SOVIET ATOMIC ENERGY VOL. 44, NO. 2

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• Publication Date: August 1, 1978
• Content Type: REPORT

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ISSN 0030.531 X
Russian Original Vol. 44, No. 2, February, 1978
August, 1978
SATEAZ 44(2) 111--'224 (1978)
SOVIET
ATOMIC
ENERGY
ATOMHAH 3HEPIWH
(ATOMNAYA ENERGIYA)
TRANSLATED FROM RUSSIAN
b CONSULTANTS BUREAU, NEW YORK
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Soviet Atomic Energy is a cover-to-covertranslation of Atomnaya
S O ET Energiya, a publication of the Academy of Sciences of the USSR.
ATOMIC
ENERGY.
Soviet Atomic Energy is` abstracted or in-
dexed in Applied Mechanics Reviews, Chem-
ical Abstracts, Engineering Index, INSPEC-
Physics Abstracts and Electrical and Elec
tronics Abstracts, Current Contents, and,
Nuclear Science Abstracts.
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 Atornnaya Energiya:
Associate Editor: N.A. Vlasov
A.A. Bochvar
N. A. Dollezhal'
V. S. Fursov
I. N. Golovin,
V. F. Kalinin
A. K.Krasin?
V. V. Matveev
M. G. Meshcheryakov
V. .B. Shevchenko
V.J. Smirnov
A. P. Ze f irov
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SOVIET ATOMIC ENERGY
A translation of Atomnaya Energiya
August, 1978
Volume 44, Number .2 February, 1978
CONTENTS
Engl./Russ.
JUBILEES
Seventy-Fifth Birthday of A. P. Aleksandrov ..... ...........:....... 111 107
Twentieth Anniversary of the International Atomic Energy Agency (IAEA)
- I. D. Morokhov .. .............. . ............ _ ........ 115 110
ARTICLES
Prospects for the Development of Chemical Technology of Factories of the
Nuclear-Power Generation Fuel Cycle - B. N. Laskorin, A. K. Kruglov,
D. I. Skorovarov, V. F. Semenov, B. A. Chumachenko, E. A. Filippov,
A. M. Babenko, and E. P. Vlasov ...............................12Y ' 118.
Nuclear Superheating of Steam, Results and Prospects at the Present Stage
B. B. Baturov, G. A. Zvereva, Yu. I. Mityaev,
and V. I. Mikhan ........................................... 131 126
The Principal Technical Problems and Prospects for the Creation of Gas-Cooled
Fast Reactors with a Power of 1200-1500 MW Using a Dissociating Coolant
- A. K. Krasin, V. B. Nesterenko, B. E. Tverkovkin, V. F. Zelenskii,
V. A. Naumov, V. P. Gol'tsev, S. D. Kovalev, and L. I. Kolykhan........ 138 131
Physicotechnical Aspects of Nuclear and Chemical Safety of Power Plants with
Gas-Cooled Fast N204 Reactors - V. B. Nesterenko, G.. A. Sharovarov,
S. D. Kovalev, and V. P. Trubnikov ............................. 144 137
Physical Properties of Fast Power Reactor Fuels and Their Effect on the Fuel
Cycle - 0. D. Bakumenko, E. M. Ikhlov, M. Ya.. Kulakovskii,
B. G. Romashkin, M. F. Troyanov, and A. G. Tsikunov ............... 147 140
Atmospheric Release of Volatile Fission Products from Operation of Nuclear Power
Reactors and Spent Fuel Reprocessing Facilities and Prospects for
Extracting the Products - B. Ya.. Galkin, L. I. Gedeonov, N. N. Demidovich,
R. I. Lyubtsev, I. V. Petryanov, B. F. Sadovskii, V. N. Sokolov,
and A. M. Trofimov .......... ....... .................... 153 145
Problems in Transporting Reprocessed Nuclear Fuel - A. N. Kondrat'ev,
Yu.. A. Kosarev, and E. I. Yulikov .... ....... .... .. ...... ... 158 149
BOOK REVIEWS
Yu. A. Surkov. Gamma Spectrometry in Space Investigations - Reviewed by
Yu. V. Sivintsev ......................... . ............... 163' 154
Development of Methods of Solidification and Burial of Radioactive Waste from
Fuel Cycle - V. V. Dolgov, B. S. Kolychev; A. A. Konstantinovich,
V. V. Kulichenko, B. V. Nikipelov, A. S. Nikoforov, Yu. P. Martynov, /
S. N. Oziraner, V. M. Sedov, and V. G. Shatsillo................... 164 155/
Principal Prerequisites and. Practice of Using Deep Aquifers for Burial of Liquid
Radioactive Wastes - V. I. Spitsyn, M. K. Pimenov, V. D. Balukova,
A. S. Leontichuk, I. N. Kokorin, F. P. Yudin, and N. A. Rakov.......... 170 161vZ
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CONTENTS
Engl./Russ.
DEPOSITED, ARTICLES
Calculation of Parameters of Weak-Signal Detection in Mass and Electron
:Spectrometers in Pulse-Counting Mode - M. L. Aleksandrov, M. S.. Kobrin,
and N. S. Pliss .......................................... . 179 169
Calculation of Parameters of Neutron Thermalization in Lead - Sh. Kenzhebaev 180 169
Thermal Expansion of Uranium Carbide with Additives Imitating Stable Fission
Fragments in 8% Burnup of Heavy Atoms - A. A. Ivanov, .
