SOVIET ATOMIC ENERGY VOL. 40, NO. 2

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP10-02196R000700070002-9
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• Document Page Count: 129
• Document Creation Date: December 23, 2016
• Document Release Date: April 10, 2013
• Sequence Number: 2
• Publication Date: August 1, 1976
• Content Type: REPORT

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Russian Original Vol. 40, No. 2, February, 1976
August, 1976
SATEAZ.40(2) 119-242 (1976)
SOVIET
ATOMIC
ENERGY
Al UI-Ir1An %mtrl vin
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
CONSULTANTS BUREAU, NEW YORK
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SOVIET
ATOMIC
ENERGY
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 Energiya:
Editor: M. D. Millionshchikov
Deputy Director
1. V. Kurchatov Institute of Atomic Energy
Academy of Sciences of the USSR
fv1oscow, USSR
Associate Editor: N. A. Vlasov
A. A. Bochvar
N. A: Dollezhal'
V. S. Fursov
I. N. Golovin
V. F. Kalinin
A. K. Krasin'-
Copyright ? 1976 Plenum Publishing Corporation, 227 West 17th Street, New York,
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
August, 1976
Volume 40, Number 2 February, 1976
CONTENTS
Engl./Buss.
ARTICLES
Peaceful Use of Nuclear Energy and the Problem of Nonproliferation of
Nuclear Weapons - I. D. Morokhov, K. V. Mysanikov,
and V. M. Shmelev ................................................. 119 99
Atomic Science and Technology in the National Economy of the USSR
- A. K. Kruglov .................................................... 123 103
Experience in the Construction of Large Power Reactors in the USSR
- N. A. Dollezhal' and I. Ya. Emel'yanov ............................. 137 117
Physical Start-up of the RBMK-Reactor of the Second Unit of the V. I. Lenin
Nuclear Power Station, Leningrad - I. Ya. Emel'yanov, M. B. Egiazarov,
V. I. Ryabov, A. D. Zhirnov, V. P. Borshchev, B. A. Vorontsov,
A. N. Kuz'min, Yu. I. Lavrenov, V. S. Romanenko, Yu. M. Serebrennikov,
and A. P. Sirotkin .................................................. 147 127
High-Temperature Reactors as a Factor of Scientific Progress in
Power Generation - N. A. Dollezhal' and Yu. I. Koryakin ............... 154 133
Methods of Mathematical Modeling and Optimization of Nuclear Power Plants
- L. S. Popyrin .................................................... 166 145
Operative Monitoring System for the Energy-Liberation Fields of the Reactors
in the Beloyarsk Nuclear-Power Station - N. Ya. Kulikov, ]. I. Snitko,
A. M. Rasputnis, and V. P. Solodov ................................... 174 152
Structural-Geological Features of Uranium Deposits in Collapse Calderas
- V. A. Nevskii, N. P. Laverov, and A. E. Tolkunov .................... 178 155
Continuous Underground Ore-Mining Operations with the Aid of Nuclear
Explosives - V. V. Gushchin, K. D. Vasin, B. I. Nifontov,
Yu. L. Odrov, K. V. Myasnikov, V. M. Kol'tsov, G. N. Kornev,
and V. A. Degtyarev ................................................ 185 162
INFORMATION
Atomic Energy in the USSR in the Ninth Five-Year Plan - L. M. Voronin
and E. Yu. Zharkovskii .............................................
LETTERS
Effect of Neutrons Reflected from the Walls of a Room on Pulse Parameters
in Fast Reactors - V. F. Kolesov .................................... 194 171
Slipping Conditions in the Problem of the Minimum Critical Mass
- A. M. Pavlovichev and A. P. Rudik ................................. 197 173
Effective Half-Life of 252Cf - V. K. Mozhaev ............................... 200 174
Measurement of the Effective Cross Section for the Fission of 252Cf by Fast
Reactor Neutrons - E. F. Fomushkin, E. K. Gutnikova, G. F. Novoselov,
and V. I. Panin ..................................................... 202 176
Unusual Mineral Associations in the Oxidation Zone of Sulfide-Free Uranium
Deposits - V. N. Levin and L. N. Belova .............................. 204 177
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CONTENTS
Effective Gamma-Ray Attenuation Coefficients for Radioactive Ores
- G. F. Novikov, A. Ya. Sinitsyn, and Yu. D. Kozynda ............ . .
BOOK REVIEWS
Yu. A. Gulin. The Garnma- Gamma Method of Investigating Oil Wells - Reviewed
by E. M. Filippov ....... ................................. .
CMEA CHRONICLE
19th Conference of the CMEA Permanent Committee on Atomic Energy Use
- Yu. I. Chikul ....................................................
Work of the Coordinating Scientific-Technical Council on Reprocessing
Irradiated Fuel of AES - V. I. Zemlyanukhin ..........................
Results of the Work of the Coordinated Scientific-Technical Council on Radiation
Techniques and Technology (KNTS-RT) - A. K. Zille ...........................
Journal of Collaboration ............... ................ ....... .............
CONFERENCES AND MEETINGS
A Conference on the Problems of the Design, Assembly, Starting, and Operation of
Atomic Electric Power Plants - Yu. I. Mityaev. .:............................ .
Conference on the Technical Applications of Superconductivity - A. G. Plesch..........
4th International Conference on Thermal Emission Energy Conversion
- A. I. Kulichenkov ........................................................
7th European Conference on Controlled Thermonuclear Fusion and Plasma Physics
- Z. I. Kuznetsov .........................................................
Soviet West German Symposium "Armatures and Pumps for Power Stations"
- R. R.lonaitis ...........................................................
Technical Conference Nuclex-75 - L. N. Podlazov .................................
Conference of Specialists on Data Processing for Reactions with Charged Particles
- L. L. Sokolovskii ........................................................
24th Session of the Scientific Committee of the United Nations - R. M. Aleksakhin
and A. A. Moiseev .........................................................
Engl./Russ.
210 181
211 181
213 182
216 184
218 185
220 186
224 188
227 190
229 192
231 193
EXHIBITIONS
Soviet Exhibitions at the 4th International Exhibition of the Nuclear Industry
Nuclex-75 - V. A. Dolinin .................................................. 237 195
BIBLIOGRAPHY
I. Ya. Emel'yanov, P. A. Gavrikov, and B. N. Seliverstov. Control and Safety
of Nuclear Power Reactors - Reviewed by V. I. Plyutinskii ................ 239 197
Yu. V. Seredin and V. V. Nikol'skii. Principles of Radiation Safety in Prospecting
and Exploration for Minerals - Reviewed by E. D. Chistov ................ 241 198
The Russian press date (podpisano k pechati) of this issue was 1/23/1976.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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PEACESFUL,?USE OF NUCLEAR ENERGY AND THE
PROBLEM OF NONPROLIFERATION OF
NUCLEAR WEAPONS
I. D. Morokhov, K. V. Myasnikov, UDC 621.039/623.454.8
and V. M. Shmelev
An analysis of the population increase of the world and of energy requirements per capita shows that
as much energy will be consumed in the world from 1970 to 2000,as was consumed in the last 20 centuries.
The energy demand at the end of the century will be approximately three times greater than present levels.
Even today the most highly developed capitalist countries are feeling the lack of "energy sources. This cir-
cumstance has created great interest in the use of a new energy source - nuclear power.
The advantages of nuclear power from the economic and ecological aspects are extremely encourag-
ing. The economic indicators for large nuclear power installations and future improvement in nuclear
technology, the development of breeders for example, convincingly demonstrate that the traditional energy
sources (coal, oil, gas) will be increasingly supplemented and displaced by nuclear fuel. The International
Atomic Energy Agency (IAEA) estimates that by the year 2000, 50%of the total energy balance of the world
will consist of electrical power and about 50%of all electrical power will be produced by means of nuclear-
power installations (Figs. 1, 2).
The energy crisis has significantly increased the competitiveness of nuclear power in comparison
with the usual sources of electrical power. Today, not much more than 20 years after the start-up of the
first nuclear-power station in Obninsk, there are more than 100 nuclear-power stations operating in the
world with a total installed power close to 40 million kW, and according to predictions the power from
nuclear-power stations will be about three billion kW by the year 2000. As is clear from Fig. 2, nuclear
power should occupy a leading position in the energy picture of developing countries even in the next few
decades. Among these countries, there are those that will consider it more economic to construct nuclear-
power stations and not consume their own oil and gas reserves for power purposes. It is precisely nuclear
power that makes it possible to free humanity from the threat of a power shortage and to provide power for
a developing civilization.
In addition to power, nuclear technology and nuclear methods are becoming indispensable in industry,
agriculture, medicine, and geology for monitoring atmospheric contamination and in other fields.
The swift progress of the peaceful use of nuclear energy brings special acuteness to the problem of
preventing the proliferation of nuclear weapons. The development of nuclear technology leads to wide-
spread "creeping" of nuclear materials over the entire planet which can give rise to the preconditions for
the proliferation of nuclear weapons. Nuclear materials intended for peaceful uses can be diverted into
military channels for the production of nuclear weapons or other explosive devices. The possibility of such
diversion exists when thermal and fast reactors are used where plutonium, which can be used for military
production, is produced as the result of irradiation of uranium. Enriched uranium itself can be used for
the manufacture of nuclear weapons.
According to the IAEA, the amount of natural uranium in power reactors will rise to 15,000 tons by
1985 and the amount of enriched uranium to 70,000 tons (Fig. 3). According to the same data, 700 tons of
plutonium will be in storage, and the amount of plutonium in the fuel of thermal and fast reactors will be
Translated from Atomnaya Energiya, Vol. 40, No. 2, pp. 99-102, February, 1976.
?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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150 100
2000
1990
1980
1970
1960
1950
50 0 0,25 0,50
Fig. 1
- 1990
- 1985
1980
1975
1970
Fig. 1. On the left, worldwide power production (trillions of kWh); on the right, fraction
used in the form of electrical power: 0) total; ^) electrical power.
Fig. 2. Electrical power production (thousands of MW): on the left, total installed power
for world's electrical power stations; on the right, installed power for electrical power
stations in developing countries-.0.) all stations; ) nuclear power stations.
more than 500 tons (Fig. 4). The fact that even by 1980 the peaceful production of plutonium and enriched
uranium will be sufficient for the manufacture of tens of atomic bombs per day is clear evidence of the
acuteness of the problem. At the same time, if even a small portion of the stored plutonium is used in
accordance with existing international agreements for the production of explosives for peaceful purposes,
humanity will obtain a huge and inexpensive source of energy.
Many studies on the peaceful use of underground nuclear explosions have been carried out over the
past decade in the USSR, USA, England, France, and several other countries. At the Geneva conference in
May 1975, to consider the effect of the Treaty on Nonproliferation of Nuclear Weapons, the IAEA presented
a review of its activities in accordance with Article V of the treaty. The report presented the conclusions
of four international conferences held by the agency. in 1970, 1971, 1972, and 1975 with respect to two basic
problems: the technical and economic feasibility,and the safety of underground nuclear explosions. Be-
cause of the exceptional compactness of the explosive devices and the low cost of the energy produced,
peaceful nuclear explosions make it possible to create new, efficient technical procedures, to carry out
large-scale civil-engineering projects, and to provide new scientific opportunities. In the report, peaceful
nuclear explosions were divided into three groups according to the degree of investigation of the technology:
1. Recognized industrial forms of application of peaceful explosions.
2. Large-scale experiments under field conditions.
3. Laboratory development and theoretical studies.
In the first group are such examples of the use of peaceful nuclear explosions as the elimination of
accidental gas blowouts, the intensification of oil extraction, the creation of underground storage cavities
in rock-salt masses, and the creation of open reservoirs for water storage.
In the second group are the experiments performed in the USA in low-permeability gas formations
for the purpose of developing a commercial technology for gas extraction. Calculations by American scien-
tists show that the national commercial gas reserves will be doubled or tripled by a successfully developed
technology. This group also includes experimental nuclear explosions with removal of soil for the develop-
ment of a technology for the construction of large industrial works - canals: Panama, Pechora-Kolva
(USSR), Orinoco-Rio Negro (Venezuela), and'Kra (Thailand). The use of underground nuclear explosions
should reduce expense and construction time significantly. In the construction of the Kra canal, e.g., the
saving would be two billion dollars out of a total construction cost of six billion dollars.
In the third group are the most complex areas for the use of peaceful nuclear explosions such as the
extraction of oil by distillation from oil-bearing shale, leaching of copper at the point of deposit, production
of geothermal energy, and disposal of radioactive wastes. These are projects which provide for the use of
peaceful explosions for scientific purposes.
Considerable experience and experimental data have accumulated in recent years which make it pos-
sible to predict the seismic and radiation consequences of underground nuclear explosions with a definite
engineering approximation. Specially constructed elements are used for the reduction of radiation conse-
quences. In the report on specific examples, it was shown that the potential irradiation of the population
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from contained nuclear explosions may be only an
insignificant fraction of the dose level established
by the International Commission on Radiation Pro-
tection. For explosions with soil removal, it may
appear necessary in some cases to evacuate the
local population temporarily in order to reduce
irradiation dose levels.
Fig. 3. Amount of
natural (1) and en-
riched (2) uranium
loaded into power
reactors.
1975 19e0 Year
Fig. 4. Amount of
stored plutonium (1)
and amount of pluton-
ium in reactor fuel
(2)
The seismic consequences of explosions can
be predicted with a sufficiently high degree of accu-
racy including possible costs for reconstruction.
By their very nature, the seismic consequences be-
long more with the factors which determine the eco-
nomic efficiency of peaceful explosions since the
safety of the population can be assured with a high
degree of reliability by the working out of special
measures.
The prevention of the proliferation of nuclear
and thermonuclear weapons of mass destruction re-
presents an urgent problem for mankind. Therefore,
a need arose for the development of methods guaran-
teeing that materials used in peaceful nuclear activities would not be diverted into military channels. A
system of such guarantees, and of methods and means for accountability and inspection of nuclear mater-
ials, has been developed and used by the IAEA for more than 15 years.
The Treaty on Nonproliferation of Nuclear Weapons, which became effective in 1970, created a juri-
dical basis for the use of international guarantees. By the middle of 1975, 43 governments had concluded
inspection agreements with IAEA and had put them into operation; about 30 agreements are in the stage of
development or legal formulation. On June 30, 1975, there was under the control of IAEA 47 nuclear power
stations, 115 reactors, 29 plants for the processing, manufacture, and reprocessing of fuel, and 195 other
separate areas of nuclear-material accounting. In a single year (from July 1974 through June 1975), the
agency performed 502 inspections without finding any cases of treaty violations. Experience in the applica-
tion of IAEA guarantees clearly showed that its control mechanisms functioned reliably to ensure satisfac-
tion by all governments of their obligations.
The treaty facilitates the widespread introduction of nuclear power into the economy of developing
countries. The technical assistance extended by IAEA to developing countries in the use of nuclear power
is expanding year by year. In the last five years, the total amount of such help was 23.8 million dollars in
comparison with 23.5 million dollars in the preceding eleven years.
The agency has performed preparatory work in providing service for the use of peaceful nuclear ex-
plosions. It consisted of four technical meetings on this problem which made it possible for nonnuclear
countries to evaluate the potential merits of peaceful nuclear explosions, developed principles and pro-
cedures for international inspection after their occurrence, and created a special subsection for providing
services and further study of this problem.
The Treaty on Nonproliferation of Nuclear Weapons, being an international instrument creating
serious barriers to the proliferation of nuclear weapons, not only is not an obstacle in the path of exten-
sive application of nuclear power for peaceful purposes, but facilitates such application and creates favor-
able conditions for international cooperation in this field.
An important event in strengthening the means for nonproliferation of nuclear weapons was the con-
ference to consider the effect of the treaty during the past five years. Experience during the five years
showed that all signatories to the treaty strictly observed its principles. The most important political
outcome of the conference was the further expansion of the group of treaty signatories on the eve of the
conference, or during the course of the conference, to include West Germany, Italy, Belgium, the Nether-
lands, Luxembourg, Gambia, Ruanda, and Libya. At the same time, a considerable group of "near-nuclear
governments" and two nuclear powers remain outside the framework of the treaty.
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The conference emphasized the importance of a system of accounting and inspection for nuclear ma-
terial from the viewpoint of the responsibility of treaty signatories and of the importance of collaboration
with IAEA for cooperation in achieving the guarantees in accordance with Article III of the treaty. The con-
ference expressed vigorous support for effective guarantees by the IAEA on the application of this inspec-
tion to all peaceful nuclear activities of treaty nonsignatories that import nuclear materials and special
technical equipment and for recommendations for strengthening the protection of nuclear materials against
theft, which were adopted to a considerable extent on the initiative of the USSR.
The conference confirmed the responsibility under Article IV for all treaty signatories to facilitate
as far as possible the most complete interchange of equipment, materials, and scientific and technical
information, pointing out that the treaty provides favorable conditions for the expansion of international
collaboration in this field. The conference adopted recommendations on further development of interna-
tional collaboration, in particular on studies of the creation of regional centers for the nuclear fuel cycle,
which are extremely important for the accomplishment of control.
The conference devoted considerable attention to Article V concerning peaceful nuclear explosions,
which provides thattreaty signatories not having nuclear weapons will obtain the benefits from peaceful
use of nuclear explosions "through the appropriate international organ in which governments not possessing
nuclear weapons are properly represented." On the initiative of the socialist countries, the conference
adopted a resolution that the IAEA will be just such an international organ. The conference called for a
continuance of the work in this field and entrusted the agency with a central role in problems associated
with the provision of services relative to peaceful nuclear explosions. Of great value is the proposal
adopted by the conference on the recommendation of the socialist countries that the potential benefits of
peaceful nuclear explosions be accessible to governments not possessing nuclear weapons and not treaty
signatories under appropriate international inspection and by means of procedures developed by IAEA.
These proposals, accepted by the conference, eliminate the need for the construction of nuclear devices
by nonnuclear governments for peaceful purposes, which is very important from the viewpoint of the non-
proliferation of nuclear weapons.
An important position in the treaty is occupied by Article VI which specifies the responsibility of
signatory governments to conduct negotiations to halt the nuclear arms race and also to achieve universal
and total disarmament.
