SOVIET ATOMIC ENERGY VOL. 43, NO. 5

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• Publication Date: May 1, 1977
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SOVIET
ATOMIC
ENERGY
ATOMHAH 3HEPrwFI
(ATOMNAYA ENERGIYA)
TRANSLATED I FROM RUSSIAN
Russian Original, Vol. 43, No. 5, November, 1977
C CONSULTANTS BUREAU, NEWYORK
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SOVIET ' F'
ATOMIC
ENERGY,
Soviet Atomic Energy is abstracted or,in-
dexed in Applied Mechanics Reviews? Chem
ical Abstracts, Engineering Index, INSPEC-
Physics Abstracts and Electrical and Elec-
tronics Abstracts, Current Contents, and
-Nuclear Science Abstracts.
Soviet Atomic Energy is a cover-to-cover translation of Atomnaya
Energiya, a publication of the Academy of Sciences of the USSR.
,An agreement with the Copyright Agency of the USSR (VAAP)
makes available both advance copies of the Russian,journal and
original glossy photographs and artwork. This serves to decrease
the necessary time lag between publication of the original and
publication of the translation and helps to improve the quality
of the latter. The translation began"with the first issue of the
Russian journal.
Editorial Board of Atomnaya r nergiya:
Editor: O. D., Kazachkovskii
Associate Editor: N. A. Vlasov'
A. A. Bochvar
N. A. Dollezhal'
V. S. Fursov
I.'N. Golovin
V. F. Kalinin
A. K. Krasin
V. V. Matveev
M. G. Meshcheryakov
V. B. Shevchen'ko
V. I. Smirnov
A. P1 Zefirov-,
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
May, 1978
Volume 43, Number 5
November, 1977
CONTENTS
Engl./Russ.
ARTICLES
Soviet Atomic Science and Engineering Today ................................ ..... 971 323
Water-Cooled-Water-Moderated Reactors in Nuclear Power Generation of the Soviet
Union - V. A. Sidorenko ............. .......................................
Development of Uranium-Graphite Channel Reactors in the USSR - A. P. Aleksandrov
Program and State of Work on Fast Reactors in the Soviet Union - 0. D. Kazachkovskii,
A. G. Meshkov, F. M. Mitenkov, K. T. Vasilenko, V. M. Gryazev, G. V. Kiselev,
L. A. Kochetkov, V. B. Lytkin, V. V. Pakhomov, N. I. Savin, M. F. Troyanov,
V. I. Shiryaev, V. A. Tsykanov, and D. S. Yurchenko ..........................
992
343
Contemporary State of Research on Controlled Thermonuclear Fusion
- V. A. Chuyanov. ..........................................................
1000
351
Basic Approaches to Nuclear Power Station Safety in the USSR - V. A. Sidorenko,
0. M. Kovalevich, A. Ya. Kramerov, and Yu. E. Bagdasarov ...................
1009
360
Economic Aspects of the Development of Nuclear Power and Fuel Cycle Plants in the
USSR - N. P. Dergachev, A. K. Kruglov, V. M. Sedov, and S. V. Shuklin .......
1014
365
Nuclear-Energy Complexes and the Economic and Ecological Problems of Nuclear Power
Development - N. A. Dollezhal', V. N. Bobolovich, I. Ya. Emel'yanov,
A. I. Churin, Yu. I. Koryakin, A. S. Kochenov, A. Ya. Stolyarevskii,
V. A. Chernyaev, N. N. Ponomarev-Stepnoi, and A. M. Protsenko ...............
1019
369
Atomic Energy and the Environment - E. I. Vorob'ev, L. A. Il'in, V. A. Knizhnikov,
D. I. Gusev, and R. M. Barkhudarov ..........................................
1025
374
Nuclear Instrument Making, the Measuring-Informational Basis of Atomic Science and
Engineering - V. V. Matveev, I. S. Krasheninnikov, I. D. Murin, and K. N. Stas' .....
1034
384
Development of Radiation Techniques and Technology in the Soviet Union - A. S. Shtan'
and E. R. Kartashev .........................................................
1043
390
Collaboration between COMECON Members in Power Reactor Design, Including Some
Aspects of Nuclear Fuel Cycles - I. Barbur, A. Barchenkov, L. Molnar,
A. Panasenkov, V. Tolpygo, V. Khake, and B. Shcherbinin ......................
1053
402
Technological Aspects of Manufacture of Fuel for Different Power Reactors
- F. G. Reshetnikov, P. M. Verkhovykh, R. B. Kotelnikov, V. I. Kushakovskii,
and V. M. Makarov ..........................................................
1059
408
Plutonium Fuel and Fuel Elements for Power Reactors - I. S. Golovnin,
A. S. Zaimovskii, T. S. Men'shikova, N. P. Agapova, Yu. K. Bibilashvili,
V. A. Tsykanov, E. F. Davydov, V. M. Gryazev, V. I. Kuz'min, V.I. Syuzev,
V. M. Sedov, and N. P. Dergachev ............................................
1063
412
INFORMATION
Nuclear Power Station Construction in the USSR - L. V. Timofeev .... :...............
1069
418
Atommash Construction - L. V. Timofeev ..........................................
1072
420
The Russian press date (podpisano k pechati) of this issue was 10/ 25/ 1977.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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ARTICLES
SOVIET.ATOMIC SCIENCE AND ENGINEERING TODAY
The creators of Soviet atomic science and engineering, i. e. , scientists, engineers, and workers, are
proud of their successes as they celebrate the 60th anniversary of the Great October Socialist Revolution
together with the entire Soviet nation. Having designed and constructed the world's first atomic power plant
in a record-breaking time, the Soviet people opened up an era for the growth of nuclear power engineering.
In this achievement by the Soviet Union the international community saw an expression of its hopes for the
use of atomic energy for the good of mankind.
Large atomic power plants have been built and are in successful operation in our country, making it
possible to satisfy the growing energy demands of a number of important industrial regions of.the country.
At the same time, the most promising types of reactors are being studied and chosen for atomic power plants.
The Novovoronezh, Kolsk, and Armyansk atomic power plants have been provided with water-cooled-water-
moderated (VVER) reactors, the Leningrad, Kursk, and Chernobylsk plants, with RBMK uranium-graphite
channel-type reactors; and the Shevchenko atomic power plant, with a BN-350 sodium-cooled fast reactor
which, in addition to electricity, also generates industrial heat for distilling seawater. The Siberian, Belo-
yarsk, and Ulyanovsk atomic power plants are operating successfully. The Bilibinsk atomic heat and elec-
tric power plant with a uranium-graphite channel-type reactor built especially for the conditions of the Far
North is in operation in Chukotka. Atomic power plants have been built in Czechoslovakia, the German
Democratic Republic, Bulgaria, and Finland with the assistance of the Soviet Union.
The following large,atomic power plants with reactor blocks of 1 million kW or more are now under
construction in the Soviet Union: Smolensk, Kalinin, Rovno, South Ukrainian, Ignalin,.and others.
Two years after the first atomic power plant was put into service, construction was started on the
atomic icebreaker Lenin, .the flagship of the Soviet arctic fleet. The second nuclear-powered ship, Arktika,
is now in service and the physical start-up of the reactors of the Sibir is..now under way. In the year of the
60th anniversary of the October Revolution, the Arktika completed a heroic voyage, viz. , overcoming the
ice of the ocean it reached the geographical North Pole.
Soviet scientists and engineers are confidently advancing along the road to harnessed thermonuclear
energy. A developed thermonuclear reaction has been achieved in the Tokamak-l0. Achievement of a con-
trolled thermonuclear reaction on a commercial scale holds out great promise for providing energy for
future society.
Practical results have been obtained in the direct conversion of heat generated in a nuclear reactor.
The experimental thermoelectric and thermionic converters, Romashka and Topaz, have been demonstrated
in successful operation. Radioisotopic energy sources have been developed for use in meteorology and navi-
gation.
The application of atomic technology in the national economy has not been confined to power generation.
In thousands of plants it is employed in the control of technological processes and in quality control of products;
in geology, it is used in prospecting for commercial mineral deposits; in medicine, it is used extensively for
diagnosis and treatment; and in biology and agriculture, it is employed to accelerate the 'selection of micro-
organisms and plants.
The level of development attained by the atomic industry has made it possible to assign it the major
national economic tasks formulated and adopted by the Twenty-Fifth Congress of the Communist Party of the
Soviet Union (CPSU). The Main Directives for the Development of the National Economy in 1976-1980 envi-
sage the commissioning of 13-15 million kW of capacity in atomic power plants; to continue the construction
of atomic power plants with reactors each of 1-1.5 million kW; to provide for the leading development of the
nuclear power industry in the European part of the USSR; to speed up the construction and utilization of fast
reactors; and to proceed with preparatory work on the use of atomic energy for central-heating purposes.
Translated from Atomnaya Energiya, Vol. 43, No. 5, pp. 323-324, November, 1977.
0038-531X/77/4305-0971 $07.50 ?1978 Plenum Publishing Corporation 971
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The atomic industry is a characteristic example of the accelerated assimilation of scientific develop-
ments: from scientific search through basic research and technical development to the practical application
of the results. Many institutes of the Academy of Sciences of the USSR were drawn into the work on the
creation of the atomic industry, the most prominent industrial organizers were put in charge, special bran-
ches of scientific-industrial organizations were set up, factories with their own personnel were brought in,
and a considerable part of the work, from search to introduction into practice, was conducted simultaneously.
The experience gained from the development of the atomic industry demonstrated the positive aspects of such
organization and this found expression in the simplification and acceleration of the process by which scienti-
fic developments are introduced into practice as well as timely, high-grade training for personnel for a new
branch of the national economy. With the constant control and assistance of the Central Committee of the
CPSU, the atomic industry in a very short time became capable of solving problems of immense complexity.
This proved possible because throughout the existence of the Soviet state the party and the government have
paid constant attention to the organization and development of scientific establishments and leading branches
of industry.