V. S. Belevantsev, Z. F. Evkina, V. A. Zelyanin, and S. N. Bashlykov.... 181 170
The 27A1(n, p) 27Mg Cross Section for 14. 9-MeV Neutrons - V. I. Melent'ev
and V. V.' Ovechkin......... ............................... 183 171
Interpolation Formulas for Calculating the Integrated Coherent and Incoherent
Scattering Cross Sections - O. S. Marenkov and B. G. Komkov ......... 184 172
LETTERS
One Error of the Radioisotope Method of Measuring the Continuity of a
Two-Phase Flow - V. A. Kratirov, A. N. Kazakov, V. S. Gurevich and -
N. A. Kukhin . ? ....... 185 173
Effect of Impurity on Sintering of Uranium Dioxide - V. I. Kushakovskii,
B. A. Zhidkov, and A. M. Loktev ....... . ...: .... 188 175,
Method of Graphical Calculation of Extraction Process for Systems with Two
Extractable Macrocomponents - A. M. Rozen, M. Ya. Zel'venskii,
and L. A.. Kasumova .. .
? 190 176
Quantum Yield and Electrons from the Cylindrical Casing of an Isotopic y-Ray
Source - R. V. Stavitskii, M. V. Kheteev, G. A. Freiman,
I. G. Dyad'kin, V. A. Velizhanin, L. A. Stulova, and E. V. Borisenkova...... 193
178
Efficiency of a and y Radiation in the Formation and Regeneration of El Centers
in Quartz - L. T. Rakov and B. M. Moiseev ... .................. . 195 180
Inversion Probes in Gamma-Gamma Methods V. A.. Artsybashev.......... 197 181
Activation of Molybdenum and Tungsten in 'a Cyclotron
- I. O. Konstantinov, V. V..Malukhin, N. N. Krasnov,
and A. D. Karpin ..... . ....... .. ...................... 200 183
COMECON CHRONICLES
Cooperation Diary . .... ............... .. ..........
203 186
BOOK REVIEWS
P. Zweifel. Reactor Physics - Reviewed by V. I. Pushkarev .. , , ? ? . ? ? .. , ? ? . 207 188
CONFERENCES AND MEETINGS
Session of Section of Physicotechnical Problems of Power Engineering, Academy
of Sciences of the USSR - Yu. Klimov ..... ............:......... . 208 189
First All-Union Conference on the Scientific-Engineering Foundations of
Waste-Free Production - V. N. Senin .................... .. 209 190
First All-Union Conference on the Analytical Chemistry of Radioactive Elements
- B. F. Myasoedov, A. V. Davydov, and N. P. Molochnikova .......... 211 191
Construction of Atomic Power Plant in Finland ........................ 213 193
Sixth Conference on Engineering Aspects of Lasers and Their Application
- V. V.. Aleksandrov and V. Yu. Baranov ........................ 215 194
Conference on Radioecology - Yu. B. Kholina ......................... 217 196
Seminar on the Use of Low-Potential Nuclear Heat - Yu. I. Tokarev ......... 219 197
NEW APPARATUS
Laboratory Apparatus with J3 Source for Research on Radiation-Chemistry
Processes - G. Z. Gochaliev, S. I. Borisova, S. L. Serkova,
D. N. Makhalov, and A. I. Yarkin ................. .. . . ... .... 222 199
The Russian press date (podpisano k pechati) of this issue was 1/ 20/ 1978.
Publication therefore did not occur prior to this date, but must be assumed
to have taken place reasonably soon thereafter.
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February 13, 1978, was the 75th birthday of that eminent Soviet physicist, Academician Anatolii Petrovich
Aleksandrov, President of the Academy of Sciences of the USSR, and Director of the I. V. Kurchatov Institute
of Atomic Energy.
Translated from Atomnaya Energiya, Vol. 44, No. 2, pp. 107-109, February, 1978.
0038-531X/78/4402- 0111$07.50 ?1978 Plenum Publishing Corporation 111
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A. P. Aleksandrov was born in the town of Tarasche in the Ukraine in the family of a teacher. Upon
finishing technical high school in Kiev he worked as an electrician. In 1923 he taught physics and chemistry
in school and at the same time studied in the Kiev University Department of Physics and Mathematics, from
which he graduated in 1929.
His first scientific paper "High-voltage polarization in ceresin," published in 1929, attracted the atten-
tion of Academician A. F. -loffe who invited Aleksandrov to the Leningrad Physicotechnical Institute (LFTI). It
was here that Aleksandrov became a scientist.
In his first years at the LFTI, Aleksandrov worked on dielectrics. . He did research on breakdown in di-
electrics and on the properties of polystyrene, a promising new material for electrical and radio engineering.
In the mid-1930s the foundations were being laid for a new science, the physics of polymers, In view of this,
it became of considerable practical, as. well as scientific, interest to ascertain the electromechanical proper-.
ties of polymers. It was precisely this area of research that attracted Aleksandrov most of all. Foreseeing
an enormous future for high-molecular compounds, together with his co-workers (and in the case of some
studies, in collaboration with P. N. Kobeko) he pursued physical research on polymers.
All of the investigations carried out by Aleksandrov during this period are characterized by an endeavor .
to extract the maximum practical results from fundamental research. This has been especially clear in his
subsequent work.
During the Second World War Aleksandrov was in charge of naval work to provide protection for ships
against magnetic mines by methods developed before the war in his laboratory. In addition to his immediate
co-workers, he was actively assisted in this work by many co-workers from other LFTI laboratories, includ-
ing I. V. Kurchatov. Protection for ships by this method made a great contribution to the successful opera-
tions of the Soviet navy.