Important international agreements have been developed in this field in recent years with the active
participation of the Soviet Union and other socialist countries. Of outstanding value for the cause of peace
and international security are the Soviet-American agreements signed in 1972 and 1973 on the prevention
of nuclear war, on the limitation of systems for defence against ballistic missiles, and on certain measures
in the field of limitation of strategic offensive weapons. During the third Soviet-American high-level
meeting in the summer of 1974, new important agreements were reached, including a treaty on the limita-
tion of underground tests of nuclear weapons. In accordance with this treaty, the USSR and USA are con-
ducting negotiations on peaceful nuclear explosions. This is making a significant contribution to the matter
of general prohibition of nuclear-weapon testing.
The conference stressed that adherence to the Treaty on Nonproliferation of Nuclear Weapons by
nonnuclear governments was the test method for mutual assurance of the renunciation of nuclear weapons
and an effective measure for strengthening their safety. Treaty signatories confirmed their great interest
in the prevention of further proliferation of nuclear weapons. They confirmed their vigorous support of the
treaty, their unshaken devotion to its principles and purposes, and their responsibility to carry out its
regulations completely and more effectively.
The time which has passed since the day the Treaty on Nonproliferation of Nuclear Weapons became
effective has demonstrated its widespread international recognition as an effective instrument to check the
proliferation of weapons of mass destruction and to promote the peaceful use of nuclear energy.
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AT.-OM,IC7.~S_CIENCE AND TECHNOLOGY IN THE
NATIONAL ECONOMY OF THE USSR
A. K. Kruglov UDC 621.039/338
Little more than 30 years have passed since the first nuclear reactor in the world first went critical,
yet now the nuclear industry in various countries is an independent field occupying a leading place in the
national economy.
Atomic science in the USSR not only promotes the creation of the new production needed for nuclear
power which has been successfully developed, but is also, involved in the introduction of its achievements
in the national economy. The energy developed by nuclear fission is employed in a number of new direc-
tions.
Radioactive isotopes are used for creating small atomic batteries to supply energy to apparatus and
instruments, and also to power artificial organs in man. They are widely used in medicine, industry, and
.agriculture not just for marking atoms (radioactive indicators), but also as sources of ionizing radiation
for changing the material properties.
The nuclear explosives which are produced enable us to use the energy of the atom in order to con-
struct canals, reservoirs, underground storage chambers for various liquids and gases, and also to create
new methods of extracting useful minerals.
The achievements of nuclear physics, the science that provides the basis for the, atomic industry,
have lead to the creation of charged-particle accelerators of various types and designs, which are widely
employed in the national economy, giving boosts to the accelerated development of various branches of
technology.
The development of nuclear power engineering in the USSR, already enables the nuclear power sta-
tions to generate energy at a cost per kilowatt-hour which is lower than can be achieved on thermal-power
stations, particularly in regions far removed from sources of fossil fuels [1].
TABLE 1. Experimental Values of Depth of Improvements in the cost effectiveness of the nu-
Burn-Up of the Nuclear Fuel at the Novo- clear-power stations are being achieved both by increasing
the unit powers of the reactors and also by improving the
vorenezh Nuclear-Power Station, kg/ton Ti technology of reactor construction and the production of
Bum-up
Unit I
Unit II
239Pu
Total
16,4
18,9
14,3
22,6
nuclear fuel. In a relatively short period of time (about
20 years) since the commissioning of the world's first nucle-
ar-power station at Obninsk, the national economy is bene-
fiting from nuclear reactors which have 200 times the unit
power of the reactor in the first nuclear-power station.
Two channel-type uranium-graphite reactors (type RBMK)
Translated from Atomnaya Energiya, Vol. 40, No. 2, pp. 103-116, February, 1976.
?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. No part of this publication maybe reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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Fig. 1. The nuclear-powered
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TABLE 2. Concentration of Actinides (g have been commissioned at the V. I. Lenin nuclear-power
/ton) in Irradiated Fuel from Thermal station at Leningrad, each with an installed power of 1 GW,
Reactors with Normal Water and Fast while the water-moderated water-cooled vessel reactor
Breeder Reactors [5] (type VVER) at the "50th Anniversary of the USSR" nu-
Thermalreactor. Fast reactor. clear-power station at Novovoronezh has been running
Isotopes burn-up33,000M burn-up 80,000 successfully for 10 years [2]. The power of the four units
(days/ton MW days/ton on this station comprised 1455 MW ate he beginning of
242Cm 8 42,3 1975. Five units with 1-MW type VVER reactors are
2"Cm 31 41,9 actively under construction.
u1Am 50 1460
243Am 92 711 The second unit at the Kola nuclear-power station
237Np 450 180 V
238pu 168 1840 was commissioned in 1974 with a type VVER-440 reactor,
249pu 5300 100
240pu 2140 52400 the total power of the first stage of this station being 880
241pu 1100 14400 MW. Construction has now started on the second stage,
235pu 340 9020 7 involving two type VVER-440 reactors.
236U 4 520 10
238U 940000 719000 Armyansk, Kursk, Chernoby'sk, Smolensk, and Kali-
nin nuclear-power stations (this is a far from complete
list of the stations constructed with types VVER and RBMK
reactors) constitute the basis for the development of nuclear-power engineering in the USSR in the near
future [3]. During the 20-year period of development of nuclear technology, there has been a significant
improvement in the cost effectiveness of both nuclear-power stations and nuclear-power engineering which,
as we have already noted, has become competitive with conventional power engineering. At a conference
on the planning, adjustment, and operation of nuclear-power stations, data were given on the high cost ef-
fectiveness of the stations. For example, the cost of electrical energy at Novovoronezh was 0.752 kopecks
per kWh in 1973 and 0.655 in 1974, which is below the cost of the electrical energy generated by the modern
steam power stations in the European part of the USSR [3].
Successful operation of the types VVER and RBMK reactors enables us to set about creating more
economical thermal reactors with unit powers of about 1.5 GW [1].
Experience with the operation of the Novovoronezh station has illustrated the reliability of the nuclear
fuel, ensuring an average depth of burn-up of 33.3. 103 MW days/ton U from the time of loading (7.974). It
should, however, be noted that this energy output arises not just from 235U; a significant contribution to-
wards the output of energy from a thermal reactor is made by a secondary fuel, plutonium, which is formed
from 238U during the operation of the reactor.
Table 1 gives the results of tests to determine the fuel burn-up in the elements of the two reactors at
the Novovoronezh nuclear-power station [2]. These results were obtained by the gamma-spectrometer
method without disrupting the fuel elements themselves. The errors in determining burn-up did not exceed
f 15%r.
Any increase in the duration of neutron irradiation of the nuclear fuel in a power-station reactor not
only leads to an increase in the total energy generation (burn-up of the original fuel), but also increases
the relative role of plutonium in the total generation of energy, i.e., reduces the specific expenditure of
235U per unit power.
If we compare the effectiveness of nuclear fuel and fossil fuel, and assume that 1 ton of conventional
fuel = 7000 ? 103 kcal, then with complete burn-up of nuclear fuel in thermal reactors [(35 to 40) -103 MW
days/ton of original slightly enriched uranium], a ton of fuel elements generates energy equivalent to 100-
120 ' 103 tons of conventional fuel. With existing energy intensities of the fuel, such a burn-up would be
achieved within three years on a power station, an increased burn-up being limited by the cost of the fuel
elements.
In fast reactors, which employ other forms of coolant, the specific energy intensity is significantly
higher, and a considerably greater burn-up can be achieved more rapidly. In thermal reactors, even with
very high burn-ups of uranium and plutonium, the energy output due to fission of the plutonium,nucleus does
not exceed the energy developed by fission of the nucleus of natural uranium (Table 2), whereas in fast
reactors, the breeding of nuclear fuel is possible. Scientists of the FEI, under the direction of A. I. Leipun-
ski i, have already shown in 1948-1949 that it is possible to involve not just 1 or 20/,.of mined uranium in the'
power-generating program, but ten times this, as by a combined use of uranium and plutonium it would be
possible to convert practically the whole of the 238U into fissionable plutonium. A study into the creation of
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TABLE 3. Characteristics of Isotope Energy a physical and engineering basis for fast reactors was
Sources carried out in 1954, starting with the BR-1 reactor, fol-
Specific (Half lowed by a series of experimental reactors: BR-2, BR-3,
e_ u M
and BR-5. In order to work out a design for the active
life,
energy, ~ ??v a ?~ t!
zone of the reactors and study the economics of the fuel
N N U.
1 i~g
Yr (IDi -, Fa ai n? `y S ~ 3
elements, an experimental reactor type BOR-60 was built
at Dimitrovgrad, on which experiments were success-
Zi6Po 144,0 0,378 0,5 5,401 31,4
242Cm 122,5 0,445 0,5 6,213 27,6
fully carried out in 1969.
'92Ir 59,7 0,204 0,2 1,1 154,0
i44Ce 26.7 0,78 1 1,409 120,0
At the present time, a nuclear-power station is
6?Co 17,5 5.26 5 2,607 65,0
227Ac 14.85 21.7 10 34,332 4,94
being built at Shevchenko, equipped with a fast reactor
i?oTm 11,35 0,354 0,4 0,321 525,0
type VN-350,designed to generate 150 MW of electrical
232U 5,0 73,6 10 40.174 4.2
244Cm 2,89 17,9 10 5,895 28,6
power and desalinate 120,000 tons of sea water a day.
9OSr 0,936 27,7 10 1.1 154,0
238Pu 0,58 86.4 10 5.59 30,3 A fast reactor type BN-600 is being built. Experience
i37Cs 0,411 29,7 10 0.786 215,0
gained in operating these reactors should define the ad-
i47Pm 0,338 2,65 3 0,062 2725,0
3H 0,36 12,26 10 0.019 2,7.104
vantages and drawbacks of the chosen equipment schemes
and active zones, provide a more-accurate breeding fac-
tor, and supply all the data needed to create more-econo-
mical fast reactors, the main type of reactor for meeting
future power demands [4, 1].
The book "From Scientific Research to an Atomic Industry" [6] and the jubilee handbook, published in
connection with the 20th anniversary of the commissioning in the Soviet Union of the world's first nuclear-
power station [7], give many physical and engineering characteristics of practically all types of power
reactors in the USSR, whether existing, under construction, or still in the planning stage. Successful opera-
tion of the icebreakers Leningrad and Arktika (Fig. 1) illustrates the wide possibilities of employing nu-
clear reactors with the fleet, in transport roles [7].
Scientists and engineers involved in the manufacture and reprocessing of fuel elements (after re-
moval from reactors) solve problems which are concerned not just with radioactive wastes, but also with
the fission products that are produced in large quantities and are used in nuclear engineering as isotope
sources.
As can be seen from Table 2, for each unit weight of fuel element submitted for radiochemical pro-
cessing from fast reactors with a burn-up of 80,000 MW days/ton, the concentration of plutonium isotopes
is increased by a factor of ten, while the concentration of the transpiutonic elements is increased multi-
fold.
The competition of the original fuel for the fuel elements used in fast reactors includes significant
concentrations of the isotopes of plutonium in addition to the 238U. As an example of this, we could note
that the concentrations shown in Fig. 2 of actinides produced during irradiation of fuel in a fast reactor
include for each ton of uranium and plutonium, in addition to the 130 kg of 239Pu, the following amounts of
short-lived plutonium isotopes (in g/ton): 238Pu = 2590; 240Pu = 51,800; 241Pu = 26,000. Therefore, when
manufacturing fuel elements for fast reactors, the total activity of the original fuel can be very high in-
deed.
The design data given in Table 2 has been obtained under the conditions that exist in thermal reactors,
in which the original enrichment of 235U comprised 3.3%, while the average specific power comprised 30
MW/ton. The amount of fuel recovered in the cycle comprises 1/3 of the annual fuel loads. Fast reactors
use 238U (78%) as fuel together with isotopes of plutonium; the average specific power is 148 MW/ton, while
1/3 of the fuel load is transferred after 153 days.
The use of plutonium in nuclear-power engineering and the consequent radiation safety problems that
arise have formed the subject of several articles [8], so that there is no point in dwelling on these topics
in the present article.
The increased rate of growth of nuclear-power engineering also accelerates the rates of mining and
processing uranium ores, which at the present time are the basic source of nuclear fuel.
For processing these ores, Soviet scientists and engineers have developed technical processes which
enable us to extract other.elements in addition to uranium, such as phosphorus, molybdenum, the rare
earth elements and other minerals that are valuable to the national economy [9]. Soviet scientists have
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 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070002-9
TABLE 4. Characteristics of Nuclear Ex- presented papers on the complete utilization of uranium
plosions in Rock Salt [17] ores, not just to Soviet conferences, but also to interna-
L 'Radius of
Di stributionof
Yield,
Pepth of
chambe
fissuring above
kton
bharge, m
r
form ed, m
chamber, m
1,1 I
16i
13,0
83
3.4
365
18,7
60
5.3
828
1714
64.
25.1
600
. 32,0
-
tional conferences [10]. Complete utilization of lean uran-
ium ores enables the cost of the uranium to be brought
down and, in the case of uranium-phosphorus ores for ex-
ample, enables us to extract a larger amount of the rare
earth elements which are used in many fields of industry,
together with valuable phosphorus fertilizers. The con-
tent of foodstuffs in fertilizers is 40-50%. The rare earth
elements are used as catalysts for cracking oil and as
alloying agents for cast iron and steel, while such ele-
ments as europium and yttrium are used in the manufac-
ture of color picture tubes.
The main requirements for the material of the active zone of a reactor have compelled specialists in
the atomic industry not only to create a series of new special alloys, but also to arrange their mass pro-
duction. Such materials as zirconium and its alloys, special materials containing neutron absorbers, and
others are widely used at the present time in nuclear-power engineering and other fields of the national
economy.
Nuclear fuels are supplied for the development of nuclear power in various countries, and in the form
of fuel elements to the COMECON countries and Finland.
Isotopes in the National Economy of the USSR
Not all the radioactive products separated after radiochemical treatment of the spent nuclear fuel are
disposed of by reliable burial. Some of them can be used in the national economy. This applies not only to
the fission fragments, radioactive products such as 90Sr, 137CS 144Ce, 147Pm, but also to isotopes specially
obtained by neutron bombardment, such as 60Co, 99Mo, 170Tm, and 210Po when,. as a rule, lighter isotopes of
these same elements are used as targets. The transuranic elements comprise a separate group of radio-
active elements.
The use of radioactive isotopes in the national economy of the Soviet Union is expanding every year.
It is now difficult to name any branch of science, technology, industry, agriculture, medicine, etc. in which
these true friends of man do not operate. The technology and reliable production capability for the majority
of radioisotopes and stable isotopes of practical use have now been developed in the Soviet Union.
The production list of the Isotop All-Union Combine exceeded 3300 in 1975, including:
Compounds of radioactive tracer isotopes ....................1034
Compounds with stable isotopes ............................. 718
Sources of a, 0,y, and neutron radiation ...................... 1622
This product list is based on 156 radioactive and 240 stable isotopes [11, 121.
By employing such a emitters as 210Po, 238Pu, and 239Pu a large number of sources can be produced for
use as radioisotope static neutralizers in the textile, printing, rubber, photographic-film, and other indus-
tries. Radioisotope signal fire-fighting equipment is widely employed in the USSR, in which isotopes of
plutonium are used in indicator amounts. The total output of these devices in 1975 was 300,000 units.
There are 246 types of 0-radiation sourcesbasedon 11 isotopes, including tritium, 14C, 60Co, 85Kr, 106Ru,
144Ce, 147Pm , and 204T1. These sources are manufactured with awide range of activities from 0.015 micro-
curie to 300 curies.
There is a large assortment of sources of y and x radiation varying both in radiation power and
physical dimensions.
The isotopes of plutonium, 210Po and 252Cf,are widely employed in the manufacture of neutron sources.
Plutonium-beryllium sources have an intensity up to 5.107, polonium-beryllium up to 4' 108, and californ-
ium from 1.5. 107 to 1 ?109 neutrons/sec.
The use of radioactive isotopes can develop along three main directions:
1. As radioactive indicators (the tracer-atom method).
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2. As sources of penetrating radiation for use in radioisotope instruments for automatic monitoring
and regulation of industrial processes and as radioactive fuel.
3. As powerful sources of ionizing radiation for acting directly on materials to initiate various reac-
ions and changes in structure.
The tracer-atom method is a powerful research tool which can enable us to uncover many complex
processes taking place in metallurgy, in the chemical and oil industries, and machine manufacturing, to de-
termine the nature of the activity of plants, animals, and human beings, to obtain a picture of the move-
ments of ground water and rivers, and to define the regional distribution of agricultural pests. The method
is widely used in both science and technology, in which it can be used, e.g., to determine the rate of wear
in machines. There are 740 compounds used as radioactive indicators, marked by various radioactive iso-
topes, including 14C (300 compounds), 3H (125 compounds), 32P (36 compounds), 35S (50 compounds), and 36CI
(20 compounds).
Radioisotope instruments are extensively employed in the national economy. One of their main ad-
vantages in relation to monitoring and measuring instruments based on other principles lies in the absence
of any contact with the material or agent being tested. Radioisotope instruments possess high sensitivity
and a high speed of reaction, they are indispensable for determining the characteristics of explosion-proof
and flame-proof materials, chemically aggressive liquids and gases, and viscous or friable materials.
In the chemical and oil refining industries, radioisotope instruments enable us to exercise direct
monitoring and regulation of the level of acids and alkalis in nontransparent reservoirs, to follow processes
at high temperatures and pressures. to determine the concentrations of materials in solution, etc.
In metallurgy, radioisotope instruments have been successfully used to automate the process of
charging blast furnaces, check the level and overflow of metals, simultaneously observe the wear in the
linings of open-hearth furnaces, and provide direct monitoring of the thickness of rolled sections.
In the machine-manufacturing industry, radioisotope instruments make it possible to monitor the flow
of products, ensure the interlocking of plants, etc.
In the building industry, hydraulic engineering, and agriculture, radioisotope instruments are exten-
sively employed for measuring the density and moisture content of various materials and determining their
uniformity.
