)
M
Ma
0
2
0,5
Fig. 7. Number of absorbers
in core to compensate burnup
effects: 1--units with burna-
ble absorber 10B; 2--units with
unburnable absorber; M--number
To compensate for effects of burnup of absorbers located in the core
with rapid circulation of fuel in the stacking; t/TK--ratio of elapsed
reactor, special absorbers of spherical time to run time
shape (like the fuel elements) are added
to the fuel charge of the core. As the reactor operates with circulating
fuel, provision is made for the capability of changing the nwnber of absorbers
in the spherical packing to maintain criticality.
Fig. 7 shows the change in the necesfiary number of absorbers during a reactor
run for two cases for burnable absorber (based on 10B) and for nonburnable
absorber.
The use of a"burnable" absorber, which does not necessitate a change in the
number of absorbers during a run increases the negarive temperature effect
of the moderator due to blocking of the absorber as temperature is increased.
Fast circulation of fuel in the reactor determines the particulars of 135Xe
poisoning due to entrainment of the irradiated fuel from the reactor with a
large flux of thermal neutrons to the irradiator, where the thermal neutron
flux is near zero with multiple repetition of the process.
A peculiarity of 135Xe poisoning of the reactor during fuel circulation can
be discovered by solving the following system of equations:
aPi/01 -I rJIiI/Jz - Q)TSfTu'I - x iPi:
r7p~e/df t~dp~c!Jz iDtE fill''X, + 7�~('~ ~~~~�~'S~� - ~Prl'~r^X~�~
where P1=pI(t, z); PXe - PXe(tl z) are the concentrations of 135I and 135Xe
at time t at point z of the circuit; it is convenient to take as coordinate
z the distance from the upper level of the core to the given point of the
circuit in the direction of fuel movement; ')T(t, z) is thermal neutron flux
density, EJT(t, z) is the macroscopic thermal-neutron fission cross section
of the core, ai,Xge are the constants of radioactive decay of 135I and 135Xe;
WI, Wge are the yields of the corresponding products upon fission of 235U;
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Px.
0,04 tqs4,y h
s~ e a
0,02 8.
0 45 r/N
Fig. 8. Distribution of 13sXe con-
centration heightwise of reactor
with circulating fuel for differ-
ent circulation times:
where pXe, Ps are concentrations,
QcXe, acs are the microscopic ther-
mal-enutron absorptior, cross sections
for 135 Xe and 235U respectively
aXe is the macroscopic thermal-neutron
absorption cross section of 135Xe; v is
the velocity of fuel displacement during
circulation.
4
~
1
Fig. 9. Distribution of heat
release heightwise of the core:
1a--considering (16--disregarding)
nonuniformity of temperature dis-
tribution heightwise of the core
in the presence of fuel circulation
(Tu = 4.4 hr); 2--using the principle
of one-time passage of fuel elements
through the core
At any time t, the solution of the system of equations should be periodic
with respect to z with period ZK equal to the length of the fuel circulation
loap. Let us note that ir.stead of z we can introduce the variable
. :
T dz'/v,
rK
where t,, dz'/v is the time taken by the fuel to complete one cycle.
In the case where the time of circulation is close to the period of decay
of 13sXe, strong nonuniformity is observed in the distribution of xenon con-
certr.ation heightwise of the core (Ref. 8)y however, there is little change
in the average xenon cWncentration.
It is clear fron the distribution of heat release heightwise of the core
(Fig. 9) that wY.en fuel is circulated the distribution of heat release is
not very nonuniform (curves 1a and 16), whereas with one-time passage of fuel
through the core in the equilibrium state one observes strong distortion of
the distribution of heat release heightwise of the core toward an increase
at the beginning of the zone (curve 2), which is cond+icive to equalizing the
temperature of the core material.