It was in this period that the talent of Aleksandrov was forcibly revealed, not only as a scientifist but
also as an organizer of scientific-engineering development and design and as a skillful leader in the practical
implementation of such developments.
A profound knowledge of physics, the ability to see the engineering aspects of a problem and possible ways
of solving them, and authority as an attentive, benevolent, but at the same time strict and insistent person are
qualities which help Anatolii Petrovich solve major and responsible problems.
The year 1943 was noteworthy in the history of science .and technology of our country. That was the year
that Soviet physicists began work on a major scientific-engineering problem of the 20th century, that of har-
nessing nuclear energy. As is known, Igor' Vasil'evich Kurchatov was in charge of the scientific side of the
work. Aleksandrov was involved in the work with his laboratory and soon came to head a large body of scien-
tists and engineers.
The greatest development of the activities of Aleksandrov has been associated with the application of
atomic energy in many areas of the national economy. In'1948, when he was appointed deputy to Kurchatov,
Aleksandrov devoted his talent as a scientist and his great experience and energy to the development of reactor.
construction. His amazing versatility and erudition have been displayed in reactor development. An outstand-
ing physicist, he'has directed and organized the work of designers, technologists, materials scientists, and
electrical engineers, and with his brilliant comprehensioniof,all the details he has proposed solutions and
evaluated the results. Aleksandrov sees not only the general outline and the principal features of any design,
he also sees the fine details. Such an approach gives confidence that the solutions adopted are correct and
this is the approach he teaches to others.
Choice of clear-cut and feasible problems, sensible organization of research and experimental work,
his attraction for designers and industrial organizations in the early stages, and, finally, his enthusiasm
enable Aleksandrov to avoid the "submerged rocks" associated with the promotion of scientific advances and
to maintain close, fruitful ties with industry.
Under his scientific leadership, major scientific-engineering work has been done on the construction
of the atomic industry in the USSR. The construction of the first atomic power plants, the development of a
series of research reactors (VVR, SM, IGR, etc.) were the first successes on this road. Special mention
should be made of the fact that the construction of research reactors in various scientific centers of the country
has led to intense development of a number of areas of physics, biology, and chemistry.
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After the death of Kurchatov in 1960 Aleksandrov succeeded him as head of the Institute of Atomic En-
ergy. Under Aleksandrov the reliable and economic reactor plants VVER-440 and RBMK-1000 were developed
for atomic power plants and are now built in the Soviet Union and abroad.
While paying much attention to the development of concrete plants for the first atomic power installa-
tions, Aleksandrov clearly saw the prospects of further development of nuclear power and took care that the
results of atomic research be introduced on a broad scale in other branches of the national economy. In 1968
at the Seventh World Power Engineering Congress (Moscow) he said that.... "in the long term nuclear power
stands out as a power industry of multipurpose complex plants engaged in electricity generation and other
forms of production. . .. Clearly, the development and all-round extension of the forms of technology which
can be converted to nuclear energy resources is one of the cardinal practical tasks confronting our generation
along with the development of fast breeder reactors with a high breeding ratio...." These ideas are being-
actively developed at the I. V. Kurchatov Institute of Atomic Energy and in other organizations in the form of
new energy and technological reactor plants.
Aleksandrov was the initiator of the application of atomic energy in shipping. Under his direct guidance
and participation, high-quality marine power plants have been developed and built. Atomic icebreakers operat-
ing on the most difficult segments of the northern sea route have transformed the strategy and tactics for con-
voying ships. The atomic icebreaker Lenin, the world's first atomic-powered surface vessel, went into ser-
vice in 1959, and has been used to appreciably extend the shipping season. The atomic icebreaker Arktika,
fitted with an improved power plant, has reinforced the successes of the icebreaker Lenin; navigation in the
western sector has become almost year-long. In 1977 the Arktika completed its unprecedented voyage to the
North Pole in a record short time, thus showing that for our icebreakers there are no unattainable places in
the icy seas.
The expanse of the scientific interests of Aleksandrov is exceptional and hence the development of many
areas of basic and applied research, ranging from thermonuclear fusion to biology, within a single institute is
not surprising.
Aleksandrov has taken an unflagging interest in the physics of the condensed state, an area of science in
which he worked in his youth. This interest is heightened by the fact that the development of atomic science
and engineering has confronted solid-state physics with new questions and has at the same time placed in the
hands of researchers new equipment and methods for studying the properties of solids. Aleksandrov attentively
follows and supports work on solid-state physics both at the I. V. Kurchatov Institute of Atomic Energy and at
other research organizations of the country.
Along with this research, Aleksandrov supports and develops work on the practical application of super-
conductivity for the needs of atomic engineering and the national economy as one of the major directions of the
present scientific-technological revolution. And here once again one sees the ability of Aleksandrov to com-
bine scientific research with development for industry and by his knowledge and persuasion to unite sizeable
staffs of scientific, design, and industrial organizations for solving major scientific-engineering problems.
While he heads an institute with a huge staff and diversity of scientific-technological subject matter,
Aleksandrov looks after not only the construction of plant and the financing of work but, perhaps above all, is
concerned about maintaining an atmosphere of goodwill and of enthusiasm for the work. He has succeeded in
doing this by virtue of his enormous personal charm and extremely respectful attitude to each employee of the
Institute and his work, but, obviously, mainly by arousing enthusiasm for any unknown phenomenon, new
problem, or new instrument. To comprehend a new theory, to become aware of new experimental facts, and
to examine a different, nontraditional approach to any known problem are all important and interesting to
Aleksandrov.