As we have already seen, the introduction of radioisotope neutralizers of static electricity has solved
the problem of dumping static charges that arise in various industrial processes and of preventing fires
during the transfer of inflammable liquids. Data exist to illustrate their high cost-effectiveness. The an-
nual saving achieved by the introduction of one neutralizer in the textiles, printing,and rubber industries is
from 1200 to 4000 rubles. The gamma flaw-detection method is used in several fields of industry, enabling
us to monitor the quality of welded joints in high-pressure boilers and lines, and to detect fractures in the
reinforcement rods of ferroconcrete structures, blowholes and cracks in metal parts and castings.
The percentage distribution of radioisotope instruments by branches of industry is as follows.
Machine manufacturing ................................... 16.6
Foodstuffs ............................................... 13.4
Metallurgical ............................................ 12.8
Mining .................................................. 11.7
Chemical ................................................ 11.1
Light industry ............................................ 9.0
Building ................................................ 8.5
Other fields of industry ................................... 16.9
Certainly, these figures might change due to renewal of the instrument catalog; they do, however, re-
flect clearly the range of uses of this form of atomic technology in the national economy.
When selecting isotopes for radiation sources, it is advisable to consider not only the type of decay
and the radiant energy, but also characteristics such as specific generation, projected period of service,
and the chances of obtaining large quantities of the isotope.
Table 3 gives some of the characteristics of the basis isotopes which are suitable for use as radio-
active fuel and other purposes. Bearing in mind the fact that isotopes with half lives from 100 days to 100
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 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070002-9
Fig. 2. Surface equipment for 50,000 m3
condenser created by an underground nu-
clear explosion.
yr are used as energy sources, only 14 isotopes are given,
all of which are available in large quantities [12].
When choosing a suitable operating life for the isotope
sources, it is necessary to take into account the radiation
stability and thermal stability of the chemical compounds,
together with those of the fuel ampoules, and the actual ser-
vice life to be expected of the device itself.
Table 3 gives data on 60Co,137Cs, 192Ir, 170Tm, as well
as 75Se, 241Am, and others which are used for monitoring
the quality of products made of steel and other materials by
the gamma flaw-detection method. This method has become
one of the principal methods of nondestructive testing of ma-
terials and manufactured products, and is suitable for a wide
range of thicknesses, from 0.5 to 200-250 mm steel equiva-
lent.
In recent years, radioactive isotopes have found wider application as powerful sources of ionizing
radiation in direct action on various processes and materials for the purposes of improving manufacturing
techniques and obtaining new materials and compounds to meet the needs of modern technology.
The heat resistance of polymer materials and products made from rubber can be greatly improved by
irradiation; the strength of wood can be increased and the quality of cotton fabric improved. The first in-
dustrial radiation-chemical plants have been set up and operated successfully at Volgograd, Kazan, Grozny,
Podol'ski, and other industrial centers.
In agriculture and the food industry, ionizing radiation offers the possibility of increasing yields,
accelerating the breeding of new types of plant, and increasing the storage life of food products. Tests car-
ried out under production conditions using presowing radiation methods and radiation treatment of products
have given encouraging results at a number of installations.
The use of manufacturing methods based on radioactive isotopes enable us to achieve significant eco-
nomies each year in many fields of the national economy.
In medicine, radioactive isotopes and nuclear radiation have been successfully used in the diagnosis
of complex illnesses, for the treatment of malignant tumors, for the radiation sterilization of materials,
instruments, and medicinal preparations.
The Izotop combine is currently supplying more than 35 radiopharmaceutical preparations to medi-
cal institutes in the Soviet Union and abroad, while more than 37 preparations are undergoing medical
trials. The isotopes 1311, 198Au, 32P, 133Xe, and 99Tc are of greatest use in radioisotope diagnostics and
therapy. About 650 institutes and medical establishments in the Soviet Union employ tracer preparations
based on radioactive isotopes, designed for medical or medicobiological purposes.
Stable isotopes such as 2H, 15N, 13C, and 180 are also employed in chemistry and biology for studying
the mechanisms of chemical reaction and the replacement processes of materials and living organisms.
Atomic science and technology enable us to create full-scale production of isotopes in the Soviet Union
by various methods and to expand the production base of radioactive and stable isotopes, tracer compounds,
instruments, and equipment using isotope radiation sources. Isotopes produced in the USSR are exported to
32 countries.
Employing the Energy of PSaceful Atomic Explosions
In recent years, several countries, including the USSR, USA, UK, and France, have studied the possi-
bility of using nuclear blasts for peaceful purposes. This problem has been studied at four conferences,
organized by MAGATE in 1970-1975 [13]. Specialists have showed conclusively that peaceful nuclear ex-
plosions, having exceptionally high concentrations and low specific energy costs, can be used to create new
processes in the construction and mining industries, and also for scientific investigations. These would
not present any danger to the population or the environment in the sense of seismic or radiation hazards,.
provided certain rules were observed [14].
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Fig. 3. Approximate arrangement of the
main works for transferring part of the
flow of the northern rivers into the Volga
basin [21]: 1_) Perchoro-Kolvinsk canal:
1) northern section (River Pechora to
Lake Chusovskoe); 2) southern section
(Lake Chusovskoe to River Kolba); o )
planned sites of dams.
Considerable interest exists within the USSR in the
future possibilities of using peaceful nuclear explosions for
carrying out large civil-engineering works or developing use-
ful mineral resources spread over vast expanses of terri-
to ry.
Depending on the degree of study involved, the techno-
logy of peaceful nuclear explosions can be divided into three
groups: industrial, experimental, and research.
The first group relates to such examples of industrial
or experimental-industrial uses of explosions as the elimina-
tion of dangerous gushers of natural gas, increasing the rate
of oil extraction, the creation of subterranean cavity reser-
voirs in solid rock salt and open reservoirs for storing wa-
te r.
Underground nuclear explosions have been successfully
used in the USSR to stop-up accidental natural-gas blowholes.
The flow from one of these reached 12 million m3/day, and for
a long time it was impossible to stop this by conventional
methods. A 30-kton nuclear explosion at a depth of 1550 m
permanently closed off the shaft of the blowhole [15, 16].
Experimental explosions have been set off to speed up
the production of oil. A long period of operation of the site
after three explosions had been set off showed that the prod-
uction rate was 27-60%greater than expected. due to the
creation of artificial fissuring [15].
An artificial reservoir for water, with a total capacity
of about 20 million m3, was formed by an excavating nuclear
explosion with a yield of over 100 kton; the visible crater had
a total volume of about 7 million m3.
The demand for reservoirs has greatly increased over
the last ten years due to the rapid development of the gas,
oil-refining, and other industries. The search for more-
effective methods of creating reservoirs has lead to trials
with camouflet nuclear explosions in solid rock salt. Due to
the elastoplastic properties of rock salt, it is possible in this
way to create large stable chambers (Table 4).
Figure 2 shows the surface equipment for a condenser reservoir created in the solid rock salt at a
depth of 1140 m by a nuclear explosion of about 15 kton. The chamber has a volume of about 50,000 m3 and
operates under a pressure of about 80 atm, created by gas from the gas line, the condensate being extracted
at the- surface [17]. Successful operation of this condenser reservoir illustrates its advantages compared
with metal reservoirs built on the surface: the cost of construction is less by a factor of 3-5 than a conven-
tional reservoir, the cost in metal is lower by a factor of 10-20, the construction time is several times less,
and the industrial land use is lower by a factor of hundreds. Furthermore, the cost of ensuring safe opera-
tion of the reservoir is greatly reduced [18].
Starting with a defined yield of explosion, this method is more advantageous, and at the same time
more universally applicable, than the method by which chambers are washed out of the solid rock salt.
Constructing reservoirs by means of nuclear explosions is more effective in regions of new industrial de-
velopment before the actual construction stage is reached, as in these cases more powerful explosions can
be used and the reservoir can be sited near to the factories themselves.
The second group relates to explosions which can be used in the construction of large hydraulic works
(canals) and for stripping areas of their useful minerals. Plans of this nature are known to include the
construction of the Pechorb-Kolvinsk canal in the USSR, the Orinoko-Rio Negro canal in Venezuela,and the
Kra in Thailand. It has been calculated that the use of underground nuclear explosions during the construc-
tion of these canals would considerably reduce the cost and the time needed for construction [19].
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 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070002-9
Fig. 4 Fig. 5
Fig. 4. Photograph of the development of group underground nuclear blasts on the route
of the Pechoro-Kolvinsk canal.
Fig. 5. "Bulat" equipment for producing wear-resistant coatings.
TABLE 5. Basic Characteristics of Mo- The article "Rational utilization and protection of wa-
bile Equipment for Neutron-Activation ter resources" [201 reports on large-scale works that have
Analysis [241 been carried out in the USSR to redistribute the pattern of
ti river flows by transferring water to regions with a deficit
' o o L on their water-economy balance. The advantages are
stressed of transferring part of the flow of northern rivers
9 g g into the Volga basin, to the extent of 20-25 km3/yr (40-50
H 5 z o. a ? e x w km3/yr in the future), which would be equivalent to increasing
NGI-1, 2 80-140 1.5-2 3.107 3.108 1-10 the flow of the Volga by 15-20%. The creation of the 112.5-
NGI-4 110 1-1,5 0.8.10' 1,5.108 1-33 km Pechoro-Kolvinsk canal (Fig. 3) basically reflects this
NGI-5 150 1,5 0,6.108 6.108 1-10 problem, and envisages the growing demand for water in
the central and southern regions of the European part of
the USSR being met, the level of the Caspian Sea being
stabilized, and the output of electricity from the Volga
hydroelectric cascade being increased. It is proposed to construct the northern section of the canal (65 km
long, useful crass-sectional area 3000 m2) by means of nuclear blasts. About half of this section passes
through a zone of flooded alluvial deposits. No experience in the creation of channels in this type of ground
is available, so there was considerable doubt as to mechanical effects of an explosion, the stability of the
banks of any channel formed, and the seismic and radiation effects. In order to study these problems, an
exceptional blast was set off under similar geological conditions [13, 22]. Three charges of 15 kton were
used in the experiment, in three boreholes at a depth of about 128 m. The distances between the boreholes
were 163.1 and 167.5 m. Figure 4 shows a general view of the development of a blast 5 sec after the
charges were detonated. The blast formed a channel 700 m long, 340 m wide, and from 10 tol5mdeep. The
sides of the channel were formed at an incline of 8-10%and these have shown practically no alteration with
time. Thus, for the first time it was shown to be possible to form large channels with sufficiently stable
banks in a thick cover of weakly flooded ground [221.
A project has been developed for the intensification of mining for useful minerals using nuclear
blasts. In the project, large deposits of light metals on the northeast frontier of the Soviet Union can, ac-
cording to preliminary estimates, be stripped by open-cut mining techniques of up to 900 million m3 of
rocky ground [15].
The third group contains the most complex fields of use of nuclear blasts: the distillation of oil from
shales, the leaching of copper from deposits, the release of geothermal energy, the creation of reservoirs
for hazardous and radioactive waste products, etc. Plans exist for the utilization of blasts for scientific
purposes [13].
Peaceful nuclear explosions represent a new field of application of atomic energy, requiring further,
more detailed investigation. There are grounds to suppose that their potential is neither fully realized nor
exhausted.
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 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070002-9
U
CIS v>
Power in beam,
kW
w E
C
o-
Weight,
o
` ,.
av.
impulse
o
a'0
5i
E E a
c6
Size, m
tons
T
U
G N
U U
7.Y o .
Ow w E
N N
Q n
5
RTD -1
Resonant transformer
1,0
3
18
60
0.2.5
910,0;:.1.5
1,9
)lektron-1
Transformer
0,7
7
7
-
y 0,7 x 3.0
110
9LIT-500
The same
0,5
1
700
18
-
0.3 x 0.5
0.04
IT-1
? ?
1,0
8
10000
360
-
00.4x0.6
0.12
IT-3
>.,>
2,5
10
40000
01,0 x1.3
0,8
ELT,-2
? ?
1,5
25
215
-
-
01,3x 2.4
7,0
KGE-2,5
Cascade generator
2,5
20
20
-
-
, 3,0 x 6.2
3210
LU9-8-5V
Linear accelerator
8,0
5
3500
-
-
5,Ox0,7x1,75
2,0
(emitter)
LU9-13-9
The same
13,0
9
11000
-
-
5,5x1.5x2,35
5,0
(emitter)
LUE-10-1
10,0
1
1000
2000
1,5
2,75x1,0x0.8
2.0
(e
LU E-15
15,0
1,5
1500
10000
2,0
4.5x1,5x2,0
5,5
(emitter)
B-25
Betatron
25,0
-
-
40
-
-
2,5
B-35
The same
35,0
-
-
250
-
-
5,0
B-50
50,0
-
-
800
-
-
20,0
Fig. 6. Implantation equipment type
ILU-4 for surface alloying.
The problem of disposing of radioactive wastes is of great significance to the development of the
atomic industry. One of the most promising ways of solving this problem, according to the results of many
years experience, is by the disposal of solutions containing radioactive matter in underground absorbent
beds in the Dimitrovgrad region [23]. The results of this trial were put to partial use on factories of the
USSR Ministry of the Chemical Industry.
Absorbent strata have been found in the vicinity of several chemical combines and turned, without the
use of nuclear blasts, into proving grounds for underground burial of concentrates from chemical treatment
of toxic materials. Table 4 gives the volumes of underground reservoirs produced by means of experimen-
tal nuclear blasts. These give us reason to suppose that in regions where there are no underground absor-
bent strata it would be possible to create underground chambers by means of nuclear blasts for burying
dangerous and radioactive wastes. This supposition stems from hydrogeological and engineering-geological
questions concerning the radiation possibilities of the underground vaults used for industrial waste prod-
ucts.
The development of nuclear physics, plasma physics, solid-state physics, and atomic materials has
not merely created a scientific foundation for atomic engineering, it has also actively facilitated the intro-
duction of the so-called "fall-out products" or by-products of these branches of science in the national
economy. For example, until quite recently, charged-particle accelerators have been used only for re-
search into nuclear physics. As their design improved, however, it became clear that accelerators could
be employed in various branches of the national economy. In industry, agriculture, and medicine charged-
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Fig. 7. An atomic building in the USSR. A general view of the construction of the
Armyansk nuclear power station. Photograph by V. Bratchikov.
particle accelerators are finding even wider applications, the accelerator developed by the D. V.
Efremov NIIEFA, for example.
Activation analysis is widely used in geological surveying and metallurgy for determining the con-
tents of various elements. Neutron-activation methods employing accelerators have been used to analyze
contaminants during the production of very-pure structural materials. The test material is irradiated by
neutrons formed by bombarding special targets with accelerated electrons. The neutrons activate contami-
nants present in the test material, forming short-lived isotopes, by the study of which it is possible quickly
to determine the presence of certain elements with a sensitivity measured in hundred-thousandths of a
percent. The time needed for analysis is greatly reduced and fewer laboratory technicians are needed.
Table 5 gives the characteristics of a number of devices used in the national economy.
Neutron generators of type NG-150 are used in stationary laboratories. These have a flux of about
1011 neutrons/see.
In addition to neutron generators, y-radiation sources, hydrogen-ion, deuterium-ion, and helium-ion
accelerators are also used for activation analysis.
A great deal of attention has been devoted during the last ten years to the problems of designing
accelerators for use in flaw detectors and radiation engineering processes, and also in medicine. Such
accelerators have a number of advantages over cobalt sources:
a. They enable us to obtain bremsstrahlung having a higher penetration and with doses in the hundreds
of rads/min ? m at a distance of up to 1 m from the target.
b. Besides brehmsstrahlung, we can obtain beams of electrons, protons, mesons, etc. with energies
ranging from hundreds of kiloelectrovolts to hundreds of megaelectronvolts.
c. They enable us to form uniform dosage fields with the clean boundaries needed for therapy, for
example.
d. They ensure radiation safety as far as auxiliary operations are concerned; e.g., there is no need
to periodically reload and bury spent highly radioactive charges.
Radiologists throughout the world are especially interested in the linear electron accelerator, as this
is very simple in design compared with other types of accelerator, the introduction and extraction of parti-
cles is very simple, and the energy and power of the radiation doses can be regulated. They can produce
powerful doses (not just at high energies but over the whole energy range) in large highly uniform fields.
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Linear accelerators have been created for flaw detection at energies of 6-9 MeV with bremsstrahlung
intensities of 300-1500 R/min and an energy of 15 MeV with intensities of up to 10,000 R/min (types LUE-
10-1, LUE-15-1.5). Flaw detector accelerators are able to penetrate steel products to depths of more than
400 mm, due to their high power and the penetrating ability of the x rays they generate. Linear accelera-
tors for flaw detectors with energies of 6-9 MeV have been successfully operated at the Izhorsk factory at
Kolpino, enabling the test time to be reduced by a factor of 10-12 compared to the cobalt equipment used
previously, and increasing the visibility of flaws in product thicknesses of more than 150 mm.
Linear accelerators have been manufactured for sterilizing medical instruments, enabling the degree
of sterilization to be increased; expendible medical instruments to be put into mass production, ready for
immediate use (hypodermic syringes, catheters, blood-transfusion systems, etc.); cheap plastics, which
are unable to withstand thermal sterilization, to be used for the manufacture of instruments; antibiotics to
be sterilized which are unable to withstand any method of sterilization other than radiation. Two linear
accelerators have been built for the medicinal preparations factory at Kurgan.
Linear accelerators are now being built with energies in the 5-30 MeV range, for treating cancers by
braking (x ray) radiation and electron beams. Such accelerators have been successfully used in medical-
research institutes at Moscow, Kiev, Minsk, and Obninsk. The use of linear accelerators for treating
cancer shows that they possess greater biological effectiveness than cobalt radiation sources.
Experience of operation (the accelerator at the Central Institute for the Improvement of Medicine has
been operating since 1966) and the achievements of accelerator technology have enabled a new improved
therapeutic linear accelerator to be built with an energy of 15 MeV, which has good characteristics, is easy
to control, and is designed for quantity production [24].
The Tire Industry Research Institute has operated a linear electron accelerator manufactured by the
NIIEFA with an energy of 7-8 MeV and a power of 3.5-5.0 kW. The accelerator is designed for studying the
technology of radiation vulcanization of rubber tires with the aim of increasing the service life of automo-
bile tires. Table 6 gives the basic types of accelerators used in various fields of the national economy.