REFERENCES
1. Aleksandrov, A. P., ATOMNAYA ENERGIYA, Vol 25, No 5, 1968, p 356.
2. Aleksandrov, A. P., Ponomarev-Stepnoy, N. N., "Nuclear Power and Technical
Progress" in: "Atomnaya energetika XX let." [Nuclear Power of the Twen-
tieth Century], Moscow, Atomizdat, 1974.
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3. AlekSandrov, A. P. et al., "Bystryye i teplovyye geliyevyye reaktory dlya
proizvodstva elektroenergii i vysokotemperaturnogo tepla" [Fast and Thermal
Helium Reactors for Producirig Electric Energy and High-Temperature Heat],
IAEA, Vienna, 1976, Vol 1.
4. "Sostoyaniye i perspektivy razvftiya rabot po VTGR v SSSR" [Current State
and Outlook for HTGR Research in USSR], report TC-109/3 at meeting of
Technical Committee on HTGR's, IAEA, Vienna, 12-14 December, 1977.
5. Ryabukhin, Yu. S., Breger, A. Kh., ATOMNAYA ENERGIYA, Vol 7, No 2, 1959,
p 129.
6. Breger, A. Kh. et al., "Osnovy radiatsionno-khimicheskogo apparatostroyeniya"
[Principles of Radiation-Chemical Equipment Making], Moscow, Atomizdat, 1967.
7. Way, K., Wigner, E., PHY;;. REV., Vol 73, 1948, p 1318.
8. Mayenshteyn, F., et al., "Gamma Rays Associated With Fission" in: "Trudy
Vtoroy mezhdunarodnoy konferentsii po mirnomu ispol'zovaniyu atomnoy energii.
T. 2. Izbrannyye daklady inostrannykh uchenykh" [Proceedings of Secand
International Conference on the Peaceful Use of Nuclear Power. Vol 2.
Selected Papers of Foreign Scientists], Moscow, Gosatomizdat, 1959.
9. Sakharov, V. N., Malofeyev, A. I., ATOMNAYA ENERGIYA, Vol 3, No 10, 1957,
p 334.
10. Bunney, L. R., Sam, D., NUCL. SCI. ENGNG, Vol 39, 1970, p 81.
11. Dosta, L., De Tourrecil, R., J. NUCL. ENERGY, Vol 26, 1972, p 431.
12. James, M. F., J. NUCL. ENERGY, Vol 23, 1969, p 517.
COPYRIGAT: Institut atomnoy energii im. I. V. Kurchatova, 1977
Atomizdat, 1979
6610
CSO: 8144/1052-A
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UDC 621.039
SOME PROBLEMS OF HEAT EXCHANGE AND HYDRODYNAMICS IN HTGR CORE COMPONENTS
(SURVEY) -
Moscuw ATOMNO-VODORODNAYA ENERGETIKA I TEKHNOLOGIYA in Russian No 2, 1979
(signed to press 8 Jun 79) pp 142-148
[Article by Yu. N. Kuznetsov and V. L. Lel'chuk]
[Text] The ?aper gives some r.esults of theoretical and
experiment:l studies of heat exchange anc: hydrodynamics
in the amLu]ar channels, rod bundles, and in channels with
permeable walls as applied to helium-cooled high-temperature
reactors.
An examination is made of inethods of calculating thermohy-
draulic processes in the primary circuit of HTGR's and GCFFc's.
Mathematical Modeling of Processes of Convective Heat Exchange in Rod Bundles
and Annular Channels. One of the authors (Kuznetsov) has been doing theo-
retical research on convective heat exchange in HTGR cores with rod fuel ele-
ments, in which the peculiarities are associated with comparatively low heat
transfer coefficient, complicated geometry of the channel, variability of
thermophysical properties of the coolant and large volinnetric flowrate.