As far back as during the first five-year plans an experimental base that was advanced for that time
was established for nuclear-physical research. Subsequently, this base was expanded through the construc-
tion of larger and larger facilities, including the unique accelerators at Dubna and Serpukhov.
Soviet scientists have enriched atomic science with major achievements such as: the discovery of spon-
taneous fission of uranium; hypotheses about the neutron-proton structure of the nucleus and the exchange
character of nuclear forces; the discovery of nuclear isomerism in artificial isotopes; the discovery of
Cherenkov-Vavilov radiation; research on artificial radioactivity; the development of a theory and methods of
calculation for the chain reaction of 235U; the discovery of new transuranium elements, including kurchatovium;
the discovery of new nuclear particles; measurement and refinement of the fission cross sections of nuclei
for thermal and fast neutrons; and many others.. Nuclear physics, which even in the 1930s seemed to be abs-
truse, abstract, and far removed from practical application, has grown into an important independent branch
of science, has had a great impact on other branches of science and engineering, and has laid the foundations
of nuclear power engineering.
Soviet scientists have played a leading role in the development of the concept of fast reactors possess-
ing the property of breeding nuclear fuel and thus increasing the efficiency of uranium utilization tens of
times. Such reactors are now in the stage of experimental-commercial development. The experience gained
from the operation of the world's largest fast reactor in the USSR convincingly confirms the promise of this
line of development.
The accelerated development of nuclear power generation today is one of the principal lines of work in
the rational balancing of fuel and energy resources, especially in the European part of the USSR. It can be
said with full justification that the advent of nuclear power has been extremely timely: mankind is facing the
problem of a shortage of fuel and energy resources. Nuclear power postpones the problem and, in the long
range, after the introduction of fast breeder reactors and mastery of thermonuclear power, will practically
completely eliminate this problem. Along with the optimal solution of the fuel problem, nuclear power also
facilitates protection of the environment.
Electricity can already be obtained from atomic power plants at lower cost than from power plants oper-
ating on organic fuel. In the future, the economy of atomic power plants will increase further, on the one
hand, because of the continuous improvements in their design and, on the other hand, as the result of the
rising cost of organic fuel as more and more inaccessible and remote deposits are developed.
The application of atomic energy for central heating is an important immediate task. Later on methods
will have to be developed for obtaining high-temperature heat in nuclear reactors for industrial purposes,
for metallurgical, chemical, and other production.
Science and the results of its development are inseparable from the policy of the party and the govern-
ment: science has become a direct productive force. Fruitful basic and applied research in the realm of nuc-
lear science and technology is being pursued in many scientific establishments provided with the latest word
in equipment. Successes are facilitated by the consolidation and coordination of work done in this field in the
socialist countries. "We consider it necessary, " Comrade L. I. Brezhnev said at a meeting with the leading
officials of the Academies of Sciences of the socialist countries, "to encourage the development of basic sci-
ence in every way, to see to its being organically united with applied research, and to speed up the introduction
of scientific discoveries into the national economy" [Kommunist, No. 4, 9 (1977)].
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The continuous care of the party and the government for Soviet nuclear-physical science and atomic
industry and the selfless work of scientists, engineers, and other specialists have ensured decisive achieve-
ments in this field and laid the foundation for future major successes in the utilization.of atomic energy for
the development of the material and technical base of Communism.
WATER-COOLED-WATER-MODERATED REACTORS IN
NUCLEAR POWER GENERATION OF THE SOVIET UNION
The prospects of a new trend in nuclear power generation must be based first and foremost on its
economy. Other important requirements for a nuclear power plant as a source of energy in electric power
stations are safety and reliability. But it is obvious that the achievement of the required level of safety and
reliability appears also in the final count in the efficiency indices of nuclear power stations. The first Soviet
water-cooled-water-moderated (VVER) reactors demonstrated the validity of the scientific-technological
principles invested in them, and the structural design principles for achieving them. The high efficiency of
water as a moderator and the small difference in the moderation length in the fuel lattice and in pure water,
because of the significant contribution of inelastic moderation in the uranium, have permitted the use exclu-
sively of a compact fuel lattice design, and the production of a high power from a reactor with a small-sized
core.
Close-packed fuel lattices ensure a significant cross effect of fast neutron multiplication. A compact
core design with the minimum quantity of structural materials allows a comparatively good thermal neutron
utilization to be achieved. These qualities enable the required neutron multiplication factors in the fuel
lattice to be obtained, with a quite high resonance capture in 238U. The neutron spectrum in the close-packed
lattices used is found to be.relatively hard, and fission and absorption processes in the epithermal energy re-
gion acquire considerable importance. Thus, a high rate of plutonium buildup and its significant contribution
to the production of energy for a deep fuel burnup are ensured. As a result, the conversion factor of fissile
isotopes in a water-cooled-water-moderated reactor is found to be markedly better than, e. g. , in a similar
graphite-water reactor.
The choice of sintered uranium dioxide as the fuel has proved to be promising, since it is stable to
erosion by water (coolant) and, for this reason, ensures the least contamination of the primary circuit in
the event of fuel element failure, and fuel elements of uranium dioxide maintain their efficiency with quite
deep burnups, which promotes further, improvement of the fuel cycle of these reactors. However, the use of
uranium dioxide as the fissile material is not optimal from the point of view of the.physics of the fuel cycle,
but is a compromise, partially satisfying the physical and technical requirements. Conversion in the future,
with the improvement of nuclear fuel technology, to denser compositions based on uranium metal, shows pro-
mise for the additional improvement of the economic characteristics of the fuel cycle.
A special zirconium alloy is used as the principal structural material of the core, which allows quite
high coolant parameters to be ensured and a satisfactory thermodynamic efficiency of the cooling cycle to be
achieved. At the same time, a quite favorable neutron balance in the core has been successfully maintained.
The use of water-cooled-water-moderated reactors in nuclear power stations has led to the introduction
into power generation of turbogenerators on saturated steam. This decision has proved to be important in
principle, since it has permitted a satisfactory efficiency of the cycle (27-34%) to be achieved in nuclear power
stations with a limited coolant temperature in the reactor (300-350?C).
The deciding factors that stipulated the power of the first commercial water-cooled-water-moderated
power reactor in the first unit of the Novovoronezh nuclear power station (NVNPS) (Fig. 1) as 210 MW (elec-
trical) were the dimensions of the reactor vessel (external diameter, 3.8 m; length, -12 m), which are almost
limiting with respect to the conditions of manufacture of the vessel at the factory and its transportation by rail-
way. The requirement for the railway transportation of the reactor and other plants of VV ER reactor installa-
tions is to be decided by the future development and improvement of technical characteristics.
Translated from Atomnaya Energiya, Vol. 43, No. 5, pp. 325-336, November, 1977.
0038-531X/77/4305-0973 $07.50 ?1978 Plenum Publishing Corporation
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Fig. 1. Fiftieth-Anniversary-of-the-USSR Novovoronezh nuclear power station. The
first Soviet nuclear power station with power reactors of the containment-vessel type.
Four units are currently operating with a total nominal capacity of 1455 MW.
A high-strength low-alloy steel is used for the reactor vessels which ensures its minimum dimensions and
mass.
The first unit of the Novovoronezh nuclear power station came into operation in Sept. 1964 (in the startup
year this unit was the most powerful nuclear power station in the world).
The next qualitative stage (second generation) of development of the W ER is the W ER-440 facility. It
forms the basis of the first large-scale series of nuclear power stations, as their satisfactory economic in-
dices have made these stations completely competitive with conventional-fueled stations in almost all the re-
gions of the European part of the Soviet Union. These'reactors are extensively used also in certain foreign
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TABLE 1. Principal Characteristics of Nuclear Power Stations with Water-Cooled-
Water-Moderated Reactors
VVER-365
VVER-440
VVER-1000
Year of startup
Capaci
electrty, MicalW:
3 x 70
5x73
2 x 220
2 x 500
thermal
760
1320
1375
3000
Efficiency (net),%
27,6
27,6
32
33
Steam pressure ?Defore turbine, kgf/cm2
29
29
44
60
Pressure in primary circuit, kgf/cm2
100
105
125
160
No. of loops
6
8
6
4
Water flow rate through reactor, m3/h
36500
49500
39000
76000
Internal diameter of reactor vessel, mm
3560
3560
3560
4070
Steam production from one steam generator, tons/h
230
325
425
1469
Uranium charge, tons
42
66
Average fuel burnup in stationary cycle, MW
28,6
26-40
day/ kg U
Av. specific power intensity of core, kW/liter
111
Av. power intensity of fuel, kW/kg U
33
45,5
Specific flow rate of coolant, tons/h? MW
38
30
25
19
Water temp. at reactor inlet, ?C
250
250
269
289
Av. heating in reactor, C
19
25
31
35
Specific capital costs, " rubles/kW (electrical)
406
273
200
Cost of electric powert, kopecks/kW. h
0,95
0,743
0,643.
(0,788)
(0,569)
(0,584)
* Data according to the Novovoronezh nuclear power station.
t Design data according to the units of the Novovoronezh nuclear power station. Actual data during 1976 are
shown in parentheses.
TABLE 2. Technicoeconomic Character-
istics of the Novovoronezh Nuclear Power
Station
Electric
power output
kWh?106
Utilization fac-
tor of installed
capacity
Cost of electric
power sent out,
kopecks/kW h
1972
5413,4
0,607
0,81
1973
8647,7
0,68
0,752
1974
9664,1
0,76
0,644
1975
9138,1
0,717
0,641
1976
9750,8
0,763
0,632
countries; they are operating and continue to be constructed in the German Democratic Republic, in Bulgaria,
Finland, and are being constructed in Czechoslovakia, Hungary, and other countries.
The third generation is the VVER-1000, which is being constructed at the Novovoronezh nuclear power
station as the fifth unit and is the pilot reactor in a new series.