Aleksandrov is an eminent specialist who has participated directly in the solution of a multitude of ap-
plied problems. He has widely advocated and collaborated in every way in the development of basic research.
An inexhaustible curiosity in basic research and his encouragement of such research enabled him to use a new
understanding of a physical effect or the ability to measure something to extract a more accurate method of
solving an important engineering problem. Aleksandrov rarely observes the official hierarchy when solving
scientific-engineering problems. In the evening in his office venerable academicians, as well as junior sci-
tific workers and senior and ordinary engineers, have their heads bent over drawings spread out on the floor
or over reports and they tell him about the results of an experiment that has just been completed or outline
ideas for a new experiment.
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The attention A.leksandrov pays to people is exceptional. No important matter, personal illness, or
fatigue could prevent him from immediately coming to the assistance of someone who has fallen ill and regu-
larly phoning in the evening to the home of a hospitalized colleague.
In 1943 A.leksandrov was elected Corresponding Member and in 1953, Academician to the Academy of
Sciences of the USSR. For 15 years Aleksandrov was a member of the Presidium of the Academy of 'Sciences
of the USSR and in 1975 he was elected President of the Academy. A.leksandrov has headed the Academy at a
time when the importance of scientific research in the life of society, especially a developed socialist society,
has been growing steadily, when there has been an extraordinary expansion of the areas of research and an in-
crease in the scale of activities of the Academy of Sciences, and a growth of the complexity of the tasks of the
Academy as the principal center of basic science and coordinator of scientific work in-the country.
With, a clear perception of?the responsibilities and enormous tasks put before Soviet science and the
Academy of Sciences of the USSR, A.leksandrov gives paramount attention to the choice of the most promising
directions of scientific research, to the concentration of scientific forces and material resources uponthe most
important problems of present-day science and current goals of technical progress.
Bearing in mind the character of scientific work under modern conditions, A.leksandrov is constantly
concerned with the development of the material and technical base of science, improving the level of equip-
ment, and automating research.
In this work as President of the Academy of Sciences of the USSR, Aleksandrov has displayed scientific
erudition, on the one hand, and a wealth of experience. of work in collaboration with industry, on the other
hand. Under the conditions today, when science has become ?a direct productive force, these qualitities of the
head of the Academy are extremely important in solving problems of the practical realization of scientific
achievements. -
An important part of his activities as President concerns the development of science in the republics
and in the branches and scientific centers of the Academy of Sciences of the USSR, refinement of planning of
research and development, and improvement of the administration of all academic scientific and institutions.
For meritorious service to the country's science and technology Aleksandrov has been made a Hero of
Socialist Labor on three occasions. He has been awarded the Order of Lenin eight times, the Order of the
October Revolution, and other orders and metals. Aleksandrov is a Laureate of the Lenin Prize and of State
Prizes of the USSR. At the Twenty-Third, Twenty-Fourth, and Twenty-Fifth Congresses of the Communist
Party of the Soviet Union (CPSU), A.leksandrov was elected member of the Central Committee of the CPSU.
A.leksandrov is a deputy to the Supreme Soviet of the USSR.
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TWENTIETH ANNIVERSARY OF THE INTERNATIONAL
ATOMIC ENERGY AGENCY (IA.EA.)
The International Atomic Energy Agency (IAEA) achieved its 20th anniversary in 1977. The IAEA is an
organization, which was founded by a group consisting of 60 countries, under the aegis of the United Nations.
The purpose of the IAEA. according to Statute is the achievement of "the more rapid and more widespread
utilization of atomic energy for the maintenance of peace, health, and prosperity throughout the whole world,
The Agency guarantees that assistance given by it or through its requirement or under its supervision
or control would not be used in such a way as to contribute to any military objective" [1].
The highest authority is the General Conference, at which each member-nation of this organization is rep-
resented by one delegate. The General Conference regularly, once per year, assembles in session. The
Statute provides for the convening of special sessions of the General Conferences according to the require-
ments of the majority of member-nations or the Controlling Council.
Between sessions the Agency is guided by the Council, consisting of 34 managers. It assembles at times
set by them (as a rule, 5 sessions/yr) and is guided in its work by the Statute of the IAEA and the resolutions
of the General Conference.
As the highest authority, the General Conference discusses any problems specified by the Statute, and
also selects the members of the Council of Managers, ratifies the acceptance of countries into membership
of the IAEA, considers the annual report of the Council of Mangers, approves the submitted budget, reports
of the Council for the United Nations, and also changes of the Statute, etc.
All information presented to the General Conference is considered and accepted by the Council of Mana-
gers. In addition, the Council appoints a General Director, who is then approved by the General Conference.
He is the principal administative person and directs the Agency Secretariat.
The IAEA budget is comprised from the obligatory payments of the member-nations which, in 1977,
amounted to 37 million dollars, and voluntary payments (amounting in 1977 to 6 million dollars), intended
for rendering technical assistance to developing countries.
During 20 years, the IAEA. has been transformed into an impressive international forum. Since 1957 the num-
ber of member-nations has grown from 60 to 110. In the work of the Executive - the Council of Managers - 34
countries now participate, as against 23 in 1957 and 25 countries in 1963. During this same period, the budget
has increased, and also the strength of its personnel.. At present, it amounts to about 1300 persons, of whom
approximately one-third are specialists, and the remainder are technical and auxiliary personnel.