A theoretical study has been done on patterns of convective heat exchange
in rod bundles and annular channels for arbitrary laws of change in the thermal
load lengthwise of the channel. Yu. N. Kuznetsov has developed a technique
that enables reconstruction of local values of heat exchange characteristics
in any channel cross sections and for any law of heat supply based on studying
Clie stabil.ized heat exchange far from the inlet with some specially selected
law of heat supply. First for annual channels under conditions of constant
thermophysical properties of the coolant and constant change in the heat load
oi both surfaces lengthwise of the channel (with constant heat load on the
perimeter), the author examined the principles of stabilizing heat exchange
with increasing disatance from the channel inlet, introducing a criter-ion
that characterizes the influence of heat load variability, and is proportional
to the logarithmic derivative of the function describing the change in heat
load lengthwise of the channel.
As the heat load described by functions of exponential type changes with in-
creasing length, the temperarure field of the coolant is stabilized, but
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differs from the temperature field on the section of the heat exchanger sta-
bi?ized with respect to length for the case of a constant load. These differ-
ences increase with increasing relative rate of change in the load, and they
characterizP the thermal inertia of tne flow. This enables us to reconstruct
the influenre function for the given heat ex.�.hange process, which also charac-
terizes the thermal inertia of the coolant flow. Mathematically, the process
of finding the influence function reduces to solving an ordinary differential
equation with unit boundary conditions in the complex plane, and to subsequent
determination of some irtegral of the real part of the solution.
The major difficulties arise in description of the velocity field and transport
coefficients of the turbulent flow of coolant. A model of turbulent coolant
flow in annular channels is proposed. Results of calculations of the charac-
teristics of turbulent flaws and heat exchange by this model are compared
with experimental data of various authors.
Values and generalizing formulas for the ir..fluence function G1 are obtained
for the temperature diEference between wall. and coolant over a wide range
of working parameters and geometric dimensi.ons. It is shown that the problem
of local heat exchar.ge characteristics in the case of longitudinal flow around
infinite bundles of rods that are not too close can be reduced to investigation
of heat exchange in the inner zone (between the rod surface and the line of
maximum velocity) of such an annular channel for which the line of maximum
velocity in some sense coincides with the line of maximum velocity of the
given rod bundle. This has enabled us to u.se the above uescribed results
for studying annular channels, giving relations for the influence function
with heat exchange in rod bundles.
The resultant formulas can be used for fair;ly exact calculations of local
characteristics of heat exchange in any channel cross section with any law
of variation in heat load lengthwise over a wide range of geometric and flow
parameters : for annular channel 0< rl /r2 5 1; 104 < Re < 106 ; 0. 7 5 Pr < 100;
for triangular or sqare bundle arrays 1.5 1 mm
35--Fraction kap ensuring the required equaliza-
tion of fuel channel pawer and predetermined effective neutron multiplication
factor ke01 1(without considering the action of the reactor control rods).
In large channel reactors, (kaP - 1) is 2-5 times greater than (knc- 1) [Ref. 21.
This teads to considerable underburning of uranium in the discharged peripheral
fuel assemblies (PFA), which must be extracted from the reactor at a burnup
that is lower than that of the extracted central fuel assemblies (CFA). To
increase the burnup of uranium in the PFA's, the peripheral region of the
reactor can be charged with uranium of higher enrichment. This is economically
advantageous since it reduces the fuel component of the cost of e.lectric energy
by several percent; however, it entails production of fuel assembl?es with
differing uranium enrichment, and the burnup of the discharged uranium is
still less than could be realized by using these fuel assemblies in the center.
Let us try to define the conditions under which the burnup of discharged uranium
could be increased by transposing PFA's with completion of burnup in the central
region of the reactor core. We will consider a two-zone channel reactor in
which fuel channel power is equalized by profiling multiplication properties,
i. e. k~P > k~c. Let us assume that the law of change in k,,, as a function
of uranium burnup P, i. e. k.(P), is lineai and the same for CFA's and PFA's
[Ref. 3].