The reactor facility of the second unit of the Novovoronezh nuclear power station, the VVER-365 (Table
1), occupies an intermediate position between the first and second generations. All the principal improve-
ments of the core, developed for a commercial reactor of average capacity and then used in the VVER-440,
have been incorporated in it. In order not to delay the practical confirmation of these decisions, the nuclear
power station and the reactor facility were completed mainly with the plant developed for the first unit and
calculated on almost the same parameters. The operating characteristics of the Novovoronezh nuclear power
station are given in Table 2.
In order to solve the problem associated with the special features of the physical processes of VVER,
an experimental base was constructed, extensive experimental data were accumulated, and on this basis im-
proved computational programs were worked out.
The small neutron diffusion lengths predetermine the sharp nonuniformity of the power distribution near
inhomogeneities of the fuel lattices. The large temperature and power effects of reactivity, the large excess
of reactivity on- burnup, require high efficiencies for the control systems. The possibilities of significant
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Fig. 2. Recorder chart of the first fuel charges of the reactor of the first unit at
the "Kozlodu" (Bulgarian Peoples' Republic) nuclear power station, and the second
unit of the "Nord" (German Democratic Republic) nuclear power station. 1) En-
richment 1.6%; 102 working cassettes and 12 control and safety cassettes; 2) enrich-
ment 2.4%; 108 working cassettes and 25 control and safety cassettes; 3) enrichment
3.6%; 102 working cassettes and no control and safety cassettes.
0 20 40 60 80 100 120 140 160 180 200 220 240 260 280
T, eff. days
Fig. 3. Change of critical concentration of boron in the coolant during
burnup of the first fuel charge of the Novovoronezh nuclear power sta-
tion fourth unit.
deformations of the power distribution during burnup create the need for the development and use of a complex
scheme of interchanging the cassettes in the core during rechargings, and the use of fuel charges with nonuni-
form composition. All this, in conjunction with the high power intensity of the core, requires great detail and
a high accuracy of the neutron-physics calculations. The grouping of part of the core of the VVER-440 is
shown in Fig. 2. In the I. V. Kurchatov Institute of Atomic Energy, a large system of mathematical programs
was set up for investigating VVER reactors. The system consists of six interrelated complexes. The first
and second complexes represent the fundamental libraries of computed nuclear data and program systems for
the preparation of multigroup libraries of cross sections, worked out both in the Institute of Atomic Energy
and in the framework of the Temporary International Scientific-Research Staff of Scientists of the Member-
Countries of the CMEA. The libraries of nuclear data of the Soviet Nuclear Data Center and other countries
have been used.
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1,50 i
a 0,50
Enrichment of cassette, ?/?
1,6 2,4 1,6 1,4 2,4 16 2,4 1,6 3,6 2,4 16 24 3,6 16 16 2,4 3,6 2,4
24 2,4 1,6 2,k 3,6 2,4 2,4 2,4 1,6 36 1,6 2,4 3,6 2,4 1,6 5,5 2,4 3,6 3,6
0, 25 1 2 3 4 5 6 7 8 9 10 12 1314 15 16 17 18 19 20 24 25 25 2728 29 30 33 37 38 39 40 48 49 50 51 5f 62
No. of cassettes
Fig. 4. Power distribution over the cassettes of the first fuel charge of the reactor
of the first unit of the Kolsk nuclear power station: ) calculation by the BIPR-4
program; A) thermocouple readings; Np = 100%; T = 100 eff. days; HARK-VI = 150 cm;
CHZBO3 = 3 g/liter; six operating primary circulation pumps.
0? 0?
VO 00 0
0 0
0
0
0 0 o000
Fig. 5. Deviation of calculated distribution
of power over the cassettes of the core, from
the measured distribution (first charge of the
reactor of the second unit of the "Kozludu"
nuclear power station with 100% capacity and
charge duration of 80 eff. days): 0) experi-
ment.
The third and fourth programs are complexes for the preparation of libraries of small-group cross
sections and precision multigroup programs (one-dimensional and two-dimensional in PN-approximation)
and Monte Carlo programs (one-, two-, and three-dimensional) for detailed investigations of procedural
problems.
The fifth complex comprises programs for design calculations on the physics of VVER. Here, pro-
grams on the preparation of effective cross sections of the lattices in different states of burnup, with a differ-
ent isotopic composition, etc. (100 points with respect to energy in the thermal group, 3 in the epithermal
group, and 120 chains of fission products and heavy isotopes) are coordinated: programs for calculating boun-
dary conditions, spectrum details in the epithermal energy region, and for calculating the power distribution
inside the cassettes; one- and two-group three-dimensional physics model of the VVER with the reproduction
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03800
0
P4-
X3840 04500
Fig. 6. VVER development: 1) VVER-210 (mass of vessel and reactor: 223 and
470 tons); 2) VVER-365 (241, 523 tons); 3) VVER-440 (200, 573 tons); 4) VVER-
1000 (304, 730 tons).
VVER-1000
VVER-365
VVER-440
2
VVER-210
1375
2500
3000
760
950
1320
MW
MW
MW
MW
MW
a
b
Fig. 7. Role of the different factors in increasing the thermal ca-
pacity of the VVER in the second (a) and third (b) generations: 1, 3)
Coolant flow rate, length of fuel elements (for 3, margins up to lim-
iting values); 2) nonuniformity of heat release; 4) design of pumps
(inertialess).
of all burnup control cycles of the fuel charges, taking into account the effects of feedback, etc.; a program
for calculating the reactivity coefficients and other parameters of the point kinetics of the reactor; two-dimen-
sional four-group 7000 program, making it possible to analyze the power distribution over all the fuel elements
of the reactor. The sixth program is concerned with programs for thermophysical calculations, calculations
of safety and reliability, optimization of fuel cycles and the processing of operating data, etc.
The constant improvement of the system of programs developed ensures the following computational ac-
curacy for the main operating characteristics of VVER, %:
Coeff. of power nonuniformity of the cassettes ............. 5
Coeff. of power nonuniformity of the fuel
elements inside the cassettes ........................... 10
Duration of operating period of charge .................... 5
Eff. of control,rods ..................................... 10
Eff. of liquid absorber .................................. 5
Temp. coeff. of reactivity ........ ................. 0.5.10-4/?C
Power coeff. of reactivity ............................... 10
These values permit the reliable functioning of all operating reactors and, with sufficient certainty,
predict the physical characteristics of reactors in the design stages.
With the bringing into operation of the variants of the program being developed at present, the errors
of the calculations are reduced to errors of measurement with regular measuring systems. Figures 3-5 show
the calculated and experimental data defining the accuracy of the computational programs.
The most important factor in the development of VVER is the increase of the unit capacity of a reactor
assembly. Figure 6 schematically shows the power series of VVER: 210, 365, 440, and 1000 MW (electrical).
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12
75 150 225
Costs of external fuel cycle, rubles/kg
6
Fig. 8. Earning capacity of the chemical reprocessing. of the VVER
spent fuel: 1, 3) ThM; 2, 3) U02; 4, 6) UM; for alternatives 3, 5,
and 6 the cost of UF6 and separative work units is 36 and 41 rubles/
kg; for alternatives 1, 2, and 4 it is 100 and 54 rubles/kg, respec-
tively.
TABLE 3. Principal Technicoeconomic In-
dices of Possible Fuel Cycles for the VVER-
100 (for (p = 0.8)
Dioxide
open
cycle
Natural uranium 216
consump., ikg/M W'.
(elect.) ;yr.
Separative work, kg 132
separative work units
MW(elect. )- yr
Metal
ioxidel Metal Ilic thor-
rum
The development of the VVER with increased unit capacity has taken place over two successive stages
of increase of the parameters, and has ensured a corresponding increase of the thermodynamic efficiency of
the steam power cycle.
The direct means for increasing the thermal capacity of the VVER are reduction of the nonuniformity
of heat release in the core, increase of the coolant flow rate through the core and of the total length and sur-
face area of the fuel elements; reduction of the margin between the operating and maximum permissible values
of the parameters. The increase of the reactor capacity due to the reduction of the nonuniformity of the heat
release was accomplished by conversion from the core of the first unit to the core of the second unit of the
Novovoronezh nuclear power station.
The VVER reactors are oriented mainly on a cycle of three partial rechargings per running period. The
design duration of the working period between rechargings, in the second and third generations of reactors,,
amounts to 6500-7000 eff. hours, which ensures a good utilization factor of the installed capacity of the station
and permits recharging to be carried out once per year, in the spring-summer period which is suitable for
power generation systems. In all the reactors, with the exception of the VVER-210, this recharging cycle
is used, in which the fresh fuel is always charged into the periphery of the core, with its subsequent reloca-
tion in the central region of the core (remaining there during two operating periods), resulting in its discharge..
This ensures the necessary compensation of the heat release in the core; moreover, the difference between
the average and maximum burnup of the discharged fuel is reduced. An additional reduction of the nonuni-
formity of the heat release was provided by the introduction on the second unit of the Novovoronezh nuclear
power station of a control of the burnup by the power, by means of a solution of boric acid in the primary
coolant.
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80
I 1 Lim 1
0 41 0,2 0.3 44 0,5 0,6 0,7
Running time, reL units
Fig. 9. Adjustability of a unit with VVER-440, taking account of
the presence of a controlling group of cassettes in the core (the
change of water temperature in the case of a power tripout is ta-
ken into account): -----) holding at power N and at any instant of
rise to nominal is ensured; -) holding at power N and output
at nominal power ensured, if the halt is not more than 1 h; 1, 2)
(I/CB) (dCB/dt) = 0.05 h-1 in the case of an unlimited and 1-h halt,
respectively; 3, 4) (1/CB) (dCB/dt) = 0.20 h-1 in the case of an
unlimited and 1-h halt, respectively.
An increase of the coolant flow rate requires the development of a new, more powerful pumping plant
(or requires the number of loops in the reactor installation to be increased, which is economically disadvan-
tageous). The flow rate of the water in the core and in the reactor vessel restricts the increase of the cool-
ant flow rate in the VVER.