At the end of September and the beginning of October, 1977, the Twenty-First Jubilee Session of the
General Conference of the IAEA took place in Vienna in the headquarters, which had conducted a total of 20
years of activity. The delegates listened with great satisfaction to the welcoming message of the General
Secretary of the Central Committee of the Communist Party of the Soviet Union, Chairman of the Presidium
of the Supreme Council of the SSSR, L. I. Brezhnev, in which, in particular, he said: "The problem standing
before the International Agency of promoting the widespread utilization of Atomic Energy for maintaining peace,
the health of the people and the prosperity of the nations, is close and understandable to us.
The Soviet Union actively cooperates and is ready to develop even further cooperation with other coun-
tries in the matter of the peaceful utilization of nuclear energy, included within the scope of the IA.EA.. Our
country, widely utilizing nuclear energy for constructive purposes, is ready to share its rich experience and
scientific-technical knowledge in this field, in the name of the future progress of mankind" [2].
*First Vice-Chairman of the State Committee for the Utilization of Atomic Energy in the Soviet Union.
Translated from Atomnaya Energiya, Vol. 44, No. 2, pp. 110-117, February, 1978.
0038-531X/78/4402- 0115$07,50 ?1978 Plenum Publishing Corporation 115
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Dept, of tech-.
nical aid and
pubis.
Devision of
tech. assis-
tance
Pubis. di-
vision
Secretariat bodies.
defining the lines
of activity
Dept. of tech.
operations
Service group
in the field of
peaceful nu-
clear explosions
Division of nu-
clear power gen-
eration and re-
actors
Division of nu-
clear safety and
protection of
the environment
Dept. of ad-
ministration
Bureau of verifi-
cation of ac-
counts and ad-
ministrative-
economic servi-
ces
Budget-
finance
division
Division of
external re-
lations
Dept. of scien-
tific research
and isotopes
Standard-
ization
section
international
center of theo-
retical physics,
Trieste
Joint division
FAO/IAEA on
the utilization
of atomic ener-
y in the food
industry and
agriculture
Dept. of safe-
guards and
inspections
Assessment
of safe-
guards ef-
fectiveness
section
Development
division
Division of sci-
entific- techni-
cal information
General ser-
vices divi-
sion
Translation
(jnterpreter)
visi n
Natural sciences
division,
Scientific re-
search and labo-
ratories division
Juridicial di-
vision
IAEA laborato-
ries
Personnel
division
Mona lisk labo-
ratoryTT
First
opera-
tions di
vision
Second
opera-
tions di-
vision
Data processing
division
*Under joint supervision of IAEA and UNESCO;
twith increased participation of UNESCO and UNEP.
Fig. 1. Organizational Structure of the IA EA. Secretariat.
Scientific - Technical Activity of IAEA.
Over 20 years, the IAEA has carried out major work in the field of the peaceful utilization of atomic
energy. For the assistance of member-nations, broad programs have been developed for research, for pro-
moting the development of nuclear power generation, exchange of scientific-technical information in the field
of nuclear science and technology, the application of nuclear explosions for peaceful purposes, ensuring the
safety of the environment, new sources of power are being mastered, such as controlled thermonuclear fusion,
etc. The Soviet Union has actively participated in the accomplishment of these programs.
The scientific-technical activity of the IAEA includes various programs on the introduction of nuclear
energy in the various fields of economics of the countries.of the world [3]:
The aim of the IAEA program, conducted jointly with the Food and Agricultural Organizations of the
United Nations (FAO), is the use of isotopes and radiations in the food industry and in agriculture. The pro-
gram is oriented on the application of nuclear methods for increasing agricultural production, and also for
raising the quality of food products and the protection of crops, domestic animals and foodstuffs from harmful
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TABLE 1. Growth of Power, GW (electri-
cal), and Power Generating Reactors
1975
1980
1985
Region
capa-
reac-
apa-
reac-
capa-
reac-
city
tors
city
hors
city
Itors
Europe
31,6
103
116,1
207
382,4
509
North America
52,5
77
124,8
149
299
278
Latin America
0,3
1
2,9
5
15,4
24
Africa
3,2
4
Asia and Austra-
lia
9,1
21
36,6
56
82,3
109
Total
93,5
202
280,4
417
782,3
924
Countries not
possessing nu-
26,9
69
111,7
179
278,7
376
clear weapons
insects, sickness and injury. Important results have been obtained already for increasing the fertility of soil,
due to the rational introduction of fertilizers and a water cycle, nuclear methods have been established and
continue to be developed for determining the protein content in seed cultures, which is extremely important for
increasing the quantity and improving the quality of protein by means of mutation induction, mutant selection
and the development of methods of selection; genetic, nutrient and agronomic assessment of the mutants has
been carried out.
This has been accomplished jointly with the World Health Organization (WHO), for promoting the develop-
ment of procedures and methods of using radioisotopes in medicine, biology and also for the preservation of
the environment.
The Physics Program consists of the following divisions: nuclear physics, the use of research reactors,
plasma physics and-controlled thermonuclear fusion, industrial application and the chemistry, testing and
analysis of materials, the production and industrial application of radioactive sources, nuclear data, atomic
and molecular data.
One of the most important programs is that of nuclear power generation and reactors. This program in
conjunction with the program on nuclear safety and protection of the environment occupies the greatest volume
in the scientific-technical activites of the IAEA.
The nuclear power generation program covers all aspects of this problem - from the forecasting of eco-
nomic questions to the study of improved methods of energy conversion. The program has such divisions as
nuclear material resources, surveying assessment, supply and demand; fuel cycle technology, including fuel
element technology, reprocessing of spent nuclear fuel and the handling of wastes; study of the regional cen-
ters of the nuclear fuel cycle, etc.