Furthermore, intensive beams of heavy ions can be used to produce materials with new properties,
study radiation damage in materials, and carry out tests on structural materials relatively quickly.
By irradiating thin films with heavy ions it is possible to produce nuclear filters having apertures
with diameters from 20-40 A up to several hundreds of microns. The Joint Institute of Nuclear Research
(JINR) has developed the technology for producing these films and for creating special equipment. Nuclear
filters are now being tried out in a number of organizations, as their fields of application are diverse.
Thermonuclear investigations being carried out at the I. V. Kurchatov Institute of Atomic Energy
(IAE) and the KhFTI have also found applications in the national economy. The KhFTI have developed a
plasma method of obtaining materials with new properties by condensation of matter from the vapor phase
in a vacuum while at the same time being bombarded with ions. This method is based on the properties of
a low-voltage electric are in a vacuum. As the arc burns, the material of the cathode is vaporized. At the
same time, the cathode effect creates a rapid flow of the plasma formed during ionization of the cathode.
The substrate (component, instrument, or material), on which it is desired to deposit the coating, is
held at a negative potential, so that the plasma ions bombard a layer of condensate on the substrate. The
KhFTI has developed on industrial equipment type."Bulat" (Damask steel) which produces high-tempera-
ture wear-resistant coatings and materials with good physical, technical, and mechanical properties. Tests
have shown that durability of a cutting tool reinforced in this way is 2.5 to 5 times greater on the average.
At present, the technology of reinforcing and hardening metal-cutting tools by the ion-bombardment-in-a-
vacuum method is being introduced on a number of factories in various fields of industry [25].
In the I. V. Kurchatov Institute of Atomic Energy implantation equipment type ILU-4 has been de-
veloped for the surface alloying of semiconductors and other materials by the ion-bombardment method. The
equipment is in quantity production and is being used for semiconductor studies and investigation of radia-
tion effects in solid bodies [26]. About 50 such devices are now in successful operation, some of them in
Bulgaria, Hungary, and the German Democratic Republic.
1. V. A. Kurillin, Priroda, No. 3, 6 (1975).
2. Ten Years Experience in the Operation of the Novovoronezh Atomic Power Station [in Russian],
Novovoronezh (1974).
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3. L. M. Voronin and E. Yu. Zharkovskii, At. Energ., 38, No. 2, 113 (1975).
4. A. I. Leipunskii, At. Energ., 28, No. 4, 297 (1970).
5. F. Girardi, G. Bertesi, EUR-5214, Ispra (1974).
6. A. M.Petrosyants, From Scientific Research to an Atomic Industry [in Russian], Atomizdat, Moscow
(1972).
7. Atomic Engineering 20 Years On [in Russian], Atomizdat, Moscow (1974).
8. S. V. Bryunin, At. Tekh. za Rubezhom, No. 6, 20 (1974).
9. Soviet Atomic Science and Technology [in Russian], Atomizdat, Moscow (1967).
10. A. P. Zefirov, Questions of Atomic Science and Technology [in Russian], in The Enrich-
ment and Metallurgy of Uranium, No. 1, Part 2, TsNIlatominform,,Moscow (1971).
11. "Forms of isotope production supplied by the 'Isotop' combine and some of their fields of applica-
tion," paper' V/O Isotop on its seminar in Budapest (April 1975).
12. In: Production of Isotopes [in Russian], Atomizdat, Moscow (1973).
13. "Peaceful nuclear explosions," Byull. MAGATE, No. 2, April (1975).
14. Yu. A. Izrael', Peaceful Nuclear Explosions and the Environment [in Russian], Gidrometeoizdat,
Moscow (1974).
15. I. D. Morokhov (editor), Atomic Explosions for Peaceful Purposes [in Russian], Atomizdat, Moscow
(1970).
16. 0. L. Kedrovskii et al., in: Peaceful Nuclear Explosions II, Vienna, IAEA (1971), p. 209.
17. K. V. Myasnikov et al., in: Peaceful Nuclear Explosions III, Vienna, IAEA (1974), p. 179.
18. O. L. Kedrovskii et al., in: Peaceful Nuclear Explosions IV, Vienna, IAEA (1975), p. 227.
19. I. D. Morokhov et al., At. Energ., 39, No. 2, 148 (1975).
20. I. Borodavchenko and 0. Tolstikhin, Kommunist, No. 14, 42 (1975).
21. "Ideas on the peaceful use of nuclear explosions in the USSR," Byull. MAGATE, No. 2 (1970).
22. V. V. Kireev et al., in: Peaceful Nuclear Explosions IV, Vienna, IAEA (1975), p. 399.
23. M. K. Pimenov, in: Investigations in the Field of the Disposal of Liquid, Solid, and Gaseous Radio-
active Waste and the Deactivation of Contaminated Surfaces [in Russian], Proceedings of a Scientific
and Technical Conference of the SEV, Poland, Kolobzheg, October (1972), pp. 2-7.
24. V. A. Glukhikh, Reprint NIIEFA, P-0165, Leningrad (1972).
25. The Physics and Use of Plasma Accelerators [in Russian], Nauka i Tekhnika, Minsk (1974).
26. V. M. Gusev et al., Pribory i Tekh. Eksperim., No. 4, 9 (1969).
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EXPERIENCE IN THE CONSTRUCTION OF LARGE
POWER REACTORS IN THE USSR
N. A. Dollezhal' and I. Ya. Emel'yanov UDC 621.039.577
In the directives of the 24th Congress of the Communist Party of the Soviet Union, among the problems
related to the development of power generation in the country, reference was made to the need to put into
operation during the 1970-1975 period the Leningrad Atomic Power Station (LAPS), which will have a capa-
city of 2,000,000 kW. This means a very substantial introduction of nuclear-fission energy into electrical
power generation in the coming years. The task assinged by the directives was fulfilled: The first reactor,
with an electrical power of 1,000,000 kW, began operation in December 1973 and achieved full power by the
57th anniversary of the October Revolution; the second reactor, identical with the first, went into operation
in August 1975, and by December 1975 its power output was approximately 900 MW. The second reactor
will also undoubtedly achieve full designed power in the near future. The LAPS. named for the great Lenin,
is equipped with channel-type uranium-graphite reactors (RBMK), which have been described repeatedly
in the literature [1-4]. The theoretical and engineering principles of this reactor were developed and
tested in practice in the Soviet Union, and therefore it can rightfully be regarded as a Soviet type of reac-
tor. The construction and introduction into operation of the LAPS reactors means that one more important
landmark has been passed in the process of improving and developing this type of reactor, the conception
of which dates from the late 1940's, when the first such reactors were constructed, including the reactor
of the world's first atomic power station at Obninsk. The next landmark was the start-up in 1958 of the
reactor of the Siberian Atomic Power Station, a film of which was shown, in particular, to the participants
in the 2nd International Conference on the Peaceful Uses of Atomic Energy, held at Geneva in the same
year. After this, in 1964, the channel-type uranium-graphite reactor of the I. V. Kurchatov Atomic Power
Station at Beloyarsk was put into operation. In 1967 a second reactor, with an electrical power of 200,000
kW, was started up at Beloyarsk. These reactors, based essentially on the same technological idea, differ
in principle in that, first of all, their fuel channels are cooled by boiling water and, second of all, the steam
generated is superheated in special channels in the same reactor. More than 12 years of operation of the
reactors of the Beloyarsk APS have confirmed the viability of such a solution. It should be noted that such
solutions, carried to the point of satisfactory results, do not exist in any other country in the world, despite
many attempts that have been made. Nuclear superheating of steam before it enters the turbine is a very
tempting idea, since it not only prevents the danger of wet steam entering the turbine but also makes it pos-
sible to do without the intermediate moisture separators and without superheating of the steam between
turbine stages, which in turn makes it possible to simplify the production of steam in the reactor. For tur-
bines with large power, e.g., over 800 MW, which may be required in the construction of large atomic
power stations, the initial superheating of the steam makes it practical to use a speed of 3000 rpm instead
of the 1500 rpm now used in operation with saturated steam. Another significant fact is that the efficiency
of the entire installation is improved.
The reactors of the LAPS, like those of a number of other APS now under construction, have no
channels for the superheating of steam and produce dry steam, obtained in channels with boiling water in a
single-loop scheme and transmitted directly to the turbine. As a consequence of this, it is necessary to in-
clude between the stages of the turbine a number of moisture separators and steam superheaters. The next
stage in the improvement of channel-type uranium -graphite reactors will undoubtedly be the introduction
into the active zone of channels for the superheating of the steam. This will come with the next generation
of reactors. They will have an electrical power of 2-3 million kW; the reliable operation of such reactors
will be based on the experience obtained in the operation of reactors now under construction, with the com-
plex physics of their large active zones and with the use of modern computers for detecting at the proper
Translated from Atomnaya Energiya, Vol. 40, No. 2, pp. 117-126, February, 1976. Original article
submitted November 14, 1975.
?1 9 76 Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. No part of this publication maybe reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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Fig. 1. Theoretical thermal scheme of the LAPS: 1) ventilation stack; 2) set gas-
holder; 3) storing gasholder; 4) helium-cleaning unit; 5) KGO system; 6) monitor-
ing of integrity of technological channels; 7) compressor; 8) gas-loop condenser;
9) SUZ pump-heat exchange installation; 10) separators; 11) regulating assem-
blies; 12) fuel channel; 13) SUZ channel; 14) TsVN-7 pump (4 pumps); 15) pre-
cooler; 16) regenerators; 17) cooling pumps; 18) separators; 19) APN tanks; 20)
emergency feed pump; 21) compressed air; 22) emergency cooling system of the
reactor; 23) bubblers; 24) technological condensers; 25) separator-superheaters;
26) TG-1 turbogenerator; 27) TG-2 cleaning; 28) condensers; 29) KN-I conden-
sate pumps; 30) condensate cleaning; 31) units for ignition of explosive mixture; 32)
KN-1I condensate pumps; 33) low-pressure preheaters; 34) deaerators (7 atm); 35)
electrical feed pumps.
time the phenomena going on in the active zones and the effects on the automatic control devices. We must
believe that scientific and technical progress in the next few years, particularly in the fields of metallurgy,
physical chemistry, and instrument design, will lead to the realization of these ideas as early as the next
5-10 years.
The theoretical thermal scheme of the V. I. Lenin APS is shown in Fig. 1. It is a single-loop
scheme, which differs from the known schemes for boiling reactors. The difference lies only in the design
of the reactor, in the present case an RBNK-1000 channel-type uranium-graphite reactor. In estimating
the advantages of this type of reactor, the following considerations are weighed: the existence of extensive
experience in the construction andoperation of such reactors; the absence of any specific and new tech-
nological processes, so that it is possible to get orders filled by the machine-construction industry without
unduly great expense for the retooling of factories, and consequently, without requiring very long delivery
times; the possibility of constructing reactors of any dimension by using mass-produced elements and
assemblies, i.e., there are practically no limitations on the increase of the unit power of the channel reac-
tors; the structural separation of the moderator and the coolant, making possible a fairly flexible choice
of the substances and materials used for them, thereby ensuring effective heat removal in the active zone,
with good neutron balance; the possibility of recharging an operator reactor with fuel without reducing the
power, thus improving the economic indicators of the atomic power station, since in this case there is
practically no need of a reactivity excess for fuel burnup; the simplification of the system for monitoring
the condition of each channel, and the possibility of operational replacement of fuel assemblies which have
developed leaks; the fact that the cooling loop of the reactor consists of many smaller loops of small-
diameter pipe, improving the safety of the installation; the possibility of easily adapting the reactors to the
conditions of the fuel market; the possibility of continuously introducing new structural elements and as-
semblies, with the use of the most modern advances in nuclear-fuel and reactor-material production tech-
nology; the convenience and simplicity of introducing nuclear superheating of steam into the scheme of the
APS.
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Fig. 2. Overall view of the LAPS in operation. Photo by V. Volkov.
The development of channel-type uranium-graphite reactors is inextricably connected with progress
in the technology of reactor materials. The problem that had to be solved was that of constructing a power
reactor with a satisfactory fuel cycle and, at the same time, a fairly satisfactory efficiency for the power
station as a whole. This required, above all, new structural materials which would retain their strength to
high temperatures and would have a low cross section of neutron absorption. Such materials - zirconium-
based alloys for the structure of the channels - were produced. As in the case of the Beloyarsk APS, a
one-loop thermal scheme was adopted for the RBMK reactors, and ordinary boiling water was selected as
the coolant. This solution was based on many years of experience with the operation of boiling reactors.
The cylindrical stacking of an RBMK reactor consists of individual graphite columns with axial cavi-
ties which contain the fuel channels and the SUZ channels. A fuel channel is a tubular construction whose
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Fig. 3. Machine room of the TAPS.
central position, situated in the active zone and made of ?a zirconium alloy, is joined to the upper and lower
parts, which are made of stainless steel, by means of special adapters. The fuel channel contains a cas-
sette with two heat-generating assemblies, each of which consists of 18 fuel elements. A fuel element is a
zirconium-alloy tube measuring 13.5 x 0.9 mm and filled with pellets of uranium dioxide. The active zone,
11.7 m in diameter and 7 m high, contains about 1,700 fuel channels with 195 SUZ channels.
The coolant is water introduced from below into each channel, heated to the boiling point, and partly
vaporized in the active zone. The resulting steam -water mixture is removed from each channel into the
separators. The saturated steam at a pressure of 70 atm is directed to two turbines, rated at 500,000 kW
each, and the separated water, mixing with the feed water, is delivered by the main circulation pumps to
the inlets of the channels through a system of distributing collectors. The reactor is equipped with: a con-
trol and protection system which, on predetermined signals, takes the reactor to different power levels,
until it has reached the subcritical state; a system of physical monitoring of the distribution of energy
generation with respect to the height and radius of the active zone; a system for monitoring the tightness
of the seal of the fuel-element jackets; a system of channel-by-channel monitoring and regulation of the
coolant flow rate; a system for monitoring the integrity of the channels in the reactor. Because there are
so many parameters to be monitored, an automatic system of centralized monitoring is used, making it
possible to measure and record the parameters of each block. The system includes a digital computer for
processing the information and for the operational calculation of a number of parameters that are important
in the running of the reactor.
The working design for the RBMK reactor was completed in 1969, and the factories began to construct
the reactor in the same year. Construction work on the site was begun in 1968, and the installation of the
equipment began in March, 1971. On September 10, 1973 the first heat-generating assembly was charged
into the reactor, and physical startup was begun. The construction of the world's largest channel-type
nuclear reactor in such a short time was possible because the engineers and technicians had behind them
more than 20 years of experience in the successful operation of such reactors. The physical start-up pro-
cess included charging of the channels with fuel assemblies and rods of additional absorbers. At a number
of specified intermediate states, the reactor was brought.to criticality in order to carry out experiments
in determining its neutron-physics characteristics. As a result, the initial charge of the active zone was
formed, the reactivity effects.and the effectiveness of the control rods were determined, and recommenda-
tions were worked out for the manner in which the control rods should be withdrawn when the reactor was
brought up to power.
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Fig. 4. Discharging and charging machine for the RBMK-1000.
From November 14 to December 21, 1973 power start-up on the first block of the LAPS was carried
out and the process of bringing it up to rated power was begun. In the earliest stages, the criterion for
safe operation of the active zone was the condition that the possibility of a heat-exchange crisis had to be
precluded in the channel with maximum power and minimum water flow rate. The power of the block was
increased by consecutive stages to 500 MW, and in the spring of 1974, after a second turbogenerator was
connected, the power was increased to 600 MW. During this period special attention was paid to the in-
vestigation of the energy generation fields in their active zone and to their equalization and stabilization.
For monitoring the state of the active zone before the regular, system was brought into operation, the LAPS
staff used a complex.of programs of physical and heat-engineering calculations, prepared for an external
computer but using as the initial data the readings of the energy-generation monitoring sensors, flow-rate
meters, control-rod position indicators, etc.
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Fig. 5. Arrangement of equipment in the RBMKP-2000 (cross section).
The calculations were used in determining the distribution of power and excess values before a crisis
in the reactor channels; the coolant was distributed in the channels in accordance with the power values.
In July 1974 the block was raised to a power of about 800 MW. At this power level the staff finally adjusted
and put into operation the regular system of operational monitoring of the state of the active zone, making
use of design algorithms for the calculation of excesses before a crisis and the calculation of heat-engi-
neering reliability. On November 1, 1974 the leading block with the RBMK reactor was brought to the
nominal power value of 1 million W.
All the basic parameters of the reactor and the block as a whole agreed with the design values.
The equipment of the second block of the LAPS was installed two years later. In May 1975 charging
of the reactor began and physical start-up took place, and the block was brought to a power level of about
800 MW by October. Figure 2 shows an overall view of the V. I. Lenin APS at Leningrad in operation.
Figure 3 shows an overall view of the machine room, with four K-500-65 turbines. The experience ob-
tained in the process of starting up the first block made it possible to reduce to less than half the length
of time required for the main stages of the process of bringing the second block up to power. One may
expect that the period of start-up and adjustment operations can be shortened even more. For this it will
be necessary to formulate typical programs optimized on the basis of the results obtained in the adjustment
and startup of the first blocks, as was done, for example, in the program entitled "Physical startup of an
APS with reactors of the RBMK type." The amount of start-up and adjustment work must be determined by
tests of the installed equipment and complex testing programs, and obviously only those of the investigative
operations should be used which, during the work on the previous blocks, yielded results which are un-
acceptable for one reason or another in the given case.