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In the simplest mode of continuous recharging, the central and peripheral
regions of the reactor are supFlied independently with the same kinds of fuel
assemblies (Fig. 1, I), the uraniian burnup in the discharged CFA's and PFA's
being Po and P respectively, where Po> P. This mode does not provide for
transpositions and additional burnup of the fuel assemblies, and is the gener-
ally accepted mode of operaticn for channel reactors.
Another reloading mode can be imagined in which all fresh fuel assemblies
ai�e loaded into the peripheral region of the reactor, operated there up to
the same uranium burnup as in the first mode, and then transposed and addition-
_ ally burned in the center (see Fig. 1, II).. If operation of the reactor in
charge charge discharge charge discl}arge chairge
po_p
I
~
center periphery I center periphery center periphery
! ' I
'p
p
discharge discharge transposition transposition discharge
I II Er
Fig. 1. Some modes of reactor recharging: I--independent recharging of center
and periphery; Il--all fresh fuel assemblies loaded into the periphery and
all additionally burned in the center; III--all fresh fuel assemblies loaded
into the periphery, and some additionally burned in the center
such a mode is possible, then uranium burnup in the CFA will vary from P to
some level P' > P at which the transposed fuel assemblies will be extracted.
For continuous transpositions and linear dependence k.,,(P), the condition of
criticality for the same profiling of multiplication properties will be the
same average burnup of uranium in this region as in the first mode with inde-
pendent recharging of center and periphery, i. e.
- ~ ,
whence
l'' - 1'� I' .
Since P' > P,
1' t !'o/'' .
Conditions (1) and (2) are illustrated by Fig. 2.
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k~(P
knc
OD
a
P)
\ 0.
~ -
\ -b ~
i
po-o vo
knc - -
00 P
` I
I
- - - - ~~-i
Fig. 2. Burnup of discharged
uranium when the center is
Continuing our crnnparison of modes I and
II (see Fig. 1), let us assume for the sake
of simplic ity that the specific power of
the uranium in the CFA's and PFA's (q, MW/kg)
is the same and does not change with burnup
(P, MW-days/kg). Let the number of CFA's
be n, and the number of PFA's be An. Then
for mode I the expenditure of CFA's and PFA's
per day is
vo n(j/l'o; u = iirtqlL';
The average burnup of uranium discharged
from the reactor is
1'ore , P" 1 A I I ^ i
/I i /~l'p n - ~ ~rn )
' 1 ^ ~ p /
charged with fresh fuel assem- In mode II (if it is possible), the additional
blies (a) and with fuel assem- burnup of uranium acquired by the fuel assem-
blies with burnup P(b) blies after transposition from the peripheral
to the central region will be APC= Po- 2P
(s Fig. 2b). Since the central region in this mode is continuously supplied
only with fuel assemblies that have been depleted on the periphery, their
expenditure in the central region must be equal to the expenditure on the
periphery, i. e.
nq Anq
whence
p 1 {^:.':1 p"'
Using expressions (1) and (4), we find that the average burnup of uraniiun
discharged from the reactor in mode II is
. ,
1'ii.- - 1-1 11 1 .
~ t I 2A
1~~)
Comparing expressions (3) and (5), we readilY see_that mode II is feasible
and is more advantageous than mode I(i. e. PII> PI) when condition (2) is met.
Thus mode II, in which all fresh fuel assemblies are loaded into the periphery,
used there until reactivity is exhausted, and then all fuel assemblies are
additionally burned in the center, is advantageous only if the uranium burnup
in the PFA's is less than half the uranium burnup in the extracted CFA's in
mode I. In reactors of large dimensions, where neutron leakage is small,
such a condition is not realizable in practice since it leads to a dip in
neutron distribution in the center of the reactor core, and mode IT is infea-
sible.
However, condition (2) can easily be met if PFA's are broken down into two
groups, and rather than transposing all PFA's to the center, we transfer only
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the PFA's of the first group with relatively low burnup of uraniwa P1< Po/2,
while the remaining PFA's are used on the periphery up to some uranium burnup
P2 (see Fig. 1, III). Let us consider mode III also in comparison with mode
I and let us assume that the specific power of the uranium q in the CFA's
and PFA's is the same and does not change during burnup.