The program on the Nuclear Safety and Protection of the Environment has its aim in ensuring the safe
utilization of nuclear power and the protection of people and the medium from the injurious effects of nuclear
radiation from radioactive and nonradioactive effluents from nuclear facilities. Altogether, the work in the
establishment of standards of safety, recommendations and guidance, asistance, and service given to the
member-nations of the IAEA. on standards of radiation safety are well known to specialists. They are con-
sidered mainly as the national standards of safety in many countries of the world, including the Soviet Union.
The modes of achievement of the IAEA. programs are very varied: symposia and conferences, active
working groups and groups of experts, meetings of specialists, etc. In this connection, the special impor-
tance for the future development of world nuclear power generation of the Salzburg Scientific-Technical Con-
ference on Nuclear Power Generation and Its Fuel Cycle, held in May 1977, should be mentioned. The con-
ference showed that the solution of the immediate and future points of the problem are being approached in dif-
ferent ways in the world, which is explained by the special features and requirements of the economics of in-
dividual countries. This discussion on the routes and tendencies of the development of nuclear power genera-
tion should be continued.
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TABLE 2. , Number of Plants for Repro-
cessing Fissile Materials
For prod, of fuel from uranium
24
36
For prod, of mixed uranium-plu-
21
26
tonium fuel
For enriched uranium
.10
For reprocessing spent fuel.
12
The.Information and Technical Services to, member -nations and the Secretariat occupy a special place in
the activities of the IAEA.
The development of an automated system of collection and distribution of scientific-technical informa-
tion (ISIS,system) is a great achievement. The system, created on the initiative of the Soviet Union, started
to operate in 1970 and has developed rapidly in recent years. The number of items processed annually has in-
creased from 4000 in the first year of operation to 65,000 at the present time. Now the ISIS system is caused
by 46 member-nations and 13 international. organizations. The ISIS Atomindex is a unique international refer-
ence journal on nuclear science and technology.
The IAEA. has available a library with a large stock of specialist literature. It has also connections
with national libraries and there is a high rate of exchange of literature according to enquiries from member-
nations and the Secretariat.
The IAEA carried out a widespread publishing activity and issues the journal "Thermonuclear Fusion,"
the series "Reviews on Atomic Energy," a monthly Bulletin, and also the proceedings of conferences, sym-
posia, etc.
The Soviet Union participates actively and directly in the scientific-technical activities of the IAEA,
sending its own specialists on scientific-technical' and organizational means,. directing highly qualified scien-
tists, specialists and administrators to work in this organization. The Permanent Representation of the Soviet
Union at international organizations in Vienna renders great assistance in liaison and cooperation with the
IAEA. The participation of the Soviet Union in the work of the IAEA wins high praise from the Secretariat and
member-nations. The role, importance and authority of the Soviet Union in the IAEA, undoubtedly has grown,
especially over recent years.
Technical Assistance to Developing Countries
One of the first places in the activities of the IAEA is occupied by the rendering of technical assistance
to developing countries, which includes the transmission of technical knowledge and skills in the fields of utili-
zation of nuclear energy for peaceful purposes, support for efforts toward a more efficient achievement of work
in the field of nuclear power generation and ensuring that the transmitted technical skills and knowledge could
be applied after rendering this assistance.
The modes of the rendering technical assistance are diverse: services of.experts, provision of plant,
granting of scholarships, and training of national personnel.
Since 1958, 82 countries have utilized the services of 3000 experts and detached specialists. During
this period, 20 million dollars worth of plants and materials have been supplied, 3000 scientists, engineers,
and administrators have carried out training in more than 180 regional and interregional training establish-
ments.
In attaching great importance to the rendering of technical .assistance to developing member-nations of
the IAEA, the Soviet Union has supplied to these countries at the requests of the Secretariat, plant and ma-
terials to the account of its voluntary payments, and has also trained national personnel.
From 1969 to 1976, of the total sum of voluntary payments of the Soviet Union of 2.8 million rubles in
the national currency of the IAEA, more than 2 million rubles already has been realized. On the account of
this payment, 15 scientific-familiarization trips of specialists from developing countries have taken place.
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TABLE 3. Installed Capacity of Nuclear Power
Stations, Number of Facilities, and Quantity
of Nuclear Materials under Safeguards of the
IAEA (on Jan. 1, 1977)
11973 11974
Installed nuclear power
station capacity, GW (el.)
No. of nuclear power stations
Other reactors
Facilities for the manufac-
ture of fuel elements and
for the chemical repro-
cessing of fuel
Other facilities or zones of
material balance
Total facilities
Plutonium, kg
Enriched uranium, tons
element
isotope
Raw material, tons
5
8
10
20
27
36
43
60
107
110
103
120
20
26
29
35
140
254
288
315
365"
4730
6300
9035
12 000
1865
2305
3096
5 000
43
53
66,7
150
3370
3910
4440
6000
From 1977 annual .trimonthly courses on the application of nuclear, methods to agriculture will be held
in the K. A. TimiryazevAU-Union Agricultural Academy in Moscow. In 1978-1979, it is planned to organize a
course at the Novovoronezh Nuclear Power station on the operation of water-cooled/water-moderated reactors.
The possibility is being considered of founding annual courses in Moscow on the application of nuclear methods
in medicine. For the first time, a scientific-technical tour has been organized and successfully conducted
on safeguards, with a visit to nuclear facilities of the Soviet Union.