The first year of operation of the LAPS has confirmed the high efficiency of the reactor and the main
equipment of the station. On the basis of the results obtained in the start-up and adjustment operations,
necessary changes were made in the design of some assemblies, the technological scheme, and the regimes
of operation, corrections are being made in the design materials, and steps designed to improve the char-
acteristics of subsequent blocks using the RBMK reactor are being developed and carried out. During the
period of operation, reactor shutdowns were due mainly to the need for removing from the reactor the
additional absorbers at the proper time and charging additional fuel into the active zone. After the process
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Electrical power, MW
1000
1500
2000
Thermal power of the reactor, MW
3200
4800
5400
Efficiency,
31.3
31.3
37.0
Active-zone dimensions, m:
height
7
7
6
diameter (or width and length)
11.8
11.8
7.75 x24
No.of channels:
vaporizing
1744
steam -superheating
872
Uranium charge, tons
192
189
226
Enrichment,
1.8
1.8
1.8/2.2
Av. uranium burnup in discharged channels, MW days /kg
vaporizing channel
18.1
20.2
steam-superheating channel
18.9
Dimensions of fuel-element jackets (diameter x thickness), mm:
vaporizing channel
13.5 x 0.9
steam -superheating channel
10x0.3
Material of fuel-element jackets:
vaporizing channel
Zirconium alloy
steam -superheating channel
Stainless steel
Water flow rate through reactor, tons/h
37,500
29,000
39,300
Pressure in separators, arm
70
70
85
Steam capacity of reactor, tons/h
5800
8800
8580
Steam flow rate of turbine, tons/h
5400
8200
7580
Parameters of steam before turbines:
pressure, arm
65
65
65
temperature, ? C
280
280
450
of recharging the channels on the operating reactor by means of a recharging machine (Fig. 4) has been
set up, the number of shutdowns will be determined by the graph for conducting planned preventive over-
hauls.
To sum up, we can list the main problems which were successfully solved in the process of con-
structing the RBMK channel-type uranium-graphite reactor:
1. Zirconium alloys - the principal structural material for the active zone - were developed and
tested under reactor conditions.
2. A tightly sealed connection between the stainless steel and the zirconium alloy was constructed.
3. A design was worked out for a heat-generating assembly operating in a stable manner in a boiling
coolant.
4. Sensors for monitoring the energy generation, to be used inside the zone, were developed, and on
the basis of these sensors, systems for the monitoring, control, and stabilization of the energy-
generation fields were set up.
5. A system and a set of algorithms for the centralized monitoring and operational estimation of the
state of the heat-generating assemblies was worked out by means of a computer.
6. Conditions were worked out for the effective removal of heat from the graphite stack to the coolant.
7. The operating regimes of high-powered nuclear reactors in a complex with a 500-MW turbine us-
ing saturated steam were tested.
The LAPS is the first of a series of APS with reactors of this type which are being constructed in
the USSR. At the present time, work is being completed on the installation of equipment on the first block
of the Kursk APS, and intensive preparations are in progress for startup of the reactor. Installation of the
reactor on the first block of the Chernobyl APS is in progress, and the construction of the Smolensk APS
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Fig. 6. Arrangement of equipment in the RBMKP-2000 (ground plan).
has begun. Each of these stations will include four blocks with RBMK reactors having an electrical power
of 1000 MW each. The decision has been taken to construct a second two-block complex for the LAPS and
a number of other APS with RBMK reactors.
Soviet channel-type uranium-graphite reactors of the RBMK type, with a unit power of 1000 MW, re-
present a step forward in the development of channel reactors. An analysis of the reactor characteristics
after the nominal thermal power of 3200 MW is reached showed that there are considerable reserves in the
design of the reactor. A number of parameters determining the limiting power of the reactor, such as the
temperature of the metal structure and the graphite stacking, actually proved to be somewhat lower than the
calculated values. Therefore, the question that naturally arose was whether the power could be increased
by making minimal changes in the design of individual assemblies. The idea of such changes was supported
by the fact that the main circulation pumps, of the TsVN-7 type, have the necessary reserve in the pres-
sure they develop. The designers of the RBMK reactor enthusiastically tackled the task of investigations
involving design, calculation, and experimentation to confirm the technical feasibility of this idea and to de-
terminetthe allowable limit of the additional power. The most important problem was to increase the criti-
cal power of the fuel channel, i.e., the power at which there occurs at the surface of the fuel elements a
heat-exchange crisis accompanied by an unacceptable rise in the temperature of the zirconium jacket.
This problem was successfully solved by introducing heat-exchange intensifiers into the regular heat-gen-
erating assembly. Tests were conducted on a number of variants of intensifier design; the optimal variant
was one which used lattice intensifiers with an axial twist in the coolant flow. Such lattices are set up, with
a pitch of 80 mm, only on the 3.5-m-long upper heat-generating assembly. Tests on a test stand indicated
that the critical power of an RBMK channel with heat-exchange intensifiers is about 1.5 times what it would
be without them. After less than one year, in July 1975, the technical design for the RBMK-1500 reactor
was brought out; this showed the technical feasibility of increasing the useful power of the RBMK reactor
to 1500 MW by intensifying the heat exchange in the fuel channels while keeping unchanged the structure of
the reactor as a whole. A number of questions will undoubtedly require further research (e.g., vibration-
wear tests of the heat-generating assemblies with intensifiers), but the most important questions were suc-
cessfully resolved and technical justification for the decisions was established. The technical design for
the RBMK-1500 reactor has been approved, and a decision has been taken to construct APS with reactors
of this type.
The intensive development of nuclear power and the trend toward increasing the unit power of reac-
tors confronts specialists with the problem of working out a design that will make it possible to construct
reactors from unified and standardized assemblies, i.e., without restructuring the machine-construction
industry base and complicating the installation of the reactor. The possibilities of uranium-graphite
channel reactors have made it possible to find ways of solving the problem. The first step in this direc-
tion is the design for the RBMKP-2000 section-block channel reactor, with an electrical power of 2000
MW. The distinguishing feature of the design of the RBMKP-2000 is that its shape is not the traditional
cylindrical one but that of a rectangular parallepiped consisting of separate sections (Figs. 5 and 6).
Through the use of sections of uniform type, it becomes possible to set up a reactor of almost arbitrarily
large power using identical arrangements, both for the reactor and for the structural elements of the build-
ing [5]. Each section includes the necessary equipment and control and monitoring instruments and con-
sists of separate transportable blocks. Special attention should be drawn to the fact that nuclear superheat-
ing of steam can be conveniently arranged in a section-block reactor [6]. The vaporizing and steam-
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superheating sections are uniform in design; the difference lies in the design of the heat-generating assem-
blies of the vaporizing and the steam-superheating channels and also in the fact that the vaporizing sections
include circulation pumps and separators. The sections may be regarded as independent zones of the reac-
tor, in which it is possible to make controlled changes in power within the required limits and there is a
certain degree of independence with regard to the cooling-water and steam loops. All of this creates fav-
orable conditions for localizing breakdowns and repairing some equipment without shutting down the reac-
tor.
The RBMKP-2000 reactor consists of eight vaporizing sections, four steam-superheating sections,
and two end-face sections. Each section has upper and lower blocks with separation of the pipes, lateral
blocks, supports, graphite stacking, and vaporizing or steam-superheating channels. The lateral sections
serve essentially as neutron reflectors and are equipped with special cooling channels. In working out the
design, maximum use was made of the experience obtained in the construction and operation of the RBMK
reactor. The vaporizing and steam- superheating channels of the RBMKP-2000 are practically identical
with the channels of the RBMK-1000, i.e., are of tubular construction, with the central portion made of a
zirconium alloy. This central portion is connected to the stainless-steel upper and lower portions with
similar steel-zirconium adapters. The required temperature conditions of the zirconium tubes of the
steam-superheating channels are ensured by slightly superheated steam passing through the annular gap
between the shell of the heat-generating assembly and the pressure pipe. The vaporizing and steam-super-
heating heat-generating assemblies are identical in construction, but the fuel elements of the steam-super-
heating channels are different from the fuel elements of the vaporizing channels: instead of a zirconium-
alloy jacket measuring 13.5 x 0.9 mm, they use a stainless-steel jacket with an external diameter of 10 mm
and a wall thickness of 0.3 mm. Unlike the RBMK reactor, in which the upper and lower supporting and
shielding metal structures are filled with a serpentinite charge with a low thermal conductivity, in the
RBMKP-2000 reactor the thermostating is achieved by filling these structures with water.
The coolant circulation is achieved as follows. From the deaerator the feed water is fed into the
downcomers of the separators, mixed with saturated water, and fed by the circulation pumps into the
vaporizing channels of the reactor. From the channels the steam-water mixture enters the separators.
The saturated steam is fed into steam-superheating channels, heated to 450?C, and fed at a pressure of 65
atm through steam ducts to two turbines having a power of 1000 MW each. Table 1 shows the main charac-
teristics of the RBMK-1000 and RBMK-1500 high-power uranium-graphite channel reactors with a boiling
coolant and of the RBMKP-2000 reactor with nuclear superheating of the steam.
The special features of channel reactors make it possible to modernize them steadily and continu-
ously, and therefore the technical and economic indicators of APS with such reactors will be improved.
This is clearly demonstrated by the design for the RBMKP-2000 reactor, which uses the progressive sec-
tion-block principle of reactor construction, a principle whose possibilities would be difficult to over-
estimate. The successful solution of a number of problems listed below will make possible further im-
provements in channel reactors:
1. The development of reactor-materials technology, including the production of high-temperature
zirconium alloys which will make it possible to improve the parameters of the vaporizing loop of
the reactor and will be usable in steam-superheating channels.
2. Further research on the intensification of heat exchange in channels with a boiling coolant, the de-
velopment of various designs for intensifiers, and the experimental verification of their operating
capacity.
3. The improvement of the design of channels and heat-generating assemblies with superheating of the
steam inside the reactor.
4. The development of means of effective heat removal from the graphite stacking in order to reduce
the temperature of the graphite and use nitrogen instead of helium for filling the stacking.
5. The improvement of safety measures at reactor installations as a way to increase the number of
circulation loops and reduce the diameter of the pipes, including the development of more effective
systems for emergency cooling of the active zone and the localization of coolant leaks.
6. The investigation of possibilities of regulating the reactor by using a liquid absorber.
The high reliability of uranium-graphite channel reactors, the relative simplicity of their construc-
tion, the possibilities of achieving high safety levels in the event of damage to the pipes of the cooling loop,
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and the practically unlimited possibilities of increasing power, the possibility of recharging fuel while the
reactor is in operation, the flexibility of the fuel cycle, the convenience of introducing nuclear superheating
of steam, and a number of other advantages make this Soviet type of reactor one of the most important in
the country's large-scale power industry and open favorable prospects for the further improvement and de-
velopment of these reactors.
1. A. M. Petros'yants, From Scientific Research Atomic Industry [in Russian], Atomizdat, Moscow
(1972).
2. Twenty Years of Atomic Power [in Russian], Atomizdat, Moscow (1974).
I. Ya. Emel'yanov et al., in: Experience in the Operation of Atomic Power Stations and Ways to the
Further Development of Atomic Power [in Russian], Vol. 2, Izd. FEI. Obninsk (1974), p. 166.
4.. A. P. Aleksandrov et al., ibid. Additional issue, p. 3.
5. N. A. Dollezhal' [3], p. 233.
6. N. A. Dollezhal' et al., Problems of Atomic Science and Technology. Series on Reactor Construction
[in Russian], No. 2 (9), TsNIlatominform, Moscow (1974).
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 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070002-9
PHYSICAL STARTUP OF THE RBMK-REACTOR* OF
THE SECOND UNIT OF THE V. I. LENIN NUCLEAR
POWER STATION, LENINGRAD
1. Ya. Emel'yanov, M. B. Egiazarov, UDC 621.039.519:621.039.524.2.034.44
V. I. Ryabov, A. D. Zhirnov,
V. P. Borshchev, B. A. Vorontsov,
A. N. Kuz'min, Yu. I. Lavrenov,
V. S. Romanenko, Yu. M. Serebrennikov,
and A. P. Sirotkin
In accordance with the program for the development of nuclear power generation in the Soviet Union,
in May to June 1975 the physical start-up was achieved at the reactor of the second unit of the V. I. Lenin
Nuclear Power Station, Leningrad (LNPS). The physical start-up program for the reactor of the second
unit of the LNPS was based on the results of the physical start-up of the first reactor [11 and provided for
a number of comparative experiments during charging of the reactor. Charging of the reactor with fuel as-
semblies FA and with auxiliary absorbers AA was carried out with dry multiple forced circulation loops
MFC and cooling of the rods of the control and safety system CSS. Although the charged reactor with dry
channels, intended for the insertion of fuel assemblies and auxiliary absorbers, does not have the greatest
reactivity, this charging principle allowed the multiple forced circulation loop to be prepared for a power
start-up, simultaneously with charging of the fuel assemblies.
For reliable control over the core and for ensuring safety during charging, a temporary control and
safety system was used together with the regular control and safety system. It effected control of the
neutron flux, the reactivity and emergency shutdown, and it comprised six emergency shutdown rods (scram
rods), four manual controls, and also the neutron source actuator with a control switch and a position indi-
cator.
The physical start-up program consisted of the following main stages:
1. Composition of the minimum critical charge without auxiliary absorbers and standard control and
safety rods (charge No. 1).
2. Completion of zone up to the maximum number of identical polycells, the so-called periodicity cells
(loading No. 2, Fig. 1).
3. Additional charging of the reactor up to 1437 fuel assemblies and 239 auxiliary absorbers. ~--
4. Shaping of the initial charge of the core, taking account of the operating experience from the reac-
tor of the first unit.
5. Estimation of the reserve of reactivity of the initial charge, and ensuring the required duration of
operation before the first fuel recharging.
6. Determination of the reactivity effects with dry multiple forced circulation loops and with cooling
of the control and safety rods.
* Water-cooled/water-mode rated channel-type reactor (high-powered).
Translated from Atomnaya Energiya, Vol. 40, No. 2, pp. 127-132, February, 1976. Original article
submitted November 13, 1975.
?1 976 Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. No parr of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, micro filming.
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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1 7
Fig. 1. Periodicity cell:
1, 2, 3) cells with con-
trol and safety rods;
auxiliary absorbers; and
fuel assemblies, respec-
tively.
7. Measurement of the energy-release fields in the cold unpoisoned reac-
tor.
8. Bringing of the reactor to the minimum level of power, controlled by
the standard control and safety system.
The necessity for experiments (in comparison with the reactor of the
first unit) originates by the difference in the number of .technological param-
eters affecting the physics of the reactor. In particular, to these parameters
may be referred the density of the graphite (1.67 g/cm3 in comparison with
1.73 g/cm3 in the first unit), the average charge with respect to 235 U in the -
fuel assemblies and the difference in the boron content in the auxiliary ab-
sorbers, etc. Comparative experiments during loading, even at the initial
stage of the physical startup, permitted those changes to be forceast which
must be carried out in the total reactor charge, in order to ensure the re-
quired reserve of reactivity and distribution of the energy-release field.
Preparation of the Reactor for Start-up
Before starting to charge the fuel assemblies and auxiliary absorbers, the following operations were
carried out:
The multiple forced circulation loops and the control and safety rods were flushed and pressurized.
Running-in of all main circulatory pumps (MCP) and the pumps of the control and safety rod loop.
During operation of all the main circulatory pumps, the multiple forced circulation loops and the
graphite brickwork of the reactor were heated up. to 150?C over two days, and after heating up the graphite
brickwork was cooled to room temperature.
The monitoring, system for the integrity of the technological channel (MITC) was put into operation.
The regular and temporary control and safety rods were put into operation.
The through-channel water-flooding system was prepared.
The loudspeaker connection between the central hall and the modular control panel (MCP) was made
ope rational.
The system for filling the multiple forced-circulation loop with water from the emergency feed pump
(EFP) tank was flushed and prepared for operation.
The drainage reservoirs were prepared for receiving water.
The general exchange and special ventilation systems were brought into operation.
The entire assembly of auxiliary absorbers was installed, with a ratio of inserts of boron steel and
stainless steel of 3 :1 in the central section, of length 500 cm and 1 : 2 at the end sections with a length of
up to 100 cm.
A complete set of fuel assemblies with openings below the interzone sensors and 100fuel assemblies
were installed.
Charging of the Reactor. Comparative Experiments
During charging of the reactor, its control and safety systems were implemented, just as in the reac-
tor of the first unit, with instruments of the temporary control and safety rods in the presence of a neutron
source in the zone.
The first critical charge without auxiliary absorbers and the rods of the regular control and safety
system in the absence of water in the multiple forced-circulation and the control and safety loops, contained
24 fuel assemblies (Fig. 2) and with the temporary control and safety rods withdrawn Keff was 1.00096 (23
fuel assemblies and Keff = 1.0050 for the reactor of the first unit). Further charging of the reactor was
carried out with respect to the periodicity cells (12 fuel assemblies, two auxiliary absorbers, and two control
and safety rods). After charging 77 periodicity cells (916 fuel assemblies and 154 auxiliary absorbers),
charge No. 2 was brought to the critical state. Further, the critical state was recorded for charges con-
taining 1437 fuel assemblies and 239 auxiliary absorbers, and 1452 fuel assemblies plus 239 auxiliary
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TABLE 1. Some Results of Comparison Experiments during Physical Start-Up of the
Reactors of the First and Second Units of the Leningrad Nuclear Power Station
Reactor condition
o p
No. of control and safety rods in-
FA
AA
3 0 $
serted in core
o
z
~
'
Keff
i
a
ao
c
N ~~ I
i w
d a
u y I H
OR
MR
AR
SRA
~
y
yw0 y
C C
ro d
'
S ~' 3 o
~
o
~ c
d O
a
v
n ? ~ I.
n o 0
0
6 3
1
23; 24 t
No.
-
No
No
-
-
-
-
1,0050;
-i,0
1.00096
2
916
154
n
?
8; -
56; 56
4; --
-
1.00000;
? +111
1,00064
3
1437
Yes
939
Yes
?
13; 8
89; 89
12; 12
20; 20
1 .00034;
-0,33
1,00016
4
1452
239
Yes
10; 1
89; 85 i
9; 12
21; 20
1.00000;
--0.5
1
1
1,0032
*OR) overcompensation rods; MR) manual control rods; AR) automatic control rods; SRA) shortened rod-absorbers.
tHere, and in future, the first and second figures are for the first and second units, respectively.
X
51 50 47 46 45 44 43 42 41 40 37 3635 34 33 32 31 32 27 26
0
0
0
0
0
i
ii
ii
i
11
11
11
1
11
0
1
01
1
0
0
0
b
0
!
Ole!
1
o
c
o '{
o
MR
o
OR
o
s.
o
o
a
ER%
""~
,, .'
ER
a
o
o
[,/.I
s
o
0
b
M
%MR
b
o
o:.
.