For mode III, using exactly the same fuel assembli2s, the burnup of uranium
discharged from the central region in accordance with expression (1) will be
PC= Pp - P1, (6)
where P1 is the burnup of uraniwn in PFA's of the `_irst group after utiliza-
tion on the periphery.
The additional burnup of uraniwn that these fuel assemblies acquire after
being transposed and used in the center will be (Pa- 2P1). Since in this
case the central region is being continuously made up only with PFA's of the
first group, the expenditure of fuel assemblies in the central region vc must
be exactly equal to the consumption of PFA's of the first group, i. e.
Uc= vi;
nq
Uc 1'0-'LI" ( (7)
t ~
vj= Anqni;
where al is the fraction of PFA's of the first group among all PFA's.
For a given P1, the value of al is determined from condition (7), and the
burnup of discharged uranium in PFA's of the first group is determined from
the condition of criticality of the peripheral region that simultaneously
ensures the required profiling of multiplication properties:
14 k., ( l') rlP k9P , (8)
where az is the fraction of PFA's of the second group, where al+a2= 1. For
linear dependence k.(P), criticality condition (8) has the quite simple form
ri I p t
Consumption of PFA's of the second group is
UZ ! .I ttq~,c.~.
1
The average burnup of uranium discharged from the reactor in mode III will be
VC~ I..
Using conditions (7), (9) and (10), PIII can be represented as
'.:P -
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A study of this expression shows that PIII has a maximum at ~I" .1. It is
important that in this case the burnup of uranium in discharged CFA's and
PFA's of the second group be the same, i. e. PZ = Pc = Po - P1 � In this case
, ;m:ix i AP
111 I1t 1 I A t
where Pi~> PI�
Thus the operating mode of a reactor in which all fresh fuel assemblies are
charged only into the periphery, used there, and then some of the PFA's with
optimum uraniian burnup are transposed to the center where they are additionally
burned, while the rest of the PFA's are used on the periphery until the reac-
tivity of this region is exhausted, is more advantageous than independent
reloading of the central and peripheral regions of the reactor.
TABLE 1
Relative increase in uraniinn burnup
with transposition and operation of some PFA's
(_�~Igl-_ in the c enter - t� 1(N), A
1'I'' I
_ -
I
I.n I II,!) I II,N 11,7 I U~Y~ I U,5 I 0,4 I u, ;
II~27
11,
11~',',11
I1,25
(1, '-'3
0,20
u,S
I5
1
1,24
1,'.'.I
1,ti
1,11
1,02
0,ri4
11,7
:i,'!1
l
:i, In
:S, 1 I
:9,111
'',?ili
'._2 ,li!
:.','_'ti
U'li
li,lii
6,1i5
I;,i>.`S
li,'J;
6,'215
~r,,9:;
5,41
4,73
As can be seen from Table l, for the given assumptions and the values of A
and P/Po encountered in practice, tlie proposed recharging method enables an
increase in burnup of the discharged uranium by several percent with a corre-
sponding reduction in the fuel component of cost of electric energy. To check
out the approximate results, a stricter two-dimensional model of the reactor
was used, enabling accounting for power and energy release of each channel.
For this purpose, the heterogeneous QUM-3-HEP program was used [Ref. 41 as
well as the VRM and VOR cunstant programs used in designing the RBMK [Ref. 51.
Let us note that the given calculations have qualitatively confirmed results
of approximate calculations, although they showed ].ess increase in uranium
burnup. For example at A= 0.36 and P/Pp= 0.78, burnup increase was -0.7%,
:is against 1.3% according to the data of Table 1. Refir.ement is mainly by
considering equalization of the neutron field with transpositions of fuel
assemblies, leading to some increase in neutron leakage, and also by con-
sideration of nonlinearity in the behavior of multiplication properties of
t}ie medium with uranium burnup.