The overall length of the fuel elements can be increased by increasing the total charge of uranium in the
core or by a reduction of the diameter of the fuel elements. The principle of the factory manufacture of the
reactor vessel and its railroad transportation limits the increase of the total uranium charge, and therefore
from the very start the necessity arose for developing fuel elements with a somewhat smaller diameter than
was accepted in foreign practice. The use of fuel elements of relatively small diameter ensured, in the first
stages of development of the VVER, a reserve in the linear thermal loading, which makes possible a consider-
able increase of the power intensity of the fuel in the core volume.
Three factors should be distinguished in the reduction of the margin between the operating and limiting
values of the parameters. The first is related to the depth, of our knowledge about the processes taking
place in the reactor. An example of the improvement of investigations and of the increase of the reliability
of the results is a study of a heat exchange crisis in the core. The decisive factor here is the maximum
approximation of the study conditions to the natural operating conditions.
In the I. V. Kurchatov Institute of Atomic Energy, the conditions of origination of a heat exchange crisis
in water cooled reactors have been studied fora longtime, and the results of other organizations as applicable
to the cores which have been developed have been generalized. Based on experiments conducted under con-
ditions as close as possible to the operating conditions of the VVER) a numerical formula was found for pre-
dicting a heat exchange crisis in the fuel element bundles of the VVER, with a mean-square error of 5.5% (for
uniform heat release). This formula satisfactorily describes the experimental data with respect to bundles
with nonuniform heat release. Further systematization of the experiments on the conditions of heat exchange
in the fuel element bundles creates a good basis for overloading of the thermal cycle of cores by disposing
of the margins in ignorance.
The second factor for reducing the margins is related to the reliability of the knowledge of the para-
meters achieved in the reactor, and conditioned by improvement of the measurement systems, in the first
phase,of intrareactor measurements.
The third factor is the increase of reliability of heat removal systems, which will enable the identical
effect to be produced in an increase of thermal capacity with a small increase of cost or a large increase of
capacity with a fixed increase of cost. This factor made the greatest contribution to the increase of capacity
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61218061218061218 0
12 18 0 6 12 18
100
80
oI
12 18 0 6 12 18 0 6 12 18 0 6 12 18 0 6 12 18 0
Friday Saturday Sunday Monday Tuesday
Hours of days
Fig. 10. Change of reactivity excess during operation of the re-
actor in accordance with a characteristic weekly load chart.
during conversion from the VVER-440 to the VVER-1000, due to the replacement of the glandless primary
circulation pumps, which have a low inertia, by pumps with external electric motors and with forced small
flows, equipped with special flywheels. Figure 7 shows the role of the,various factors in increasing the power
output.
Important factors, which ensure an increase of power, are increase of the fuel burnup and its charge in
the reactor core. It should be emphasized that railroad transportation of the reactor vessel facilitates the
construction of nuclear power stations in many regions of the Soviet union and other countries, but in practice
it limits the unit capacity of a VVER reactor to 1000 MW (electrical), as it is difficult to distribute a large
load of uranium. In practice, from the point of view of railroad transportation, the maximum dimensions of
the vessel have been reached already for the VVER-210.
In order to ensure the optimum grouping and structural solutions, an increase of the capacity of the unit
must be accompanied by an increase of the capacity and output of the main plant. The tendency to reduce the
specific capital costs in commercial installations and to utilize rationally the production capacities of factory-
manufacturers reduces to the fact that the tendency to enlargement of the plant acquires an independent nature
outside of the relation with increased capacity of the unit.
The. plant enlargement, unconditionally, requires a significant increase of its reliability and this, in its
turn, leads to a reconsideration of certain principles established in the basic layout of the facility, the group-
ing of the plant and maintenance methods. The experience built up on all VVER installations takes into account
this increase of reliability. and confirms the advisability of simplifying the layout decisions and the basic con-
figuration.
In the VVER-210, each circulation loop of the primary circuit is installed in an insulated compartment,
which allows maintenance of the loop facilities to be carried out during operation of the reactor with the other
loops. In the second unit of the Novovoronezh nuclear power station, two loops were installed in each compart-
ment, and in the VVER-440 the plant of all six loops was installed in a single compartment. A similar arrange-
ment of the four loops is provided in the VVER-1000.
In considering a capacity of - 2000 MW as a further step in construction of VVER reactors, several
routes can be mentioned for achieving this capacity. The thermal capacity of 6000-6300 MW necessary for this
unit can be obtained from the core installed in the vessel of the VVER-1000. For this, it will be necessary to
use fuel elements with a diameter of 6 mm, to carry out fuel recharging twice per year and, for the purpose of
maintaining the total feed of cooling water in the reactor, as accepted in the VVER-1000 design, it will be neces-
sary to generate saturated steam at a pressure of 47 kgf/cm2. An increase of the thermal load of the fuel ele-
ments can be provided by heat exchange intensification.
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b
Fig. 11 Fig. 12
Fig. 11. Grouping of a reactor for an automatic thermal power station: 1) absorber servo;
2) cooling for concrete; 3) intermediate heat exchanger; 4) reactor vessel; 5) tractive section;
6) cooling for iron water shield; 7) core; 8) concrete shell.
Fig. 12. Emergency localization systems for Pmax = 4 (a) and Pmax = 1 kgf/cm2 (b).
More preferable at the present time is the route which retains all the principle solutions with respect
to the core and the limiting parameters of the VVER-1000. For this unit it will be necessary to construct a
reactor vessel with an inside diameter of 5.7 m, in which a fuel charge of about 150 tons can be disposed
(scaled to uranium metal); new and more powerful steam generators must be developed. By retaining the
VVER-1000 pumps, a layout is possible with two pumps per steam generator, and by retaining the layout with
4 main circulating pumps, pumps with an output of ^- 40,000 m3/h will have to be built.
The discarding of railroad transportation of the reactor vessel leads to the necessity for using either
other means of transportation or undertaking assembly of the steel vessel in situ at the nuclear power station,
for which multilayer vessels are more suited, as they do not require heat treatment after assembly. The in
situ construction of reactor vessels is possible not only of steel, but also of prestressed reinforced concrete,
and its use will remove completely the problem of power limitation of the VVER. The impossibility of large
cracks appearing in consequence of embrittlement failure makes the concrete reactor vessel safer than a
metal reactor vessel. However, in a complex with a concrete vessel, it is advisable to use a VVER of the
boiling type, as the pressure in it is twice as low as in the nonboiling type. In order to construct high-capa-
city boiling VVER, the large positive experience in the operation of the experimental VK-50 boiling reactor
in the Nuclear Reactor Scientific-Research Institute (NIIAR) has been useful. The large volume of investiga-
tions on this reactor has enabled a higher specific power density to be achieved with natural circulation in the
core, than in the reactor of the first unit of the Novovoronezh nuclear power station. For a boiling reactor
with a capacity of 2000 MW, a reactor vessel is required with a diameter of 12-14 m and a height of 21-23 m,
and with a concrete wall thickness of - 5 m. Steam generators and other complex plants are not required in this case.
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Together with the solution of the problem of further increasing the unit capacity of the nuclear power
station reactors, the problem of improving the efficiency indices of the fuel and reducing the specific consump-
tion of natural uranium will be urgent for a long time. Up to the present time, the principal route for im-
proving the efficiency of the fuel cycle of VVER reactors is to increase the fuel burnup. In the reactors of
the second unit of the Novovoronezh nuclear power station, and in the VVER-440, the design burnup of 28,000
MW-days/ton already has been achieved on the average for the discharged fuel, with a maximum burnup of
>40,000 MW. days/ton for the fuel elements.
The core of the VVER-1000 is oriented on an average burnup of 40,000 MW-days/ton and with a maxi-
mum burnup on the average of 44,000 MW?days/ton for the fuel element; for this, it is necessary to supply
fuel enriched to 4.4% on recharging. Until mass experience in the operation of fuel at this burnup has been
acquired, the core can operate with makeup fuel enriched to 3.3%, with a burnup of 27,000 MW- days/ton.
The currently used cycle of three partial rechargings with an annual cycle cannot be assumed to be the
best for the future development of VVER reactors. In the first place, an increase in the number of partial
rechargings per running period of the fuel from three to six will allow the fuel component cost of electric
power to be reduced further by 6-8%; secondly, in the development of large-scale nuclear power generation,
it is not obligatory and even inconvenient to "be bound" to recharging to a summer minimum of power require-
ment. The most important requirement remains the provision of the maximum load factor of the station, in
any case, to reduce to a minimum the time lost. On the other hand, the duration of the recharging operations,
even for those reactor designs which are calculated on once-per-year recharging, including cooling and heat-
ing up, unsealing and resealing the reactor, all assembly and disassembly operations and replacement of
spent cassettes with the recharging machine, amounts to about 7-8 days.
This experience allows it to be Supposed that the most promising route for increasing the recharging
frequency is the improvement and simplification of the existing methods with a reduction, as the immediate
problem, of the duration of shutdown of the unit for the next fuel recharging to 1 week. Under these condi-
tions, the changeover of all VVER reactors to a regime with six partial rechargings per running period with
a 6-month cycle can be planned. In relation to the specific consumption of natural uranium in the stationary
regime, this will be equivalent in practice to a limiting regime of continuous fuel recharging. There are other
ways also for improving the fuel cycle factors of the VVER.
As the irradiated VVER fuel contains a large number of fissile isotopes, chemical reprocessing of the
fuel and the return of the reprocessed fuel to the cycle permit the consumption of natural uranium to be re-
duced by 40-50%, and the necessary capacities of the separating factories by 40%.
The development and utilization in the cores of VVER of a denser fuel than uranium dioxide (e.g., cor-
rosion-resistant compositions based on metallic uranium), would permit additionally a reduction of the con-
sumption of natural uranium by 35%, the capacity of the separation plant by 50%, andthe fuel component by 15%.
The data on the use of thorium fuel cycles in VVER (Table 3) are close to these results.
Figure 8 shows the variation range of the cost of production of the fuel cycle, for which the expediency
of chemical reprocessing of VVER fuel is retained, for a different cost of natural uranium and its enrichment
process.