0T
-
0
0
d
ER
ER
a
o
o
o' '/,'
R
ol
0
M
o
f
o i
s
o
e
o
b
0
10,
0
0
0
0
0
0
0
absorbers. Similar charges were brought to the critical state
in the reactor of the first unit. For the charges containing
1437 fuel assemblies and 239 auxiliary absorbers, the effect
of reactivity on the filling with water of the multiple forced-
circulation loop was measured by means of a reactimeter; as
in the reactor of the first unit, this was found to be +1.913.*
The results of the comparative .experiments obtained during
the physical startup of the reactors of the first and second
units are shown in Table 1. The difference in the effective
multiplication factor Keff was determined in the following
way. An identical sequence for withdrawing the control and
safety rods was adopted for both reactors on reaching the cri-
tical state. By measuring the efficiency of the control and
safety rods, amounting to the difference in the compensation
position, the difference in Keff was determined.
It follows from Table 1 that all the charges investigated
for the reactor of the second unit have a lower reactivity. The
difference in Keff varies from 0.33 to 1.1%.
7~4
Fig. 2. No. 1 charge and diagram of
the disposition of the sensors and the
temporary control and safety rods: 1,
2) cells with charged fuel assemblies
and uncharged channels; 3) cells with
regular control and safety rods; 4)
cells with sensors: a) galvanometers
(G1, G2, and G3), b) reactimeters
(PIR-1 and PIR-2), c) counter-trigger
devices (SPU-1 and SPU-2), d) power
pen recorder (EPPV), e) scram rod
boosters (UA-9-1 and UA-9-2), f) scram-
rod velocity instruments (UZS-1 and
UZS-2); 5, 6) cells under temporary con-
trol and safety rods; S) neutron sources;
MR) manual control rods; ER) emer-
gency shutdown rods.
Effect of Various Parameters
The core structure of the reactor of the second unit, in
accordance with the physical start-up program, was specified
by the identical core of the first unit. The differences which
appeared during the physical start-up of the second reactor
necessitated calculations to be carried out in order to assess
the effect of deviations of the various parameters on the multi-
plication properties. The results of the calculations are shown
in Table 2.
Analysis of the deviations from the nominal values of the
mass characteristics of the fuel in the fuel assemblies, the
graphite purity, and the content of boron in the auxiliary ab-
sorbers, showed that all these factors can be eliminated from
those significantly affecting the multiplication properties. On
contrary, a change of graphite density affects not only the multiplication properties but also the diffu-
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TABLE 2. Effect of Deviations of Various Parameters on the Multiplication Properties
of the RBMK Reactor Core
Nominal value of
parameter
Devia. from nominal
value assumed in
calculation
Uranium enrichment
1.787%
+0.01%
+0.18
Fuel density
9.30 g/cm3
+0.1 g/cm3
+1.32
Graphite density
1.67 g/cm3
+0.1 g/cm3
-0.31(-7.1 for AMZ/ M2)t
Absorption cross section of graphite
4.2 mbar
+0.1 mbar
-0.13
Boron content of auxiliary absorbers
2.0%
+0.1% I
-0.02
*Values of AK00/K.are given for the reactor, with water in the multiple forced -circulation loop.
tM2 = L2 + T.
also their overflow to the control and safety rods and the auxiliary absorber rods.
Calculations by the QUAM-2 program showed that a reduction of the graphite density led to the follow-
ing losses of reactivity for the charges being compared (see Table 1):
Charge No. 1 - 0.96% (- 1.0%,,);
Charge No. 2 - 0.91%
Charge No. 3 - 0.33% (- 0.33%);
Charge No. 4 - 0.31%r (- 0.50%).
The experimental data are shown in the brackets. Thus, the calculations confirm that a reduction of reac-
tivity in the reactor of the second unit is mainly due to the reduction of the graphite density. However, ac-
cording to the calculations, this does not lead to a noticeable change of the depth of burnup in view of the
increased plutonium production.
Formation of the Initial Reactor Charge
The reduction of reactivity which appears in the reactor of the second unit is compensated mainly by
substituting 9 auxiliary absorbers by fuel assemblies. Moreover, the interchange of several peripheral
auxiliary absorbers was effected, which gave rise to certain difficulties in the case of rechargings during
operation of the reactor of the first unit. In contrast from the first unit, auxiliary absorbers were installed
on the periphery in the lattice of the control and safety rods which, in this region are disposed approximately
twice as sparsely as at the center of the core. It was decided not to load eight channels on the periphery of
the reactor as, according to calculations, the installation of fuel assemblies in them leads to an increase of
nonuniformity of the power release.
As a result of rearrangements and transfers, the initial charge for the reactor of the second unit was
defined: 1455 fuelassemblies, 230 auxiliary absorbers,and 8 uncharged channels. The critical state of the
initial charge (the multiple forced-circulation loop and the control and safety loop filled with water) was
achieved by the insertion in the reactor of 89 manual control rods, 12 automatic control rods, 21 shortened
rod-absorbers, and 10 overcompensation rods. 21 emergency shutdown rods and 26 overcompensation rods
were withdrawn. With this situation of the control and safety rods, Keff = 1.00077, the temperature of all
core elements was -20?C and the reactor power was -1.4 kW.
Experiments on the Initial Reactor Charge
One of the problems of the physical start-up is to determine the basic physical characteristics of the
reactor, necessary for its future operation. For this purpose, in the initial charging of the reactor the
effects of reactivity were measured with dry cooling loops of the control and safety rods, fuel assembly
channels (estimation of the "steam" effect of reactivity in the cold state), the multiple forced-circulation
loop, and with installation of the interzone monitoring sensors.
At the same time, the total efficiences of the inserted control and safety rods (estimation of the re-
serve of reactivity of the cold unpoisoned reactor) and of the withdrawn control and safety rods were deter-
mined.
All negative reactivity effects were measured with a reactimeter during the introduction of reactivity
into the critical reactor. The efficiency of the inserted control and safety rods was measured by their suc-
cessive withdrawal from the critical reactor. If the efficiency of a single rod exceeded 0.30, then the
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 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070002-9
TABLE 3. Experimental and Calculated Data for the Initial Reactor Charge
Effect of reactivity with:
dry channels with fuel assemblies
dry channels with auxiliary absorbers
dry multiple forced -circulation loop
dry control and safety loop
installation of 117 sensors for monitoring the radial neutron field
Total efficiency:
of inserted control and safety rods
of withdrawn control and safety rods
Nonuniformity factor:
of radial neutron field
of neutron height field
-0.42%
-0.43%
-1.60%
-1.1 W/o
-2.02%
-1.621o
Compensated by the insertion
of 13 controland safety rods
-0.006%
8.9%
7.3/0
1.9%
1.6%
2.04 ?
1.94*
1.37
2.45%
*Obtained on fuel assemblies in which measurements were carried out by fission chambers.
tFor all fuel assemblies of the reactor.
measurements were carried out by the overcompensation method. Dehydration of the control and safety
loop was carried out in the subcritical state.
The relative power release field in the initial reactor charge was measured with small-sized fission
chambers. At the same time, five independent measurement channels, in the corresponding way to the com-
muted channels, participated in the measurements. The measurements were made at eight points with re-
spect to height in 144 fuel assemblies, having at the center dry channels for the fission chambers. The
quality of the relative measurements, carried out twice at several points, has a mean-square error of 1.6%.
The absolute thermal neutron flux'1T was determined by the activation of gold foils in and without cadmium.
The absolute power of the fuel assemblies, in which the absolute thermal neutron flux was measured, was
determined from the relation
iDT VT0/T 1/n/4 a10fN5kr,k1
WT 3.1.1010kT
where T is the neutron temperature at the point of location of the indicator; T0 = 293?K; 7 of is the fission
cross section of 235U when T = To; N5 is the number of 235U nuclei in the fuel assemblies; kr is a factor
which takes into account fission by resonance neutrons; kl is the deviation of the neutron flux measured by
the fission chambers and averaged over the height, from the neutron flux at the site of irradiation of the
indicator; and kT is the ratio of the neutron flux at the point of measurement to the average neutron flux in
the fuel.
The reactor power was determined from the formula
Qe M
WT i-1
Qi,
Wp e It
QT I Qp i-1
where QT is the relative power of a fuel assembly, measured by the fission chamber, the absolute power
n
of which was determined by gold activation; Qi is the summed relative power of the fuel assemblies, mea-
i=1
n in
sured by the fission charrbers; E Qip and QP are the total relative powers, calculated for the fuel as-
i=[ i=1
semblies n in which the fission chamber measurements were made, and all fuel assemblies m respectively.
All critical charges, and also the measured power release fields, were computed by the BOKR-COB
and QUAM-2 programs, which describe channel-wise the structure of the core. Moreover, the experimen-
tal efficiency of the control and safety rods was computed by the BOKR-COB program.
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The BOKR-COB program is a development of a program [2, 3] based on the solution of the diffusion
equations of a reactor by a finite-difference method in x-y geometry (for the cross section of the reactor).
In the program, the two-group diffusion equations of a reactor consisting of heterogeneous square cells are
solved. The nodes of the reference mesh coincide with the centers of the channels. It was shown by the
calculations of the experiments carried out on critical assemblies, and also on the reactor of the first unit
of the Leningrad Nuclear Power Station, that such an arrangement of the reference nodes is more prefer-
able than in the angles of elementary cells. The nonuniform poisoning of the fuel by xenon, as a function of
the designed distribution of the power-release field and the fuel burnup, are taken into account in the pro-
gram. The presence in the core of the control rods and other breeder channels is taken into account by
assigning the appropriate homogenized properties of the cells in which these absorbers and channels are
located. Partially inserted rods are replaced by completely inserted rods of equivalent efficiency. For
the operational calculations, a modification of the program is used -'the BOKR-COBZ program in which,
in order to take account of the partially inserted control and safety rods, experimental measurements are
used of the height neutron field by the sensors of the physical control system.
The QUAM-2 program achieves a new method of calculating heterogeneous reactors [4]. The reactor
is represented in the form of a finite lattice of channels (in x-y geometry) in an infinite moderator. The
transport of neutrons in the moderator is described by two-group diffusion equations of the Galanin-Fein-
berg type [5, 6], which are transformed to the so-called quasialbedo form similar to the finite difference
form, and which are solved by an iterration method. As a result, computer time in solving the equations
is shortened by a factor of 15-20 in comparison with the traditional heterogeneous method and amounts to
1.5 min for the calculation of a single reactor state. The QUAM-2 program enables Keff to be calculated
and also the power distribution over the reactor channels with a specified position of the completely or
partially inserted control and safety rods. The possibility is provided for taking into account the steady-
state poisoning by xenon in the uranium burnup, individually for each channel. In calculating Keff, a cor-
rection is made for the axial leakage of neutrons and the nonuniformity of properties over the height of the
reactor, and also a correction which takes account of the processes caused by moderated neutrons and
due to the presence of nonbreeding channels.
By means of the QUAM-2 program and a system of supporting programs, calculations of about 70
critical states (cold and hot poisonings and with uranium burnup) have been carried out for the reactors
of the first and second units of the Leningrad Nuclear Power Station. The mean-square error in determin-
ing Keff amounts to 0.5%and the maximum deviation does not exceed 1%.
Comparison of the calculations by the BOKR-COB and BOKR-COBZ programs with the experimental
data, shows that the calculations predict satisfactorily the criticality of the various states of the reactor.
For the system of neutron-physical constants assumed, the discrepancy in Keff does not exceed 0.9%. For
a complete charge, it does not exceed 0.5% and, taking into account the height field of the neutrons (the
BOKR-COBZ program), it amounts in all to 0.2%. It is shown that the height distribution of the neutrons
has a marked effect on the calculated value of Keff. In the calculations with sinusoidal and measured neu-
tron distributions, the difference in Keff amounted to 0.3%. Therefore, for a more accurate calculation of
Keff by the BOKR-COBZ program, it is necessary to take account of the actual height distribution of the
neutrons. The efficiencies of different groups of control and safety rods, calculated for different charges,
mainly coincide well with the experimental data. The results of the experiments and calculations of cer-
tain effects of reactivity and neutron distributions over the core, carried out for the initial reactor charge,
are shown in Table 3.
Comparison of the experimental and calculated power-release fields along the radius of the reactor,
obtained by the BOKR-COBZ and QUAM-2 programs, showed agreement at the location of the field maxi-
mum; the mean-square error in determining the power of the fuel assemblies by both programs is identical
and amounts to 9.7%.
After completion of the experiments on the initial charge, the power start-up of the secondunitofthe
Leningrad Nuclear Power Stationwas effected in July-August 1975. The power of the unit was built up
gradually in accordance with the readiness of the turbogenerators. Initially, at a total power of 500 MW,
the third turbogenerator was cut-in and then the fourth turbogenerator was brought into operation. On
September 30, 1975 the State Commission authorized the handover of the second unit.of the V. I. Lenin Nu-
clear Power Station, Leningrad to commercial operation. On October 10, 1975 in accordance with the
start-up program, the electric power output of the second unit amounted to 750 MW.
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During the power start-up, the basic decisions taken according to the results of the physical start-up
were checked and confirmed. In particular, according to the results of measurements of the power-release
fields, the rate of decrease of reactivity as a result of poisoning and burnup of the uranium, the validity of
the specification for the initial charge, and the creation of the necessary reserve of reactivity were con-
firmed.
During the physical start-up of the reactor of the second unit of the Leningrad Nuclear Power Station,
the initial reactor charge was specified, experiments and calculations were carried out which permitted a
comparison to be made with the results obtained during the start-up of the reactor of the first unit:
1. The initial reactor charge consists of 1455 fuel assemblies and 230 auxiliary absorbers (eight
channels remain uncharged).
2. There are certain differences between the reactors of the first and second units, in that similar
charges in the reactor of the second unit have a lower reactivity. The decrease of reactivity of the total
reactor charge (charge No. 4) amounted to 0.5%. This difference is explained mainly by the lower density
of the graphite in the reactor of the second unit.
3. Dehydration of the 1455 channels with fuel assemblies and the 230 channels with auxiliary ab-
sorbers reduces the reactivity by 2%.
4. The effect of reactivity during dehydration of the control and safety loop is positive and is com-
pensated by the insertion of 13 shutdown -rods.
5. The reactor has the greatest reactivity when the circulation loop is filled with water and the con-
trol and safety loop is dry.
6. Dehydration of the channels with 1455 fuel assemblies leads to a reduction of the reactivity by
0.42%, which exceeds by 0.3%the similar effect in the reactor of the first unit and confirms the more nega-
tive steam effect of reactivity on the initial stage of operation of the reactor.
7. The total efficiency of the inserted control and safety rods amounts to 8.9%.
8. In order to bring the cold, unpoisoned reactor with water in the multiple forced-circulation and the
control and safety loops to the critical state, 47 shutdown rods must be withdrawn, the total efficiency of
which amounts to 1.9%
9. A comparison of.the experimental and calculated power-release fields along the radius of the reac-
tor, obtained by the BOKR-COBZ and QUAM-2 programs, showed that the mean-square deviation between
the measured and the calculated powers of the fuel assemblies by both programs is identical and amounts
to 9.7%.
The physical startup of the reactors of the first and second units of the Leningrad Nuclear Power Sta-
tion permitted considerable experimental data to be accumulated concerning the startup of RBMK-type
reactors. By taking into account the possibility of variation of certain technological parameters of the
core materials, it will be advantageous during startup of the next units to carry out comparative experi-
ments in order to correct the initial charge and to refine the operating characteristics of the reactor.
1. I. Ya. Emel'yanov et al., in: Experience in the Operation of Nuclear Power Stations and Routes for
the Future Development of Nuclear Power Generation [in Russian], Vol. 2, Izd. FEI, Obninsk,
(1974), p. 166.
2. I. S. Akimov, M. E. Minashin, and V. N. Sharapov, At. Energ., 36, No. 5, 427 (1974).
3. I. Ya. Emel'yanov et al., in: Experience in the Operation of Nuclear Power Stations and Routes for
the Future Development of Nuclear Power Generation [in Russian], Vol. 1, Izd. FEI, Obninsk, (1974),
p. 81.
4. S. S. Gorodkov, Preprint IAE-2251 [in Russian], Moscow (1973).
5. S. M. Feinberg, in: Reactor Construction and the Theory of Reactors [in Russian], Izd. Akad. Nauk
SSSR, Moscow (1955), p. 152.
6. A. D. Galanin, in: Reactor Construction and the Theory of Reactors [in Russian], Izd. Akad. Nauk
SSSR, Moscow (1955), p. 191.
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HIGH-TEMPERATURE REACTORS AS A FACTOR
IN SCIENTIFIC PROGRESS IN
POWER GENERATION*
N. A. Dolle.zhai' and Yu. I. Koryakin UDC 621.039.576:621.039.524.2
The successes in power-generating reactor construction have led to the creation of an extremely
large branch of power production - nuclear-power generation, which has a complex but essential assembly
of scientific-technical means for ensuring its development.
At the present time, there remains a very limited number of nuclear reactor types in the whole world
for future utilization and modernization. Reactors with water cooling in three structural modifications
occupy the dominant position (Table 1); the vessel type (pressurized water and boiling water) and the chan-
nel types which, in particular, have undergone long development in the Soviet Union. Over the last 10-15
years, there has taken place in nuclear-power generation, if it can be so expressed, a natural selection of
reactor types according to the extent of their technicoeconomic capabilities and conformity to problems of
large-scale electric power production. The limited number of reactor types and modifications as a basis
of standardization and unification is an essential condition for the observed intensive growth of nuclear-
power generating capacities. However, over recent years and especially in view of the energy generation
crisis, a marked interest has been shown toward high-temperature nuclear reactors.
With the well-known convention, the factors stimulating the development of high-temperature nuclear
power generation can be divided into two groups.
1. Factors arising from the internal evolution of the development of nuclear electric-power generation
and affecting its economy.
2. External factors, either arising from the evolution of the interrelation between nuclear-power
generation and the habitable environment, or dictating the ever-increasing necessity for extracting organic
fuel from energy-containing thermal processes and the necessity for conversion of the energy carriers.