Calculations showed one more advantage of the given method of reloading, which
shows up in additional equalization of energy distribution due to transposition
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of partly burned fuel assemblies rather than fresh ones into the center, i. e.
into the region with maximum neutron flux. If reloading is done on the worl.ing
reactor, equalization may be intensified by 13sXe in the transposed fuel
assemblies. This advantage becanes especially important when an economically
advantageous increase in uraniwn burnup by simply raising its initial enrich-
ment is impossible due to an inadmissible increase in the coefficient of non-
uniformity of energy distribution because of heat-transfer conditions. Such
a situation is typical of reactors in which the minimum necessary reserve
is provided up to the limiting power of the fuel channels, and in particular
for the RBMK-1500 and the RBMKP.
The model indicated above was used for calculating steady-state conditions
of recharging the block-section RBMICP-2400 reactor with nuclear superheating
of steam [Ref. 6, 71, in which the plan provides for the given method of load
changing. Consideration was taken not only of the configuration and makeup
of the core (Fig. 3), but also of the fact that the power of the superheated
channels must be approximately half that of the evaporative channels, and
the region of elevated neutron flux density of the evaporative zones is made
up of the central regions of these zones, whereas in the superhea ted zone
the superheated channels in rows bordering on the zone of the evaporative
channels make up the regions of high neutron flux density. The evaporative
~ �s
_
I w I CFA ~ Iw I
P .a
L zone_ SC zone .
Fig. 3. Structure of the left half of the RBMKP-2400 reactor
core. The broken line shows the boundaries of sections (the
core consists of eight identical evaporative channel (EC)
sections and four superheated channel (SC) sections. Each
section contains 270 channels: 240 fuel channels and 30 for
control rods)
fuel assemblies are reloaded and transposed during operation of the reactor,
all Eresh fuel assemblies heing installed only in three peripheral evaporative
channel zones, from which some are transposed to the center. The superheated
fuel assemblies are reloaded and transposed with the reactor shut down, about
5% of the channels being renewed at one time. In the superheated zone, fuel
ussemblies are transposed from the three peripheral rows only to the three
rows of superheated channels bordering on the evaporative channel zones (Fig.
3). No transposition of fuel assemblies is provided in the other regions
of the superheated zone.
To simplify spatial calculations, a fixed periodic structure of the reactor
core was assumed. Certain average values of uranium burnup were assigned
to the fuel assemblies transposed aad reloaded from each region. In each
89
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region the distribution of the remaining fuel elements with respect to ura-
nium burnup from the fresh or transposed assembly to the unit about to be
discha;Ked or transposed was taken as discrete and uniform. Compensation of
operative reserve reactivity was also fixed by control rods. All this enabled
us to do variant calculations with minimum changes in load charting records
without considering the history of each fuel assembly. Criticality and the
required energy distribution in the reactor were ensured by varying the sought
average values af uranium burnup in the fuel assemblies at discharge or trans-
position.
TABI.E 2
Some characteristics of the RBMKP-2400 fuel cycle
for different methods of i�:loading fuel assemblies
I Without transposing I With transposition
Characteristic fuel assemblies of fuel assemblies
SC
Enrichment of uranium to be
2.0
3.0
2.4 3.6
loaded, %
Average burnup of discharged
21.9
24.7
27.7 30.5
uraniwn, MW-dayE /kg
Specific consumption of en-
28.8
6.4
22.9 4.8
riched uranium, metric tons/GW-yr*
Specific constunption of natural
108.6
37.9
106 35
uranium required for getting ura-
nium oE appropriate enrichment,
metric tons/GW-yr*
Fuel component cost of electric
1
.0
0.91
energy, relative units
*Power utilization factor was 0.8.
Calculations done at the same values of limiting power of the evaporative and
superheated channels showed that the proposed transpositions of fuel assemblies
into the region of elevated neutron flux density ean appreciably increase
the enriclunent and burnup of ur