The introduction into the power generation system of a country of a large number of nuclear power sta-
tions requires a reconsideration of their operating cycles. If, until very recently, work on base loads could
be provided for nuclear power stations, then in subsequent years they would have to operate as well in the
variable part of the load chart. Experience of operating the VVER confirms the simplicity of control of re-
actors and the feasibility of following a variable load. The VVER reactors have a very important quality
- negative temperature and power coefficients of reactivity and, associated with this, the capability of auto-
matic regulation and automatic limitation of power. Special attention has always been paid to the conservation
of this property of VVER cores. In particular, after the introduction into the designs of reactor installations
of liquid boron control, the choice of the maximum reactivity excess, compensated by a dissolved absorber,
has determined the conservation in all operating cycles of a negative reactivity coefficient with respect to the
temperature of the coolant.
The dimensions of the VVER reactors (by comparison with the neutron migration length) are very large.
Under conditions of a strongly flattened neutron field, variations of the reactor power accompanying the move-
ment of the automatic control rods and the change of the spatial power distribution are dangerous due to the onset
of xenon instability. However, in the VVER the negative power coefficient of reactivity effectively stabilizes
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the system, significantly displaces the boundary of instability, and also reduces possible fluctuations of the
neutron field in transient processes. The VVER-440 reactor is completely free from xenon power "wobbles. "
The radial dimensions of the VVER-1000 also guarantee stability of the neutron field in the horizontal cross
section of the core. The increase of the height of the core to 3.5 m in the VVER-1000 required a special sys-
tem of control with the high-level field, which must preserve the axial nonuniformity of the heat release in
transient processes within the scope of the permissible values. of 1~"r
In solving the problem of building an adjustable power generating facility, in addition to the basic stabil-
ity of the reactor, other limiting factors are important, relative to which the units of nuclear power stations
with the VVER reactors are undoubtedly promising. Here should be mentioned the problems of ensuring the
necessary reactivity excess for overcoming the nonstationary poisoning by 135Xe, the deviations of the para-
meters of the facility from the nominal values, and problems associated with the duration and complexity of
the technological operations during changes of operating conditions.
The solution of the first problem in the existing VVER reactor installations, in principle, is simplified
by the large reactivity excess by burnup, which can be used in transition processes during the greater part of
the running period of the reactor, without specific deterioration of the planned fuel cycle of the present day
installations. As a rule, this is related with the creation of an adjustable reactivity excess (Fig. 9). Good
grounds for solving the second problem consist in the small range of change of operating temperature of the
primary coolant with a wide variation of loads (not more than 30?C for a 100% power variation), which is char-
acteristic for the VVER installations. Cyclic loads in the plant of the primary circuit are not found to be ex-
cessive, and the automatic adjustability of the reactor contributes to maintaining the parameters of the facility
within safety limits.
The good controllability of the VVER reactors is confirmed in practice; further improvement of the con-
trol system and of the technological circuits in new designs is directed at increasing the control operability of
the facility in the case of frequent changes of the operating conditions. Already in certain designs of the units
with commercial VVER-440 reactors, complex power change charts with weekly shutdowns and daily partial
load sheddings are being adopted (Fig. 10). Based on the VVER-1000 facility, a unit with a VVER-500 is being
constructed, on which increased demands for adjustability have been imposed. The specific problem requiring
special attention remains the assurance of the necessary durability of the fuel elements, constantly operating
under variable power conditions.
A new important field in which the experience in constructing VVER reactors can be used in future years
is district heating and heat supply. The first route to the solution of this problem is the construction of auto-
matic thermal power stations (ATETs) based on the developed reactor facilities used for nuclear power sta-
tions. Changes may involve only the steam-turbine section of the station and the extent to which solutions must be
considered, directed at increasing the radiation safety of facilities, in consequence of their proximity to dense-
ly populated regions. Here, in time, a predominance of boiling reactors can be provided in reinforced con-
crete vessels, which was mentioned earlier. In order to use the VVER reactors in heat supplies, it will be
necessary to build special reactors for nuclear boilers. Specialization in the production of relatively low-
temperature heat (150-170?C) will enable the basic plant design to be strongly simplified and reduced in cost,
and a facility of increased reliability and safety to be built. It will be most advantageous for such simplifica-
tion of the VVER facilities to use natural circulation and integral plant grouping, to make maximum use of
the principles of automatic control of the facilities, etc. Experience in the development, operation, and study
of certain generations of the VVER and of the cased boiling VK-50 reactor will enable a facility to be construc-
ted which is optimum in its parameters and characteristics. In a three-circuit facility, the pressure of the
coolant in the primary circuit cannot exceed 12-16 kgf/cm2. The manufacture of the plant is simplified consi-
derably. The necessary level of economic efficiency of automatic thermal power stations can be achieved with
a single thermal capacity of the units of - 500 MW. A general view of a possible reactor facility for an auto-
matic thermal power station is shown in Fig. 11. In any case, the use of reactor facilities for heat supplies
demands special attention to further increase their safety.
In ensuring the safety of the VVER reactors, the main attention is paid to the quality and control of the
plant, and to the establishment of reliability in the course of operation.
The shielding and insurance measures provided for in the plans of the first stations conformed to the
restricted scale of the maximum planned emergencies and increased role accepted for these facilities, and
removed the factor of remoteness of nuclear power stations from populated locations. The mass spread
of nuclear power stations leads to an intensification of the technical measures for the neutralization of poten-
tial danger. These demands have been set before the plans for the latest generation of nuclear power stations
with VVER reactors, including the new group of stations with the VVER-440.
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An analysis of the nature of the occurrence. of dangerous procedures shows that the dangerous conse-
quences of accidents for the. nuclear power station itself and for the surrounding population. can be prevented
by the creation-of reliable-and strong means of cooling the core, if emergency cooling prevents melting of
the fuel, then the role of external insurance barriers of the type of leaktight compartments and health-protec-
tive zones becomes less marked..
The loss of coolant by rupture of the largest circulating pipeline of the primary. circuit is accepted as
the critical design emergency in new designs. At present, a system of emergency cooling of the core is being
developed, which will-permit the expectation. that in the event of a dangerous rupture of the largest pipeline
(with a diameter of :500 and 850 mm in the VVER-440 and VVER-1000 respectively) no melting of the fuel will
occur and the fuel.elements may only partially lose their hermetic sealing.
In addition to this,. certain alternative systems for localizing activity.to compartments of nuclear power
stations have been developed, which arecalculated on the possibility of,a dangerous rupture of the largest pipe-
line (Fig. 12)..? As applicable to the VVER-1000, (a) the installation of a leaktight ferroconcrete shell has been
specified and (b), for the VVER-440' various alternatives have been devised for localizing compartments,
operating on the typical grouping of.coinmercial nuclear power stations with a leaktight steam-generator com-
partment.
Experience in the operation of working nuclear power stations with the VVER reactors shows that these
stations are safe sources of power, and have no harmful effect on the surroundings and the population. Never-
theless, the extension of the sphere of application of nuclear power plants, their proximity to population bodies,
and in particular, the installation of nuclear boilers, inevitably w ill lead to a further intensification of safety re-
quirements. Experience in the development of the VVER reactors shows that the increased demands on safety
can be achieved while maintaining the satisfactory economical indexes of these power plants.
DEVELOPMENT OF URANIUM-GRAPHITE -CHANNEL
REACTORS IN THE USSR*
A. P. Aleksandrov and N. A. Dollezhal'
The steady increase in the demand for electric energy and the problems in the use of organic fuel re-
quire an intensive development of nuclear power. One of the most important ways in which the Soviet Union
is meeting the nuclear power problem is by the use and development of channel-type power reactors. These
are graphite-nioderated reactors cooled by boiling water or steam under pressure in vertical fuel channels.
Channel reactors were chosen because of their characteristics and Soviet experience in reactor engineering
[1-3].
Channel reactors can be refueled without shutting down the reactor. This ensures a high reliability
and technical readiness of the reactor and permits the use of various fuel and structural materials and fuel
loading systems. Standard structures and components can be widely used in constructing the reactors and
increasing their power. Although the branching of the circulation loop is a certain disadvantage of channel
reactors, it permits the construction of coolant loops of relatively small volume, and this simplifies the solu-
tion of safety problems of such reactors, particularly as their unit power is increased [4].
Experience in the Construction of Channel Reactors. The development of channel reactors began 50
years ago when research and experiments on the construction of the world's first nuclear power plant (NPP)
at Obninsk were greatly expanded. The startup of this NPP in 1954 and its successful operation not only de-
monstrated the possibility of the peaceful use of atomic energy, but also permitted the testing and verification
* The present article is based on papers by V. G. Aden, Yu. M. Bulkin, A. P. Veselkin, V. K. Vikulov,
V. P. Volkov, O. M. Glazkov, V. V. Goncharov, I. I. Grozdov, M. B. Egiazarov, I. Ya. Emel'yanov, A.
D. Zhirnov, L. R. Kevorkov, A. Ya. Kramerov, S. P. Kuznetsov, E. P. Kunegin, L. I. Luninaya, Yu. M.
Mityaev, V. I. Mikhan, K. K. Polushkin, V. V. Postnikov, V. S. Romanenko, A. P. Sirotkin, V. M. Fedu-
lenko, S. M. Feinberg, V. N. Filippov, and R. T. Shapovalov. The authors are sincerely grateful to these
colleagues.
Translated from'Atomnaya Energiya, Vol. 43, No. 5, pp. 337-343, November, 1977.