Among the first group, the efficiency of utilization of nuclear fuel both in thermal neutron reactors
and also because of the problem of fast breeder reactors is of the most decisive importance; the conver-
sion factor for the heat from the core into useful power, the specific capital investment in nuclear power
stations, and also problems of safety. First and foremost, it should be emphasized that high-temperature
nuclear power generation at the present-day level of scientific knowledge is related with reactors in which
the coolant is an inert gas - helium. However, in conformity with the mainly nonelectric power industrial
problems, other methods also are possible for removing heat from the core, based on other ideas in con-
trast from convective heat exchange, inherent with a gaseous coolant.
The distinguishing features of a high-temperature reactor with helium coolant are not only a higher
temperature level at the outlet from the core than in all the well-known types of reactors, but also a con-
siderably better neutron balance in the core and a smaller requirement for structural materials, which en-
sures a greater efficiency of utilization of nuclear fuel per unit of electric power sent out (Table 2).
These properties of a high-temperature reactor are due to an extraordinarily favorable combination
of materials or media in the core - ceramic fuel, graphite as the moderator, and helium as the coolant
(unactivated gas, having an almost zero neutron absorption cross section). Owing to this, high-efficiency
*Paper in abridged form was reported at the Scientific Session of the Division of Physicotechnical Problems
of Power Generation, Academy of Sciences of the USSR, devoted to high-temperature reactors June 6, 1975.
Translated from Atomnaya Energiya, Vol. 40, No. pp. 133-144, Februatv, 1976. Original article
submitted September 8, 1975.
01976 Plenum Publishing Corporation. 227 West 17th Street, New York, N. Y. 10011. No part of this publication nmay be reproduced,
stored in a retrieval. system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise. without written permission of the publisher. A copy of this article is available from the publisher for $15. 00.
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TABLE 1. Trend of Development of Reactor Construction in Certain Countries
Great
Britain
Type of reactor, accepted for nuclear
power stations under construction
Pressure vessel: with water under
pressure, dual circuit: with boiling
water, single circuit
Channel, heavy -water: with heavy -water
under pressure, dual-circuit, CANDU -
PHWR. Boiling, with heavy-water
moderator and light water as coolant,
single-circuit, CANDU -B LW
Pressure vessel: with water under
pressure, dual circuit: with boiling
water, single circuit
Pressure vessel: with water under
pressure, dual circuits with boiling
water, single circuit
High-temperature reactor with helium coolant (HTGR):
for the production of electric power (dual circuit);
for the production of high-potential heat and electric
power, single and dual circuit; for the production of
electric power and central heating, single circuit,
with helium turbine. Fast neutron breeder reactors:
with sodium coolant, triple circuit; with helium
coolant, dual circuit
Pressure vessel, with water under pressure, dual-circuit.
Fast neutron breeder reactor with sodium coolant
Channel, with heavy-water moderator and light boiling
water as coolant (SGHWR), single circuit. High-tem-
perature (HTGR) and with very high temperature of
helium at the outlet from the reactor (VHTR)
Channel, with heavy -water moderator and light boiling
water as coolant (CIRENE), single circuit
Type of reactor, having
been developed to the
stage of commercial
experiment
Reactor with organic
coolant and moderator.
Pressure vessel reactor
with nuclear super
heating.t S odium -
graphite reactor*
Channel, with organic
coolant (development
discontinued in 1973)
Gas-graphite, with CO2
as coolant. Heavy-water
with C02 as coolant
Gas-graphite, with CO2
as coolant.
Federal
German
Republic
Pressure vessel: with water under
pressure, dual circuits with boiling
water, single circuit
?OMRE, PI QUA.
tBONUS, PATHFINDER.
IHALLAM.
High-temperature reactor with thorium fiel cycle and
helium coolant
Gas-graphite with CO2
as coolant. Reactor with
organic coolant and
moderator (ORGEL, PRO,
ROVI)
utilization of the neutrons in the core and the significantly higher (than in the well-known reactors with
water cooling) thermal cycle efficiency (41-43%against 31-33%) is ensured.
The relatively low pressure of the gas coolant (40-60 atm) and its other properties as a medium,
makes it the most advantageous and feasible to use a reinforced concrete vessel, in which all the main
plant of the primary loop are grouped (integral grouping). The feasibility of making the vessel on the floor
of the nuclear power station eliminates serious weight and size problems of transportation and gets rid of
the necessity for designing and developing mechanical engineering facilites and workshops for the manu-
facture of such vessels. The construction of reinforced-concrete vessels is technically feasible at a unit
electric capacity of 1.5 to 2.0 million kW.
These, and other circumstances, finally will undoubtedly favor a reduction of the specific capital in-
vestment in nuclear-power stations with high-temperature gas reactors and will imply also a reduction of
the cost of useful electric power sent out. Particularly attractive in this respect is the direct gas-turbine
cycle, which can be achieved even with helium temperatures of 750-850?C at the outlet from the reactor.
Attention should be paid to the extremely important question, which is of.an independent nature and
which is associated with the future structure of the nuclear power-generation system, i.e., with the opti-
mum ratio between nuclear power stations with thermal and fast reactors. The optimum dynamic structure
of a nuclear power-generation system containing fast reactors is determined to a considerable degree by
their efficiency margin, expressing the difference between the inherent and contractural costs in the
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TABLE 2. Requirement on Natural Uranium for the Production of
Electric Power, with Different Types of Thermal Reactors using Low-.
enrichment Uranium
Consumption of natural
uranium (for different
modifications), g/MW ? d
Pressure vessel with water coolant
25-34
Channel with graphite moderator and water coolant
20-25
High-temperature with graphite moderator and helium coolant
14-18
system. In its turn, this margin depends on the ratios both of the fuel component and the magnitude.of the
specific capital investment of nuclear-power stations with fast and thermal reactors.
The magnitude of the specific capital investment in nuclear-power stations with thermal reactors
has been determined adequately. At the same time, for fast reactors this quantity has a considerable in-
determinacy, mainly in consequence of the experimental-industrial destination of the isolated fast breeder
reactors being constructed. Because of this, in a number of papers on the system analysis of nuclear-power
generation, the effect was investigated of the possible excess of the specific capital investments of nuclear-
power stations with fast reactors, in comparison with thermal reactors, on the efficiency of fast reactors.
There is still no significant experience of the construction of large-scale nuclear-power stations with
fast sodium reactors. However, the experience in planning and building the first nuclear-power stations
with these reactors is extremely alarming in relation to capital investments. For example, the specific
cost of a fast sodium reactor with a capacity of 350 MW (el.) at Clinch River (USA), marked for completion
of construction in 1982, has proved to be very high. There are grounds for fearing that the excess of the
specific cost of reactors with liquid-metal coolant over the specific cost of thermal reactors will be ex-
treme, and this will bring fast reactors beyond the limit of permissible economy (according to estimates,
it is probably no higher than 50-60%0) [1].
The main reason for the relatively high cost of nuclear-power stations with breeder reactors with
sodium coolant lies in the complex triple-circuit heat-removal system and conversion of the heat into elec-
tric power (sodium-sodium-water), in the listed multiplicity of plant, and in the complexity and impor-
tance of the systems for filling, heatingup, and monitoring of the liquid-metal loops. Therefore, as alterna-
tives fast reactors with a gaseous coolant are being considered in a number of countries. Design and
planned operations in both the USSR and abroad show that these reactors have approximately equal or
somewhat better indexes with respect to breeding and nuclear fuel utilization, the fuel cycle and heat con-
version efficiency than reactors with sodium coolant. At the same time, the absence of negative qualities
which are characteristic of a liquid-metal coolant ensure a much simpler and, consequently, a cheaper
heat conversion scheme. Fast reactors with a gas coolant, therefore, give the basis for expectation for
obtaining an equal or somewhat larger value of the specific capital investments than for thermal reactors.
The construction of nuclear-power stations with fast gas-cooled reactors maybe considered as. an economically
promising problem and a desirable alternative for solving the general problem of fast reactors.
From the scientific and engineering aspect of the problem of building fast and thermal gas-cooled
reactors, there are a number of general features, such as material problems, gas-circuit plant and the
construction of its elements and units, gas economy schemes and auxiliary services (safety, decontamina-
tion, dosimetry, etc.), grouping, andother general plant problems. Therefore, work and results of investi-
gations on thermal and fast reactors with helium coolant are of a mutually beneficial nature.
As concerns problems of safety, these problems as applied to thermal gas-cooled reactors, in view
of the qualities which are inherent to these reactors, are solved no more intricately than for other types
of reactors:
1. Integral grouping of the primary circuit plant in a reinforced-concrete vessel and a significantly
lower coolant pressure (40-60 atm) than in pressurized water-cooled reactors almost eliminates
rupture of the primary circuit.
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2. Even in the low-probability case, rupture of the primary circuit is not accompanied by a heavy for-
mation of steam from superheated water, which is inevitable for pressurized water reactors.
3. Problems of corrosion of the primary circuit and the deposition in it of radioactive deposits are
solved more easily.
4. In view of the very small contribution of the gas coolant to the reactivity of the reactor and the high
storage capability of graphite, the loss of coolant is not linked with an abrupt and significant jump
in reactivity of the reactor.
5. Melting of the core or its elements is eliminated, as it consists entirely of ceramic materials with
a very high melting or volatilization point.
These, as an example, are the basic factors ensuring from the intrinsic evolution of the development
of electric power reactor construction, stimulating the development of work on the construction of high-
temperature nuclear reactors.
Let us consider the outside factors. Certain of them were shaped only recently, and the cause and
effect of their existence is still far from being explained and understood. Therefore, at the present level
of our knowledge, only the principal trends leading to their onset can be noted.
Among this group of factors, the question of the connection between a power installation and the sur-
rounding environment recently has occupied an important place. This connection appears more strongly in
terms of the efficiency of conversion of the primary heat released in the installation, into useful power.
An increase of the conversion efficiency, all conditions being equal, is related to factors which raise the
efficiency strictly of nuclear-power-generating plants and improve the efficiency of the nuclear fuel utilization.
Finally, since recently, attention to problems of ecology associated with nuclear-power production, in par-
ticular with the discharge of heat into the habitable environment, has increased. Therefore, the question
of the efficiency of a power-generating plant is now becoming so much broader. It concerns not only the
economics of power-generating plants but also problems of preservation of the essential qualities of our
habitable environment. High-temperature reactors possess a significantly higher efficiency (4.1-43%) and
conform better to this problem. If we add, that by a further increase of the gas temperature at the outlet
from the reactor up to 1000-1100?C, which gas-cooled reactors will allow, there is a possibility in princi-
ple of building a competitive modification of a nuclear-power station with a triple thermodynamic heat con-
version cycle and with an efficiency of 55-60%. In one of these proposals [21, potassium, diphenyl,and
steam of different parameters are used as the working media. Table 3 shows the important possibilities
of electric-power generation which could be achieved by the use of high-temperature reactors.
In connection with what has been said above, the question may arise: If it is necessary to develop high-
temperature electric power-generating reactors, what should be the attitude toward water-cooled reactors
which, as is well-known, at the present time occupy the predominant position in nuclear-power generation?
This question can be answered thus. Modern nuclear-power generation as a branch of electric-power prod-
uction has completed successfully the circle of its possibilities lying, if it can be so expressed, as yet only
on the surface of the reactor power-generation phenomenon. With a certain paradox it can be considered
even that a nuclear reactor is, in principle, a new heat-generating agent, operating on a steam turbine
cycle, the parameters of which will lag very much behind the parameters achieved in thermal power genera-
tion until right up to the appearance of the power-generating reactor. In the thermal-power generation
sense, the parameters of this cycle to a certain extent are even a step backwards. Nevertheless, electric-
power-generation problems, standing before nuclear-power generation, are so vast and the demand on them
is so great that reactors with water cooling, obviously, for quite a long time still will successfully satisfy
the problems of the electric-power generating industry.
However, the evolution and logic of the development of power-generating reactor construction, and
also the possibilities of nuclear-power generation by this time require a gradual transition to a new, higher
level with more improved specific branches and, chiefly, with systemic indexes. High-temperature reac-
tors reveal this possibility. Moreover, in recent years important power generation problems have started
to be shaped, the solution of which lies beyond the limits of possibility of water-cooled reactors. The sole
reason is the low temperature level obtained by means of these reactors.
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TABLE 3. Possible Characteristics of Nu- The fact that, as a rule, they are found to be outside
clear-Power Stations with a Triple Produc- the sphere of electric power production is general for
tion Cycle them. The necessity for their solution is connected mainly
Limiu of cycle, C/bar
Eic en-
cycle
upper 1 lower cy, ?
Potassium
890/3,0
477/0,027
29,1
Diphenvyi
455/20,9
287/2,0
16,9
Water (functioning
270/55
33/0,051
in thecondensa-
(Intermediate superheat-
33,6
_tion cycle)
ing 270/8.9)
with the limitation of organic fuel reserves and the every-
where increasing deficiency of its most important forms
- petroleum and gas. Related with these problems are
the extraction by nuclear fuel from energy-rich high-tem-
perature processes, of valuable forms of organic fuel (for
example, in ferrous and nonferrous metallurgy, and
chemistry) and the conversion of energy carriers, effected
at high temperatures (gasification of low-grade coals,
production of hydrogen, etc.).
The first group of problems concerns branches of the national economy, which over a long time have
become more complicated. There is a certain advantage in this, because the special features of power re-
quirements in these fields can be specified with greater certainty, and the demands on the nuclear reactor
and its parameters can be defined. But there is also a difficulty there, as the power-technological ex-
perience, having been involved in the field, sometimes leads to the onset of branch conservatism which
mainly is manifested in a tendency to consider the nuclear reactor as a source of heat which adequately
replaces the heat produced by burning organic fuel. However, the specific characteristics of a nuclear
reactor for producing the necessary economic effect, in many cases requires a reorganization of the tech-
nology of the process, a departure from the power-technological concepts and, as a consequence, changes
of the customary constructive solutions. A nuclear reactor in these industrial environments is represented
as a factor which affects the trend of technical progress.
The second group of problems is of a systemic multified nature, as it concerns the preparation of a
product for multipurpose utilization. This is a new group of all-round complicated problems, in which the
question of the construction of a nuclear reactor occurs in a number of other, no less complicated, prob-
lems. Here, there is still the greatest indeterminacy of all.
High-temperature reactors might be used in ferrous metallurgy, where there is an extremely large
demand for oganic fuel. For example, in the Soviet Union up to 20% of the volume of organic fuel required
for the entire industry of the country is expended on ferrous metallurgy requirements. One further reason
for the marked interest in the use of the heat from nuclear reactors in this field is the poisoning of the
environment during the carrying out of classical metallurgical processes.
At the present time, ferrous metallurgy is based on the use of coking, with the production of the
metal, for example by the "blast furnace-converter-rolling mill" system. Together with the future
modernization of blast furnace production, it is proposed to introduce cokeless metallurgy processes,
with the final product being produced by the process "direct reduction of iron ore-electromelted steel
product - rolled product." Cokeless metallurgy processes at the present time are at the initial stage of
utilization. Obviously, cokeless metallurgy is very promising and it can be economically competitive
with the blast furnace process for the production of metals. Among the several cokeless metallurgy pro-
cesses are the well-known processes of solid-phase reduction of nodulized iron ore concentrate by con-
verted natural gas, with the production of pig iron, which is carried out at temperatures of 950-1250?C.
Investigations carried out in the Soviet Union and abroad have shown that in this case, the use of high-
temperature reactors may prove to be promising. If the natural gas which is expended on heating the re-
ducing gases in this section of the metallurgical cycle up to 950-1250?C is replaced by the heat from a nu-
clear reactor, then up to 50-55%of the natural gas will be saved. As Soviet and foreign investigations have
shown, an economic effect can be expected to be produced. As contractual costs of organic fuel have a
tendency toward increasing, this effect in presepective can be significant.
The economical promise and the economical need for the introduction of high-temperature nuclear
reactors in ferrous metal production processes, however, is accompanied by technical difficulties. These
are due, first of all, to the high level of temperatures required and, secondly, to the considerable differ-
ence in pressure between the nuclear-reactor coolant (40-50 atm) and the reducing gases in the direct re-
duction plant units - the shaft furnaces (6-8 atm). Therefore, in order to reliably prevent radiation con-
tamination of the metallurgical product (pig iron), it will be necessary to ensure directivity of the pressure
gradient from the nuclear-reactor coolant to the reducing gas. By using a high-temperature reactor with
helium coolant, this can be achieved merely by the construction of a shielded intermediate circuit with a
higher pressure than in the reactor circuit.
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CaBr2+ 2H20 -. Ca(OH)2+ 2HBr
2HBr+Hg- HgBrz+Hz
HgBr2 -Ca(OH)2 -. BaBr2 + HgO + Hz0
Hg0 -. Hg 1 2 Oz
6FeC12+8H20-.2Fe304 + 12HC1+2H2
2Fe304 3012+ 12HC1-6H20 -;- 6Fe3C13= 02
6FeC13-. 6FeC12+3C13
KzO2+H20-.2KOH+ 2 02
2KOH+2K-.2K20+H2
2K20 -. K202 +2K
750
200
200
600
725
1000
In connection with this, the complicated problem arises concerning the steam pipes and heat-ex-
changers for the transport and transfer of heat in the range 1000-1300?C. There are still no acceptable
solutions or suggestions for this. In general, the complex nature of the nuclear metallurgical problem
should be emphasized. It will be necessary not only to find acceptable solutions to the amalgamation of a
high-temperature reactor with the conventional metallurgical production plants, but also to find technical
solutions for both the high-temperature reactor and the metallurgical production plant units.
Another promising trend for the utilization of the energy of high-temperature nuclear reactors is
their use in the chemical industry for the production of ammonia and methanol. These products are the
basic raw materials for the production of nitrogenous fertilizers, the demands for which will increase
strongly in the near future. Modern industrial technology for the production of ammonia and methanol is
based on the large-scale requirement for natural gas, used as a technological raw material and energy-
producing fuel. In this case, the fraction of natural gas consumed as an energy-producing fuel amounts to
-45%of the total requirement; large volumes of the combustion products of natural gas are discharged into
the surrounding medium; in particular, not more than 15-20%of the heat released in the combustion of
natural gas is used in the technological process and it is used at a temperature of 1000?C and higher; the
utilization of the heat of the high-temperature, low-pressure waste gases (1 atm) presents a serious tech-
nical problem and a large quantity of low-temperature heat at a temperature of 300?C and below is dis-
charged into the surrounding medium.