On the recommendation of the government, the,Soviet delegation declared at the Twenty-First Jubilee
Session of the General' Conference 'of the"IAEA. an increase in the voluntary payment of the Soviet Union to the
technical assistance fund, in the first place to developing country-participants of the Treaty for the Nonprolif-
erationof Nuclear Weapons. This payment may be used for the purchase of Soviet plant, instruments and ma-
terials, and also for conducting IAEA educational-familiarization arrangements in the Soviet Union.
The effective combination of technical assistance with the necessary control measures will serve for
the further consolidation of the policy of nonproliferation of nuclear weapons and, consequently, a more com-
plete realization of the problems arising from the IAEA. Statute and the conditions of the Nuclear Weapons Non-
proliferation Treaty.
The Problem of Nonproliferation of Nuclear Weapons
It should be pointed out, however, that even if the activities of the IAEA. in cooperation with the wide-
spread introduction of atomic energy into the peace economics of member-nations of this organization do not
prove fruitful, in the modern setting there is no more urgent problem than the cessation of the arms race and
disarmament. The IAEA. acknowledges cooperation in the achievement of these aims. At the moment, it is
impossible to forget that the energy of the atomic nucleus can be used also as the most destructive weapon
which mankind has ever known. Therefore, the efforts undertaken by the IAEA for the prevention of nuclear
weapon proliferation acquire special importance.
At present, it can be seen with all authenticity that the development of nuclear power generation is pro-
ceeding with increasing rates, and an even greater number of countries are included in its orbit. Undoubtedly,
its intensive development will allow the greater part of all forms of energy requirement to be ensured and will
allow economy in the use of the large quantity of organic raw material for those purposes where its total re-
placement is more complicated, mainly for the chemical industry.
At the same time, in considering the positive aspects of development, it must not be forgotten that the
significant increase of the quantity of fissile materials and the number of countries possessing them increases
the potential hazard of using the accumulated nuclear materials for the creation of nuclear weapons.
Estimates show that the average doubling time of the world's nuclear power generation capacity in the
next 2-3 decades may amount to 5 years, and the installed capacity of nuclear power stations expected by 2000
A. D. may amount to 4. 106 MW (electrical). Even if these development times prove to be low but com-
mensurate with the increase of all power generation as a whole, the capacity of nuclear power stations by 2000
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A. D. will amount to 2. 106 MW (electrical). However, even this minimum estimate shows the considerable
scale of its growth [4].
The distribution of capacities and the numbers of nuclear power stations throughout the regions of the
world in the forthcoming decade are shown in Table 1.
Thus, in 1985 the capacity of nuclear power stations in countries which do not possess nuclear weapons
will have increased by a factor of 10, and the number of countries possessing nuclear power will have doubled.
The considerable increase of nuclear power -stations leads to an increase of requirements for uranium,
which will have increased from 25,000 tons in1975.to 35,000 and 160,000 tons by 1980 and 1985, respectively
Significantly, the requirements for enrichment will increase from 13,000-tons of sep. work units/yr in 1985,.
to 100,000 tons of sep. work units/yr in 1985; fuel manufacture will increase from 6000 tons in 1975 to 15,000
and 30,000 tons in 1980 and 1985. By 1980, more than 150 tons of plutonium converted to fissile fuel willhave
accumulated, and by 1985 this figure will amount to 504 tons.
It should be mentioned that the increase in the number of nuclear facilities is not identical in all stages
of utilization of nuclear material and its reprocessing. Thus, if the number of nuclear power stations in-
creases by more than 200 units by 1980, and by 600 units by 1985, in comparison with 1975, then over this
same period only a few new uranium enrichment plants and plants for reprocessing spent fuel will appear
(Table 2).
L. I. Brezhnev, in the salutory address at the Twenty-First Jubilee Session of the IA.EA. General Con-
ference, wrote: "In supporting, the development of the peaceful utilization of atomic energy, the Soviet Union
is firmly resolved, together with other governments, to consolidate in every way the international policy of
nonproliferation of nuclear weapons. It is essential-to do everything possible in order that the international
exchange of nuclear technology, involving in many countries a scientific-technical and industrial nuclear
potential, does not become a channel for the proliferation of nuclear weapons.
"We cannot shut our eyes to the fact that in the world there will always be powers who would wish to re-
ceive in their hands nuclear weapons, in order to threaten nations with this weapon. Therefore, the problem
of setting a reliable safeguard on the paths of nuclear weapon proliferation, andfor! preventing the hazard of
a nuclear war, remains now just as acute as ever.
"In solving this. problem of immediate importance, the International Atomic Energy Agency has played on
important role, and we express the hope that the IAEA will apply all efforts to ensure that the atom will serve
only the interests of peace. "
Future- consolidation of an international policy of nonproliferation, today as never before, is important
and is connected directly with the maintenance of peace, safety and reduction of the threat'. of nuclear war.
The accelerated development of nuclear power generation, which is becoming one of the principal sources
for satisfying the power generation requirements of countries, is related inevitably with the accumulation of
large quantities of nuclear materials and, as a consequence, with an increase of the danger of nuclear weapons
proliferation. The Soviet Union proceeds from the fact that the development of nuclear power generation in the
world must be combined to the fullest extent with consolidation of the nonproliferation policy.
All governments who value peace highly, must actively strive for the Treaty on the Nonproliferation of
Nuclear Weapons to become a genuinely universal instrument of international nonproliferation politics, en-
compassing all governments without exception. Unfortunately, not all countries who possess nuclear weapons,
nor all countries with significant nuclear potential, have subscribed to the Treaty, and some of them, as for
example the UAR, in fact are opposed to this Treaty and are actively preparing to carry out nuclear tests.