0038-531X/77/4305-0985$07.50 ?1978 Plenum Publishing Corporation 985
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It i
Fig. 1. V. 1. Lenin Leningrad nuclear power plant (LNPP).. At the present time two units
with channel boiling reactors, each rated at 1000 MW, are in operation. Two more units
with, the&?same kind of reactors are- under construction. .The::total: power of the LNPP will
be 4000 MW "(electrical).
of a number of technical procedures, many' of which, were successfully?utiliZed in subsequent reactors. Thus,
partial refueling, cooling of parallel fuel channels by boiling water, the nuclear superheating of steam, - the
proof that graphite stacking can'withstand a' high temperature, etc. were achieved for the first time at the first
NPP. The' xperience' gained in~the construction and operation of the first NPP was utilized in the .construction
of the Siberian' NPP which went 'bn line in 1958, and in, two 2.1013cm-3 ? sec for
the energy balance in a hybrid reactor. This value has already been attained today in the tokamaks. Thanks
to the additional energy generation in the blanket, new cheaper thermonuclear schemes become possible which
are energetically unprofitable in a pure D=T reactor. Finally, the additional energy generation in the blanket
makes possible a significant reduction in the flux of 14-MV neutrons through the first wall of the thermonuclear
reactor, thereby significantly facilitating the solution of the problem of the radiation resistance of the mater-
ials. Summing up all that has been said, one can conclude that the constructionof hybrid reactors will be techni-
cally completely realistic toward the beginning of the 1990s, since the physical conditions necessary for hybrid
reactors should be achieved in the generation of thermonuclear plants now being constructed. The timely de-
velopment of hybrid reactors would permit successfully solving the problem of supplying with fuel the current-
ly rapidly growing power by light-water reactors. Together with this the creation of hybrid reactors will in
no measure hinder, but on the contrary is a natural stage on the path to pure thermonuclear reactors, if the
latter prove to be more attractive for mankind by virtue of their "purity. "
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As a result of almost 30 years of intense research the scientific basis of thermonuclear reactors has
been laid - the physics of the high-temperature plasma has been created. The plasma parameters have been
approached right up to those required for technical utilization in experimental installations of various types.
The technology of large superconducting magnet systems and the means for additional heating of the plasma -
injectors of neutral atoms and superhigh frequency generators - along with the technology of superpowerful
lasers and relativistic electron beams are approaching reactor requirements and scales. The accumulated
scientific and engineering potential permits proceeding today to the creation of demonstration systems, and
then to industrial ones.
Performance of the appropriate developments would permit a significant lightening towards the end of
this century of the problem of supplying thermal nuclear reactors with fuel and would give mankind somewhat
later on a new alternative source of energy in the form of clean D-T thermonuclear reactors.
BASIC APPROACHES TO NUCLEAR POWER STATION
V. A. Sidorenko, O. M. Kovalevich,
A. Ya. Kramerov, and Yu. E. Bagdasarov
A Combination of Technical and Organizational Measures is
the Necessary Condition of Real Nuclear Power Station Safety
The safety of a nuclear power station is taken to mean its capacity to ensure protection against radiation
effects for the technical personnel, the population at large, and the environment both in normal use and in the
event of possible disturbances, including major accidents. Experience in the design and use of nuclear power
stations with various reactor types in the USSR and the world shows that the problem of nuclear power sta-
tion safety in normal use has practically been solved. The release of radioactive products beyond the confines
of the power station corresponds to the accepted national or international norms. The trend toward more
stringent norms cannot have a significant effect on the development of nuclear power, since it is technically
possible to reduce or entirely eliminate such release.
Experience in nuclear power station development shows that the main emphases of safety work should
be as follows:
1) high-quality manufacture and installation of the equipment, as the basis for safety and for reducing
the probability of accidents and disturbances;
2) monitoring of the condition of the equipment in all stages of use;
3) the development and implementation of effective protective measures and equipment for eliminating
the causes of accidents and compensating for any disturbances that arise (switching on reserve equipment, emer-
gency reactor shutdown, etc.) or for limiting the consequences of disruptions (emergency active-region cool-
ing);
4) the development and implementation of measures for either complete localization of radioactive ma-
terials in the event of accident or limitation of the consequences of an accident (collection systems for radio-
active products, hermetic chambers, choice of power station location);
5) the implementation of all technical and organizational safety measures at all stages of design, equip-
ment manufacture, startup, use, and repair of nuclear power stations;
6) the standardization of all technical and organizational safety measures;
7) an effective system of state supervision for nuclear power station safety, enforced by law.
The present paper considers technical measures and organizational structures used to ensure nuclear
power station safety in the USSR.
Translated from Atomnaya Energiya, Vol. 43, No. 5, pp. 360-365, November, 1977.
0038-531X/77/4305-1009$07.50?1978 Plenum Publishing Corporation 1009
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Technical Safety Measures
Quality and Monitoring. The development of nuclear energy imposes more rigid requirements than or-
dinary energy technology on the quality and the corresponding quality-control standards in the manufacture
and installation of equipment. New materials and technological processes are being developed and particular
attention must be paid to the radiation stability of the equipment, the monitoring of welded joints, the reliabil-
ity of components, etc. The necessary quality and quality-control standards rise with the importance of the
component considered.
The most significant mistake made in the early stages of development was to underestimate the probabil-
ity that defects would appear as a result of factors such as vibration, temperature oscillations and fatigue, cor-
rosion (in particular, under stress), etc. In addition, experience has shown that the remote monitoring of the
condition of equipment while it is in use can involve unexpected difficulty. Accordingly, the first reactors were
underequipped with monitoring equipment, and their design and construction was often such as to impede access
to the important points of the reactor for monitoring and repair. Subsequent experience has allowed these defi-
ciencies to be partially removed: the amount of monitoring has been increased, weak points have been reinfor-
ced, and removable structures have been substituted for fixed, nonreplaceable structures.
The basic measures implemented in current power stations permit periodic monitoring during shutdown
and repair. Procedures and instruments are being developed for continuous or periodic monitoring of the
equipment while it is in use, including the analysis of acoustic and neutron noise, stress-wave emission, etc. ,
i.e., a multiple approach to monitoring is being developed.
Disturbance and Safety Limits. The development and implementation of measures intended to maintain
nuclear power station operation within safe limits necessitates a choice of the appropriate approach and the
definition of a "safety limit" for each specific initial disturbance. Nuclear power station safety measures for
"small" disturbances (disturbances that are entirely neutralized or those whose consequences may be avoided)
are very largely identical with the measures that ensure the reliability of a nuclear power station as an energy
source. To ensure uninterrupted use, the design of any nuclear power station must include adequate provision
for the complete neutralization of the disturbance and the maintenance of continued operation in the event of
any of a number of initial disturbances (e. g. , interruption of essential supplies, turbounit failure, breakdown
of the power system, isolated disturbances in the reactor-control system)..
Between the first safety limit (complete neutralization of the disturbance and continued operation of the
plant) and the ultimate limit arising from the definition of safety (the prevention of injury to the population at
large or contamination of the environment) it is usual, in practice, to employ a whole series of intermediate
safety limits, each of which corresponds to a specific initial disturbance (or a set of simultaneous but inde-
pendent disturbances) and to formulate appropriate specifications for the protection and localization.systems.
Maximum Accident and Quantitative-Probability Method. In considering the possible disturbances at a
nuclear atomic power station, the upper limit for which the design must provide adequate safety measures is
determined by the concept of the maximum possible accident (MPA). Indeterminacy in defining the MPA at
present arises from the natural subjectivity in interpreting the problem. Throughout the world there is con-
troversy between those nuclear power specialists who believe that the measures adopted to counter potential
dangers at atomic power stations are too rigorous and those who think them too lax. The present lack of reli-
able statistical data on the probability of failure necessitates a deterministic approach in standardizing the po-
tential nuclear power station dangers, i. e. , specification of the maximum possible (limiting) accident, safety
limits and corresponding disturbances, etc.
The MPA included in the design differs for different stages of nuclear power development and for power
stations with different types of reactor. In the first generation of stations, the protection and backup (second-
ary protection) measures included in the design were determined on the basis of the limiting MPA assumed for
the station and the distance from centers of population. Stations introduced today are designed to accommodate
more serious disturbances; the level of technical provision for the neutralization of disturbances is such that
the distance from population centers is a less significant factor. Designs have now been developed for power
stations (primarily for heat supply to towns) a minimum distance from population centers or close to large
concentrations of population. Of course, such stations require a further intensification of technical safety mea-
sures and an appropriate change in approach to safety limits and the corresponding disturbances.
Protection and Localization Systems. Experience in the design and use of nuclear power stations with
thermal reactors confirms the possibility of implementing safety limits on fuel-element damage for practically
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all disturbances that do not involve loss of seating of the primary neat-carrier loop. liisturaances associated
with major rupture of the primary loop are a special case and at present determine the MPA for thermal re-
actors. For these disturbances it is necessary to consider safety limits characterized by the yield of radio-
active products beyond a certain level.
Analysis of the occurrence of accidents shows that hazardous consequences of MPA with heat-carrier
loss are almost completely eliminated by the use of reliable and powerful means of cooling the active region.
If emergency cooling prevents fusion of the fuel elements, external safety barriers in the form of hermetic
chambers and shells play a less significant role. Both in early nuclear power stations and in current stations
with water-cooled-water-moderated reactors and RBMK, reliable emergency cooling is an extremely impor-
tant part of the safety systems. The protection afforded by the localization system provides a backup capacity,
in cases where the cooling system is less than totally effective.
If instantaneous transverse rupture of a primary-loop channel of maximum diameter is the limiting
accident, particular attention must be paid to the localization systems, since for such major disruptions of the
loop less than total effectiveness of the emergency cooling system may have hazardous consequences, with the
escape of radioactive products from the active region. What is required is an optimum technical solution in-
corporating both primary and secondary protection systems. In new nuclear power station designs, some of
which are already being implemented, the emergency-cooling systems employed are such that in the case of
primary-loop rupture fusion of the fuel elements is prevented, although partial unsealing of the fuel elements
may occur. Structurally, the localization systems may take different forms: e.g., hermetic ferroconcrete
protective shells (fifth block of the Novovoronezh nuclear power station) or systems with hermetic boxes,
without sealing of the reactor chamber, in conjuction with systems for the condensation of any vapor formed.
Particular mention should be made of a development in protection technology which allows the emergency-
cooling and localization systems to be simplified, while retaining the same level of safety: the creation of col-
lapsible power structures which maintain the whole of the potentially safe main structure of the first loop (in-
cluding, possibly, the reactor body) in the event of rupture and limit the rate of heat-carrier escape.