Thus, the effective increase of production of ammonia and methanol in the volumes necessary to
satisfy the increasing demands of the national economy in these products and their reprocessed products
encounters considerable difficulties, which are aggravated by the fact that the limitations on the use of natu-
ral gas as an energy-producing fuel are being intensified, and the demands arebeing increased for reducing
the discharge of toxic substances and heat into the surrounding medium.
The high-temperature heat from nuclear reactors can be used mainly in the heat-consuming section
of the overall technological cycle for the production of ammonia - during the carrying.out of the endother-
mic reaction of the steam-water or carbon dioxide conversion of methane. By the existing technology, these
reactions take place at a temperature of 800-900?C in the presence of a catalyst. At the same time, the
construction of nuclear reactors, experience in planning, and also prospective scientific ideas and technical
development, will allow confidence in the capability of building high-temperature reactor systems which, in
the level of working temperatures and also in technical solutions, will satisfy the conditions for their effi-
cient utilization as sources of power for high-temperature technological processes of the chemical indus-
try. As a result of this, the requirement for the chemical production of natural gas will be cut down by up
to 45%; the discharge into the surrounding medium of the combustion products of natural gas will be elimi-
nated; the thermal efficiency of the entire technological cycle will be increased; and the production cost of
ammonia and methane can be reduced markedly.
To the group of more general power-generation problems, which can be solved by means of high-tem-
perature nuclear reactors, are related the gasification of lignite and coal and other forms of low-grade
organic fuel. Work in this direction has been progressing quite broadly during a number of years in the
Federal German Republic, the USA, and Great Britain. Additional impetus toward the development of the
task was given in view of the energy crisis in the western countries in 1973-1974 and by the proposed
future increase of prices of petroleum and natural gas. Several processes are being considered for the
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es,
n, nuclear buildings of the USSR. Chernobyl'sk Nuclear-Power Station. Gen-
eral view of construction. Photo by V. Bratchikova.
gasification -of coal and lignite, using the heat from nuclear reactors. However, preference is being given
to two methods: hydrogenization and steam-water gasification [3, 4]:' In comparison with conventional gasi-
fication processes, which use as heat the energy from the combustion of organic fuel, it is expected that the
use of nuclear heat will give the following advantages:
1. The gas generated will be cheaper.
2. Coal resources will be utilized more efficiently, which is equivalent to a relative increase of these
3. Costly gas can be produced from deposits at a low cost of extraction, which is particularly impor-
tant,when using cheap lignites.
4. As'nuclear`fuelis a unique form of fuel. which is classified according to the relative cost of extrac-
tion, the increase of cost of raw material (coal) will have less effect on the cost of the gas pro- ,
duced by comparison with conventional gasification [3].
'5. Contamination of the surrounding medium will be reduced significantly, e.g., the discharge of car-
bon dioxide by 60% and sulfur by 40%.
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Fig. 2. In nuclear buildings of the USSR. Chernobyl 'sk Nuclear-Power Station. Con-
struction of the reactor hall. Photo by V. Bratchikova.
Recently, attention has been drawn to yet another problem of the conversion of energy-carriers: to
obtain hydrogen, which is being designated often as the fuel of the future, by means of the heat from nuclear
reactors. It is well-known that the heat from a nuclear reactor can be utilized for the production of hydro-
gen by means of the electric power produced in a nuclear-power station, which then is used for the conven-
tional electrolysis of water. However, in this scheme the additional stage of conversion of the energy-car-
rier leads to a low efficiency of the whole conversion cycle. Therefore, the thermochemical decomposition
of water is considered energywise to be more efficient, in which the heat from a nuclear reactor is used
directly for the production of hydrogen. The power generation efficiency of this conversion amounts to 50%.
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TABLE 5. Energy-Generating and Energy-Technological Courses for the
Application of High-Temperature Reactors
Course of application
necessary temp.
level, ?C
pressure level,
arm
Improvement of nuclear-fuel ultilization, increase
of efficiency of nuclear-power station up io 41-43%:
steam -turbine cycle
550-560
240
gas-turbine cycle up to 60-61%n
850-900
40-60
triple cycle
890-900
Up to 55
Production of ferrous metals
850-1250
5-6
Production of ammonia and methanol
750-900
Up to 30
Gasification of coal:
hydrogenation
700-800
25-50
steam -water
800-950
25-50
Production of hydrogen by thermochemical decomposi-
790-1000
Up to 10
tion of water
Transport of latent heat in a cycle with reverse
chemical reactions
Several cycles have been proposed [5] for the thermochemical decomposition of water. Common for
all is the maintenance of the quantitative balance of the intermediate chemical elements participating in
the reaction cycle (Table 4). The technological schemes for these cycles are quite complex, and the chemi-
cal compounds participating in the reaction and the elements are corrosion-active and even toxic. There-
fore, their practical achievement is not a simple problem. The storage, transportation,and distribution of
the final product - hydrogen - also are complicated.
The technological scheme for the transference of heat in a closed cycle with reversible chemical
reactions is interesting [6]. For one of them (steam-water conversion of methane), heat supply is neces-
sary but in the other (by the exothermic reaction of the products produced) heat can be liberated. This
peculiarity of the process can be used for the transportation of latent energy to considerable distances [7].
A high-temperature nuclear reactor supplies a chemical converter with heat, in which an endothermic
reaction takes place with the formation of H2 and CO2. Then these gases separately are transported to the
heat consumer. The gases enter a chemical reactor in which the chemically combined energy, in conse-
quence of the exothermic reaction process between H2 and CO2, is liberated and is fed to the consumer. No
loss of energy during transportation of these gases in the cold state occurs (with the exception, obviously,
of the expenditure of energy in transporting the gas).
In all energy-conversion processes, it is essential to ensure a pressure gradient directed from the
primary reactor circuit to the chemical production circuit under irradiation safety conditions, which re-
quires the construction of an intermediate coolant circuit between the reactor circuit (if a gas-cooled reac-
tor is used) and the technological plant. The problem of using high-temperature reactors for the produc-
tion of hydrogen is also a complex problem, falling outside the bounds of solution of the problem merely of
the generation of heat by a high-temperature nuclear reactor. Despite all the intricacy and the strict com-
plexity of these problems, their solution at present is urgent.
Courses for the possible application of high-temperature reactors, the temperature levels and pres-
sures of the processes of these courses which, at the level of our knowledge may be assumed to be formu-
lated, are generalized in Table 5.
Thus, in terms of the future requirements of the national economy, in the protection of the environ-
ment, and also taking account of the everywhere increasing intensity of the fuel-energy generation balance,
it can be assumed that the attraction for high-temperature nuclear reactors for solving on an economic
basis the problems facing the national economy will be justified in the near future. To what extent will the
status and prospective parameters of high-temperature reactor construction answer these problems?
The first project for a high-temperature reactor was accompanied in England by the startup of the
Dragon reactor at Winfrith in 1964. In April 1966, its thermal power was raised to the design power of
20 MW. A large series of tests on various types of fuel have been carried out in this experimental reactor.
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Fig. 3. In nuclear buildings of the USSR. Kol 'sk Nuclear-Power Station. Unitized con-
trol board. Photo by V. Bratchikova.
Fuel elements were studied under irradiation with direct measurement of the discharge of gaseous fission
products and discharged fuel assemblies. At present, according to the statements of the supervisors of the
Dragon reactor, there are prospects for obtaining a coolant temperature at the outlet from the active zone
of up to 1250?C and, in light of promising experiments with a fuel of uranium nitride particles, it is per-
missible to consider the possibilities of obtaining even higher temperatures (up to 1450?C). In the USA, the
experimental Peach Bottom nuclear-power station with a high-temperature gas-cooled reactor (HTGR)
was brought to its operating power at 42 MW in 1967. In the Federal German Republic (Julich), in 1967, an
experimental nuclear-power station with a HTGR of electrical capacity 15.6 MW was also constructed. The
next stage in the development of HTGR reactors was the construction in the USA of the Fort St. Wain nu-
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clear-power station. In January 1974 charging of the reactor with fuel was completed and the firm "Public
Service of Colorado" obtained a license for its commercial operation. The electrical capacity of the nuclear-
power station is 342 MW and the efficiency is 39.2%. At the present time, work is taking place to eliminate
troubles which have developed, in particular helium leakage.
It is planned to construct a large-scale nuclear-power station with a HTGR in the Federal German
Republic (Wentropp) with a capacity of 300 MW (el). The helium temperature at the outlet from this reactor
will be 750-800?C. So much for the prospects and possibilities of further increasing the temperature in
HTGH-type reactors.
Recently, reports have appeared concerning the cancellation of contracts on the construction of large-
scale nuclear-power stations with HTGR reactors, due to the increase of prices of plant and materials, and
the necessity for revising plans in the course of increasing the unit capacities of the reactors, which will
put back by several years their periods of construction. These reasons are of a situational and not a scien-
tific-technical nature.
At present, all efforts directed at the production of a high temperature are based on the utilization of
fuel elements with microparticles of fuel (uranium dioxide or dicarbide) and with a multilayered coating
of pyrolytic graphite and silica. The particle sizes, dispersed in the graphite matrix, are 200-2000 W.
These fuel elements are characterized by a high degree of retention of fission products (up to 130.0-1400?C).
The production technology is being developed and improved.
Fuel nuclei of uranium dicarbide do not react chemically with the pyrolytic graphite cladding and re-
main structurally stable up to 3000?C. However, carbides have a relatively high diffusion coefficient through
graphite. The upper limit of the working temperature for extended operation of microfuel elements with
a nucleus of uranium dicarbide coated with pyrolytic graphite is approximately 1700?C. The use of uranium
dioxide as the nucleus of a microfuel element allows contamination of the cladding of pyrographite to be re-
duced, as the diffusion coefficient of uranium from the dioxide in graphite is less than from the dicarbide in
graphite. Complex fuel compositions based on nitride compounds are being considered, which will permit
the working temperature of microfuel to be raised to 1700-1800?C. The possibility of designing a ventilzated
fuel element is also being studied, for the purpose of equalizing the internal pressure in the fuel element
with the pressure of the coolant, in order to reduce thereby the stress in the fuel element cladding.
The idea is noteworthy for removing heat from the reactor core and transferring it to a technological
requirement by means of radiative heat exchange with a coarse element solid coolant - graphitic elements
- the circulation of which is affected with a servomechanism. Radiative heat exchange and the application
of a kinematic servomechanism imparts a new quality to the reactor, which permits confidence in its effi-
cient utilization for the provision of heat for high-temperature technological processes:
1. The pressure inside the reactor vessel can be maintained both above and below atmospheric with-
out reducing the power intensity of the core.
2. The solid coolant (graphite) can be returned to the core at any temperature level.
3. The heat transfer surface between the solid coolant and the technological raw material does not ex-
perience mechanical stresses and can be made of heat-resistantbut not heat-proof materials, in-
cluding nonmetallic materials.
Because of these qualities, the operating conditions of thermotechnical plant and the shielding condi-
tions of the technological raw material from radiation contamination are significantly eased, and also the
transfer of heat from the core to the technological process is ensured over a narrow range of high tem-
peratures. Moreover, the feasibility emerges in principle for building on the basis of a reactor of this
type nuclear assemblies incorporating in a single vessel a nuclear source of high-temperature heat and a
technological consumer. According to preliminary estimates, the achievement of these possibilities will
permit the production of a tangible economic effect to be calculated.
The study and working out of other ideas is also possible, for example, making the helium coolant
dusty with finely dispersed graphite in order to reduce the pressure in the reactor circuit and improving
the heat transfer at the same time with convective and radiant heat exchange.
From all that has been said above, the conclusion can be drawn that the feasibility of constructing a
high-temperature economical nuclear reactor is an important factor in the future development of technical
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progress, not only in the field of electric-power generation but also in a number of national economical
fields, where fuel plays the influencing role. Moreover, extension of the spheres of application of the heat
from the fission of heavy nuclei, undoubtedly, will contribute to the solution of a number of important power-
economical problems.
LITERATURE CITED
1. V. N. Bobolovichetal., At. Energ. 36, No. 4,251 (1974).
2. G. Rajakovics, Atomwirtschaft, 20, No. 1, 24 (1975).
3. H. Harder and R. Fischer, Nucl. News, 55, March (1975).
4. Nuclear Engineering International, 20, No. 205, 115 (1975).
5. J. Pottier, New Frontier Energy, 143 (1975).
6. R. Schulten, K. Kugeler, and H. Barnest, Rev. Energie, 237 (1974).
7. W. Hafele, IIASA Report (1973).
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METHODS OF MATHEMATICAL MODELING AND
.OPTIMIZATION OF NUCLEAR POWER PLANTS
L. S. Popyrin UDC 621.311.2:621.039.001.2.003.1
The exceptional complexity of the technological structure of the contemporary national power indus-
try'and the interconnections between individual power units predestined its division into a number of sys-
tems. The unified national power system standing at the upper hierarchical level is divided into five basic
specialized systems which together guarantee the national economy all forms of energy and fuel. an elec-
tric power system, oil, gas, and coal supply systems, and a nuclear power system presently being formed.
These systems are divided into appropriate systems of individual regions of the country. The ultimate
elements of large power systems are businesses - power producers and consumers [1]. In turn each power
producer is a complex system which includes a large number of units of various types of power plants joined
by physical, technical, and transport ties. It is expedient to represent each such complex system as a
series of hierarchically coordinated systems.
Four hierarchical levels are generally distinguished in thermal power systems: thermal power plants,
power units, groups of plant elements, and plant elements.
The final elements of the hierarchical structure of a thermal power system - the plant elements
- must be detailed further for the study of individual phenomena, processes, and structures. These in-
vestigations are performed at lower hierarchical levels, i.e., at the level of physicotechnical systems of
parts of plant elements when the problems of the development of thermal power become problems of
mechanics, thermal physics, physical metallurgy, and other related disciplines [2, 31. The importance
and complexity of the problem of the optimum design and long-term development of thermal power sys-
tems of various types are obvious. Nuclear power plants are no exception. Figure 1 shows the system of
informational interconnections of a nuclear power plant arranged according to the hierarchical principle.
The external input data result partly from optimizing the power and economic systems at a higher
level; the nuclear power plant is one element. The external input data are obtained partly from predic-
tions and expert estimates. The internal input data include a descritpion of the laws and characteristics
of the flow of technological processes, physical properties of the working media and coolants, and char-
acteristics of different kinds of plant structures. Descriptions are given of the constraints imposed on the
parameters and characteristics of the structures. The input data also include lists of design layout types
for power plant equipment and variants of the form of its process flow diagram (or the conditions of their
formation).
The information obtained by solving the optimization problem includes integral and discrete charac-
teristics of the form of the process flow diagram of the power plant, continuously variable thermodynamic
and flow rate parameters of the energy carriers, discrete characteristics of the types of structures of
units and basic plant elements, and continuously or discretely varying structural parameters of plant units
and elements. In addition to its direct use to establish optimum parameters and characteristics of a nu-
clear power plant this information also appears as internal and external feedback data. The internal feed-
back data determine the direction of further development and improvement of investigations leading to a
lower step on the hierarchical ladder, i.e., to the level of physicotechnical models of individual phenomena,
processes, and structures. The external feedback data link includes optimum technical and economic in-
dices of the plant for various conditions of its use, technological, weight, and structural characteristics of
units and elements of the equipment of an optimized plant, characteristics of surges, industrial wastes,
heat release, and other factors which describe the effect of the power plant on external systems and the
environment.
Translated from Atomnaya Energiya, Vol. 40, No. 2, pp. 145-151, February, 1976. Original article
submitted November 6, 1975.
?1 976 Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. No part of this publication may be reproduced,
stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming,
recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00.
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External systems:nuclear power, mechani-
cal engineering, electric power, environ-
ment, etc.
Extemalin-
ut data
I
-~~
---------------i Required
Models of groups' unformation
~ _4
of plant elements
Model of nuclear
power plant
Internal in -
Models of plan
elements
;Intrnal -
]feedback
Ldatq
r--
Intemal physicomechanicalsystems of
parts of plant elements
Fig. 1. System of mathematical models
and exchange of information for the opti-
mization of a nuclear power plant and its
equipment; -; ---; -?-?) input, required,
and intermediate data respectively.
In addition to the indicated streams of external and
internal data there are direct and feedback streams of so-
called intermediate data circulating within the system of
mathematical models considered. These data streams con-
tain information on optimum values of parameters obtained
by solving optimization problems at various hierarchical
levels within the power plant.
This hierarchical structure of intermediate data
makes it possible to organize a two-step optimization
cycle - downward and upward. In this case there is taken
into account the decreasing influence of more detailed in-
formation on the refinement of overall optimum solutions,
and hence the possibility of decreasing the volume of data
exchanged.
The intermediate and sought for data can also be used
to investigate the content of the models and the structure of
the interconnections between them, i.e., to construct an opti-
mum system of mathematical models. In this case supple-
mentary cycles of interconnections arise between the sys-
tem of models and the sources of internal input data.
At the present time a qualitative solution of the prob-
lem of optimizing the values of the parameters, the form of
the arrangement, and the design of an atomic power plant
are impossible without the extensive use of mathematical
modeling and a computer, leading to a solution whose suit-
ability depends on the extent to which the time of comple-
tion and the capital outlay are taken into account.
Studies of the problem under consideration have been
directed toward developing a theory and methods of a com-
prehensive thermodynamic, technical, and economic analysis and complex optimization of a nuclear power
plant based on mathematical modeling methods and the solution of complex experimental problems and the
use of a computer.
So far the following [2-51 have been developed: theoretical bases for the construction of mathematical
models of various types of nuclear power plants for complex calculational studies; methodical bases for the
use of nonlinear mathematical programming and computers for solving the power plant optimization prob-
lem; practical ways of applying the methods of mathematical modeling, nonlinear mathematical program-
ming, and computers to determineways of increasing the economy of various types of atomic power plants
by the choice of optimum neutron, physical, thermodynamic, cost, and structural parameters, and also an
efficient form of the process flow diagram.
The mathematical formulation of the problem of the complex technical and economic calculation of a
nuclear power plant of a given form can be written as follows:
Z=Z[Xg, Yg(Xg),G]E. (1)
'Dg[Xg Yg(Xg)]EO=0;
Yg