The development of such structures would allow practically complete elimination of the indeterminacy
in the idea of the MPA which is the cause of so much discussion.
Organizational Structure and Standardization of
Nuclear Power Station Safety Provisions
State Regulatory Bodies. State supervision of nuclear power station safety is the responsibility of the
following bodies:
the state committee for the supervision of working safety in industry and final supervision in the USSR
Council of Ministers (control over supervision in nuclear power) of the observance of technical-safety standards
in the design, construction, and use of a nuclear power station and its equipment;
the state inspectorate on nuclear safety (Gosatomnadzor SSSR) within the state committee of the Institute
of Atomic Energy of the USSR responsible for the observance of nuclear safety standards and regulations in
the design, construction, and use of nuclear stations;
state health supervision within the Ministry of Health (Minzdrava SSSR) of the observance of health regu-
lations and radiation-safety standards in the design, construction, and use of nuclear power stations, in order
to prevent irradiation of power station personnel and the surrounding population.
The jurisdiction of the state regulatory bodies extends over all establishments, institutions, and organi-
zations, regardless of their departmental affiliation.
Standardization Documents. The "General safety principles for nuclear power stations in design, cons-
truction, and use" are at present the main standardization documents outlining the basis for the consistent im-
plementation of technical and organizational safety measures in nuclear power stations at all stages of devel-
opment and use. As well as the statement of nuclear power station safety criteria and regulations that cover
the monitoring of safety-measure implementation, the documents outline basic general safety requirements for
various systems and for the power station as a whole, and also deal with the organizational and technical
problems of safe operation. Principles developed through experience of reactor design are outlined: the pro-
vision of reserve equipment, multiple-loop design (duplication), containment of damage, etc. It is necessary
to note certain requirements determining the range of disturbances that are to be considered in power-station
design:
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safety must be ensured for any single disturbance of the equipment in normal use, which may coincide
with an unobserved prolonged disturbance of other equipment;
the protection and localization systems must be constructed to take into account not only disturbance of
the system in normal use but also simultaneous disturbance or shutdown of one of the independent active pro-
tection and localization systems.
In these documents, a number of disturbances correspond to the safety limits and fix the presently adop-
ted MPA - rupture of the primary-loop channel, the most dangerous in terms of radiational consequences.
As well as the "General safety principles" there are detailed specific normative documents (standards,
regulations, procedures, etc.) in which requirements and procedures for nuclear power station safety are set
out.
Ensuring Implementation of Safety Requirements. The nuclear power station safety practices developed
and implemented in the USSR cover all the stages of nuclear power station design and use, and extend through-
out all the institutions and organizations associated with nuclear power.
In the course of designing a nuclear power station and its components, the observance of the safety re-
quirements is discussed at each stage of the design; it is useful to ask persons and organizations not associa-
ted with the development of the design for their point of view (appraisal, criticism, evaluation, etc.).
The manufacture of nuclear power station equipment and components is monitored. Monitoring at the
factory is carried out by special groups of personnel (technical-monitoring sections), independent of those
directly responsible for production. For particularly important components, special quality-control programs
are introduced. The external monitoring organizations are the state supervisory body (Gosgortekhnadzor SSSR),
the clients' representatives, and the designers. Their main activities are random sampling, inspection, par-
ticipation in factory testing, etc. When a particularly important component is to be accepted, a state accept-
ance commission is instituted, including representatives of all interested parties. Production samples carry
a plate bearing information on the factory or state tests.
Roughly similar procedures apply to construction and startup, and throughout the nuclear power station
operation itself. In the case of startup, the main role is that of the state acceptance commission, appointed
by government agencies from representatives of the regulatory bodies, the clients, and the designers.
Specific Safety Problems for Nuclear Power Stations
with Various Types of Reactor
Water-Cooled-Water-Moderated Reactor. Operating experience of nuclear power stations with reac-
tors of this type in the USSR and COMECON countries amounts to more than 50 reactor-years. Their develop-
ment from the first block of the Novovoronezh nuclear power station of 210-MW power to the present fifth block
of 1000-MW power indicates the evolution in the solution of safety problems.
In the event of accidents with the primary-loop flow, the Novovoronezh station reactor is protected by
an emergency flow-maintenance system of capacity 100 m3/h and a system of hermetic chambers (boxes) in the
primary loop, equipped with sprinkler devices for the condensation of any vapor. The emergency flow-main-
tenance system prevents active-region damage for tube ruptures of up to 100 mm. Special nozzles fitted at
junctions between large and small tubes limit the flow rate of incoming heat carrier. The primary-loop boxes
of the first block were designed for an excess pressure of 3 kgf/cm2. Additional secondary protection in the
case of unforeseen and improbable large-scale disturbances is provided in that the radiation-safety radius is
3 km and the power station is sited 50 km from the nearest large population center (Voronezh).
In the first group of reactors with a VVER -440 reactor, safety is further ensured by extensive monitor-
ing and by ease of access and maintenance. The hermetic boxes are designed for an excess pressure of 1 kgf/
cm2. From the boxes there is provision for a spray of vapor-air mixture at the level of the reactor-structure
breakdown through 9 adjustable valves. This spray moderates the accident process in the case of rupture of
small heat-carrier tubes when the sprinkler system fails to operate or is less than totally effective. Its maxi-
mum capacity is calculated for the release of water in the rupture of tubes of 200-mm diameter, although this
accident is more extreme than the maximum design accident.
The emergency-cooling system currently in use is able to deal with rupture of heat-carrier tubes of
maximum diameter (500 and 850 mm for VVER-440 and VVER-1000) without fusion of the fuel elements. The
main elements of the system are water stores and high- and low-pressure emergency pumps. There are
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several versions of the system for localizing activity within the structure of the power station in the event
of the maximum accident. The VVtR-1000 design envisages the construction of a hermetic ferroconcrete
shell above the reactor chamber and all the primary-loop chambers, designed for an excess pressure of 4
kgf/cm2 and capable of retaining all the heat carrier released from the primary loop. A sprinkler system
is provided for vapor condensation and cooling. The VVER-440 design includes a system of localization
chambers, based on a typical composition of series nuclear power stations, with a hermetic steam-generator
box designed for maximum excess pressure of 1.5 kgf/cm2 and without sealing of the reactor chamber. Total
condensation of the heat carrier released is ensured in an ascending series of boxes of special construction.
RBMK. Such reactors will provide approximately half of the power in the USSR in the next 10-15 years.
Pilot units of this type of 1000-MW power have been operating at the V. I. Lenin Leningrad nuclear power
station since 1973 and 1975; the next units are to be introduced at the Kurskii and Chernobyl' skii power stations.
Previously, this type of reactor was represented in the USSR by the world's first nuclear power stations, the
reactors of the Beloyarskii and Sibirskii nuclear power stations.
Safety measures in the event of accidents associated with primary-loop rupture are affected by the follow-
ing physicotechnical features of reactors of this type:
the positive vapor coefficient of reactivity or, at least, the absence of a strong negative inverse relation
on dehydration, which places increased demands on the reliability and efficiency of the emergency protection
systems;
the large volume of saturation vapor and water leading to a relatively low rate of pressure drop (there-
fore, solutions for the system of emergency cooling for the active region different from those for vessel reac-
tors are expedient);
the small accumulated activity of the water in the primary loop because of the removal of volatile pro-
ducts with the ejector gases and the possibility of detecting and replacing defective channels "on stream";
the large reserves of water for the emergency-cooling system and the possibility of reliable direct trans-
fer of the water to the active region;
the possibility, in principle, of having a "fractional" channel of narrow autonomous cooling loops, which
decreases the volume and rate of flow;
the large size, because of which it is difficult to use a general protective shell and more expedient to
employ other localization systems, e.g., a bubble tank.
At present in RBMK designs, the MPA assumed is the worst of two accidents in terms of radiational
consequences:
total rupture of the general pressure or suction collector of the main circulation pump or of the connec-
tion to the return valves at the inlet to the communal collector;
rupture of the communal collector.
On the basis of past experience of the design and use of nuclear power stations with RBMK it may be
assumed that the specific features of these reactors may fairly simply and cheaply be made to satisfy the
safety requirements for practically all convenient nuclear power station sites.
Fast Reactors. The scientific-technical techniques for fast-reactor safety in the USSR are based on
experience obtained in the development and use of the BR-5 (BR-10), BOR-60, and BN-350 reactors, the con-
struction of a nuclear power station with a 600-MW (electrical) BN-600 reactor, and the development of de-
signs for a 1500-MW (electrical) power station.
From a safety point of view, the main features of fast reactors with sodium heat carriers are associated
with the active-region properties (the absence of a moderator, high thermal stress, short lifetime of the re-
tarded neutrons for plutonium and mixed fuel, swelling of the constructional materials) and the consequences
of the use of a sodium heat carrier (large heating in the reactor, ambiguity of the hydraulic characteristics of
the heat-extraction channel, sodium vacuum reactivity effect, large induced activity of the heat carrier, and
the chemical activity of sodium with oxygen and water).
Operating experience confirms the important properties of sodium-cooled reactors from the safety point
of view: the absence of contamination and problems of spatial instability, negative temperature reactivity
effects, and also favorable conditions for the extraction of residual heat liberation. Because of the large
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margin to the boiling point (400-500?C) and the large heat-transfer coefficient of sodium at very low rates it
is possible, without hazardous consequences, for heat to be transferred from the active region and accumul-
ated in sodium and the constructional elements of the first and second loops in conditions of natural circula-
tion, even in the case of prolonged shutdown of the system for the discharge of heat to the atmosphere. The
rate of increase in heat-carrier temperature in these conditions is no more than 20-50 deg C/h.
The probability of an accident analogous to the MPA in reactors with water under pressure (rupture of
heat-carrier tube of maximum diameter) is significantly lower in a sodium reactor than in thermal reactors,
because of the low pressure in the first loop and the low corrosional activity of sodium. Hazardous conse-
quences of such an accident are easily eliminated in the integral composition (as in the BN-600).
A more dangerous and less studied possibility for fast sodium reactors is instantaneous-neutron run-
away as a result of rapid destruction of the active region and the appearance of a positive sodium reactivity
effect (other possibilities of the rapid introduction of positive reactivity may be excluded by constructional and
design measures). For modern nuclear power station designs with the BN-600 similar dangers are absent,
since in this case the integral sodium vacuum effect is negative. In future series nuclear power stations of
power 1000-1500 MW (electrical) it is positive and may reach large values. Note that, given current know-
ledge, the only conceivable situation that might lead to the appearance of a positive sodium effect is the switch-
ing off of the primary-loop pumps and simultaneous failure of all the control-rod emergency protections.
Evidently, the probability that such a situation will arise is extremely small, and may be reduced to required
values by increasing the reliability and duplication of the protection systems and equipment. All other cases
(the development of local damage, the passage of large gas bubbles or retarding materials through the reactor,
etc.) may be eliminated by appropriate constructional and design measures.
In the light of current requirements regarding possible disturbances at nuclear power stations with ther-
mal reactors, it may be said that the design solutions adopted in fast reactors are adequate for safety and im-
pose no additional restrictions on the choice of power station sites.
ECONOMIC ASPECTS OF THE DEVELOPMENT OF
NUCLEAR POWER AND FUEL CYCLE PLANTS IN
THE USSR
N. P. Dergachev, A. K. Kruglov,
V. M. Sedov, and S. V. Shuklin
Therifial vessel-type reactors, viz., the water-cooled-water-moderated (VVER) and uranium-graphite
channel-type RBMK, will be the most typical for nuclear power stations in the next 20-25 years in the Soviet
Union, as in other countries.
An assessment of the present state of the art with respect to nuclear power stations withfast reactors,
particularly with respect to the experience from the operation of the Shevchenko nuclear power station with the
BN-350 reactor, and the rate of plutonium build-up in thermal reactors permit the conclusion that nuclear
power stations with fast reactors will not be commissioned before 1985. The development of the nuclear power
industry over the next 20-25 years presents itself as a combination of nuclear power stations with thermal re-
actors and with fast breeder reactors. Studies have shown that nuclear power stations with fast reactors have
somewhat higher costs (20-30%) which, however, are compensated by lower electricity generating costs.
Rate of Development of Nuclear Power. According to the Directives of the Twenty-Fifth Congress of the
CPSU, electrical generating capacity of (13-15) ? 103 MW will be added by nuclear power stations in the USSR in
1976-1980, i.e., the total capacity of nuclear power stations in 1980 will reach (19-21).103 MW [1].
If 5-6 years is taken as the period for doubling the capacity of nuclear power stations, for making predic-
tions it is convenient to use a function determining the rate of introduction of nuclear power stations in the form
Translated from Atomnaya Energiya, Vol. 43, No. 5, pp. 365-369, November, 1977.
0038-531X/77/4305-1014$07.50 ?1978 Plenum Publishing Corporation
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The icebreaker Arktika, put into.service in 1975. ,The largest nuclear-powered
ship in the world; shaft power 75,000 hp. The Arktika does important worK for the
national economy by convoying caravans of ships in the Arctic Ocean. ' In August 1977
the ship reached the' geographical North Pole,-the first surface vessel in history to reach
this. geographical place'by active `sailing.
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NE(t)=NoEexP[a(t-to)1-N,. (1)
Here a = 0.1 g-1; NOE = 40.103 MW(electrical); N1 =19.103 MW(electrical); to = 1980 (yr chosen by normaliza-
tion conditions).
Equation (1) makes it possible to simplify the analysis and in a certain way to predict the development
of the power of the fuel-cycle plants beyond the limits of the time interval specified in the Directives.
Below are some characteristics of the reactors for nuclear power stations that constitute the
program's basis:
f VVER-100 [3] I RBMK-1000 [2] BN-1600 [4]
Thermal capacity, MW ................ 3000 3200 3750
Electrical capacity, MW .............. 1000 1000 1600
Initial charge,. ton ..:................. 65 180 3(Puf)
Enrichment of initial charge, % ........ 2.6 1.1 -
Annual recharging, tons ............... 21 50 2.2(Puf)
Enrichment of stationary charge, % ..... 4.4 1.8 -
Time to first unloading, yr ............ 1 1 1
External fuel cycle, yr . 1.3 1.3 1.3
Load factor ... .................... 0.8 0.8 0.8
In the structure of the power industry, in order to obviate possible difficulties in placing orders with
industry for reactor equipment, a 1:1 ratio is adopted for nuclear power stations with VVER and RBMK reac-
tors. A significant role in the structure of the nuclear power industry will be played by nuclear power stations
with fast reactors in view of the build-up of a considerable quantity of plutonium in the fuel of thermal reactors.
The demand for enriched uranium for the given growth rate of the power industry can be found from
2 t
Go (t) + GE (t) = I GotNEi (t) + Y.GEi I NE t (t -Ti) dt, (2)
i i 1975
where Go is the fuel of the initial charge; GE is the fuel of the stationary recharging. The quantity of enriched ?
uranium, in tons, is easily found from Eq. (2).
The Role of Nuclear Power Stations with Fast Reactors. The most feasible way of using the plutonium
produced in thermal reactors is in nuclear power stations with fast reactors.
With the given data taken into account, the commissioned capacity of nuclear power stations with fast re-
actors (NF. E) can be described by the equation
.2
d F p() = p d ,(t-i) +(Co -1)gAF.E(t-i). (3)
di GF dt
Here PI is the build-up of plutonium (Pu f) in the fuel; GF is the charge of plutonium in the complete fuel cycle
of the fast reactor, in tons [GF = (2.1-2.2)Go]; Co is the conversion factor of the fast reactor (1.3); q is the
quantity of plutonium burned in the reactor in a year, ton /yr*; GEi (t) is given by Eq. (2); T is the plutonium re-
sidence time in the cycle for thermal and fast reactors (T = T1 + T2 = 2.3 yr). The solution of Eq. (3) is of the form
2
fitPt i noE a t z-r a(ti 10)+ v (t ti) N 1 i VU-if i VY-et
P. E (t) = x e ( o) _ e i+vL t ( } Y )
~F (1-f- YT) _ ~ y \ ( ) - Y (e 1+vx ) l _. NZe i+vz , (4)
'- i tx 1 Vti) f
-F
where Pi is the fraction of reactors of the i-th type; t1 is the time required to commission nuclear power sta-
tions with BN-1600 reactors, t1 is the year - 1985; N2 is the number of nuclear power stations with fast re-
actors which can be commissioned by 1985 in accordance with the size of the plutonium pile-up, N2 =
MW(electrical); and y = [(CO - 1) q]/GF.
4.5.103
With a burn-up of more than 20.i03 MW-days/ton the plutonium pile-up (239Pu + 241Pu) in VVER
fuel is
more than 6 kg [5, 6]. In the RBMK reactor, with a burn-up of 18.103 MW-day/ton, the plutonium pile-up
(239pU + 241Pu) exceeds 2.5 kg [7].
Calculation by Eq. (4) shows that 13 GW(electrical) of capacity in nuclear power stations with fast re-
* It has been assumed that the fission of 238U - 240Pu and 242pu compensates for the radiation capture of neu-
trons on 239Pu and 241Pu.
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actors can be commissioned by 1990 and 64 GW(electrical) by the year 2000.
If in Eq. (4) we set y = 0, which corresponds to Co = 1, the capacity of nuclear power stations with fast
reactors by the year 2000 will be 47 GW(electrical), i.e., will decrease by only 25%. It thus follows that the
effect of only fast reactors on the rate at which capacity is commissioned in the next 20 years is not decisive.
Therefore, it is not an indisputable assertion to say that effects must be concentrated over the next 10-15
years on the development of fast reactors with a maximum conversion factor to the detriment of the simplicity
of design and operating reliability.
Let us consider how changes in the length of the external fuel cycle T2 for thermal and fast reactors
affect the rate of commissioning of nuclear power stations (from the point of view of ensuring fuel). If T2 is
reduced to 0.8 yr (fuel water-cooling time -0.5 yr) the capacity of nuclear power stations with fast reactors
by the year 2000 will increase by 30% (NF. E = 85 GW, GF = 5.4 tons), whereas if T2 is increased to 3 years,
NF. E = 37 GW, GF = 11 tons, i. e. , the capacity is roughly halved.
Plutonium resources in the next 20-25 years practically will not limit the growth of nuclear power
stations with fast reactors. Construction of such nuclear power stations will alow an approximate 25% de-
crease in the demand for natural uranium.
Basic Principles for the Development of Fuel Reprocessing Plants. The structure of the nuclear power
industry poses a number of requirements upon the development of fuel reprocessing plants.
1. If plutonium is to be involved in power generation by being used in fast reactors, radiochemical
plants for reprocessing fuel elements and plants for fabricating fuel elements for thermal and fast reactors
must be built over the next 10-15 years.
2. The external fuel cycle for thermal and fast reactors should be no longer than 1-1.3 years (with plu-
tonium incorporated into power generation).
3. The number of nuclear power stations with fast reactors commissioned depends on the size of the
plutonium charge in the fuel cycle.
The size of the plutonium charge in the fuel cycle of a fast reactor can be found from the expression
GF = Go [1 + 1/T fl (0.3 + Tc + Ttr + Trc + Tfabr]. (5)
where Tfl,, Tc, Ttr9 Trc? and Tfabr are, respectively, the fuel lifetime in the reactor, the fuel cooling time,
transportation time, duration of radiochemical reprocessing, and duration of fuel-element fabrication. The
fuel reserve for the first reloading or refueling is 0.3 (Tfl = 1.4 yr).
When Tc = 1 yr and Ttr = 1 month, Eq. (5) becomes GF = Go [1.98 + 0.71 (Ttr + Tfabr)]?
If we proceed from the position that the plutonium charge in the fuel cycle should not exceed (2.1-2.2)
Go, then 0.16