The campaign for a new stage of the nuclear arms race: being conducted by certain western circles un-
der the catchword of expansion of production of the so-called neutron bomb and other dangerous types of wea-
pons, does not assist consolidation of the Treaty for the Nonproliferation of Nuclear Weapons.
System of Safeguards
During the 20 years of existence of the IAEA., considerable experience of monitoring activities has been
built up. A system of legal standards has been worked out, monitoring equipment has been set up, procedures
and methods of monitoring have been developed and introduced at many types of nuclear facilities. At the
present time, the IAEA. monitors the activity of many nonnuclear countries of the world. This is done com-
pletely regularly, because one of the functions of the IAEA., fixed by its Statue, is the implementation of
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safeguards which have their aim in ensuring "that special fissile or any other materials, services, plant,
technical facilities and data, presented by the Agency either according to its requirement or under its super-
vision or control, should not be used in such manner as to further study any military purpose, and to extend,
according to the requirement of the parties, the use of these safeguards to any two-party or multiparty agree-
ment or, according to the requirement of one or other government, to any forms of activity of this government
in the field of nuclear energy. "
The system of safeguards was formulated for the first time in 1961 in the form of INFCIRC-66 and con-
tained monitoring procedures for small experimental reactors. Since then, it has been extended and modernized
repeatedly, which has been reflected in other documents. The key stage was the decision of the participating
countries of the Treaty on the Nonproliferation of Nuclear Weapons to guarantee to the IAEA the implementation
of monitoring functions according to Article III-I of the Treaty, in accordance with the proposals of INFCIRC-
153. Thus, at present, the IAEA. monitors nuclear activity resulting from the agreements concluded on the
basis of INFCIRC-153 and INFCIRC-66, Rev. 2. Based on INFCIRC-153, 44 agreements are concluded, of
which 21 are with countries which do not possess a significant nuclear activity. Based on INFCIRC-66, Rev. 2,
agreements on projects (11) and transfer of safeguards (21) are operative, and also agreements in connection
with single-party organization of nuclear activity under safeguard (8).
Under the control of the IAEA, there are about 12 tons of plutonium, 5000 tons of enriched uranium and
about 6000 tons of raw material (Table 3) [5].
Under the conditions when nuclear power generation in the world is developing and international trade
exchange of nuclear materials and plant is expanding, the improvement of IAEA activities in the field of safe-
guards is being promoted in the first plan of a number of measures directed at consolidation of the policy of
nonproliferation of nuclear weapons. The Soviet Union considers the efficient monitoring of the IAEA as one
of the principal premises for widespread international cooperation in the field of the peaceful utilization of
atomic energy.
The IAEA is entering at present a new stage of its monitoring activity, the characteristic feature of which
is a sharp increase of the volume and complexity of monitoring. In connection with this, the problem of the
maximum use of all possibilities set out in the system of safeguards arises in all its acuteness. At the basis
of the system, as is well known, lies the principle of independent verification. The IAEA. must use this en-
tirely in its own right, independently of the extent of the development of registration and monitoring in indivi-
dual governments of groups of governments. Moreover, it will be necessary in all countries using IAEA moni-
toring, that efficient systems of accounting and control of nuclear materials should be created and operated.
The subsequent achievement by the IAEA and by countries of the regulations laid down in the IAEA system of
safeguards is a pledge of effective international control in the field of nonproliferation of nuclear weapons.
As before, the question of the necessity for radical improvement of operation of the IAEA. monitoring
machine is acute. Recently, the Department of Safeguards and Inspection was reorganized. A. second inspec-
tion division was set up and a section for assessing the effectiveness of the safeguards, intended to play the
leading role in stepping up controls. It is important to strengthen the Department of Qualified Specialists and
to raise to a new level the cooperation between its divisions and sections.
The necessity for a comprehensive analysis of the activities of the IAEA. control machine has become
imminent, and the implementation of long-term and short-term plans for its improvement. This would give
the capability of more reasonably approaching a definition of the necessary manpower and financial resources,
and would stimulate on a planned basis the development of procedures and methods of control, instruments and
plant used in monitoring activities, and their operative introduction into practice, especially at the present
time, when the IAEA is approaching achievement of safeguards in a number of large-scale facilities, which
are "sensitive" from the point of view of nonproliferation of nuclear weapons. The question of the development
of a model of effective safeguards also has been put on the agenda.
Due to the increase of volume of monitoring activities of the IAEA, the question of the volume of data
received by the IAEA is important. Until recently, processing and analysis of this information received insuf-
ficient attention. The creation in the Department of a special division for the processing of information on
safeguards, the development and operative introduction of an automated system of data processing, in principle,
is of great value for the entire system of control.
The formulation of the problem of implementing within the framework of the IAEA a project for an inter-
national convention concerning the physical protection of nuclear materials, plant, and transportation is urgent.
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In attaching great importance to the activities of the IAEA in the field of safeguards, the Soviet delega-
tion made a statement at the Twenty-First Jubilee Session of the IAEA General Conference about the purpose-
ful contribution introduced by the Soviet Union in the implementation of the technical aspects of safeguards in
1978 to the amount of 300,000 rubles in national currency. This contribution may be used, in particular, for
conducting training for inspectors at the Novovoronezh nuclear power station, development of technical meth-
ods of monitoring at this nuclear power station, and for the organization in the Soviet Union of IAEA confer-
ences and courses on safeguards. The Soviet Union, for their part, is prepared to render further assistance
to the IAEA in work on the strengthening of the system of safeguards, which is important for peace [6].
It would be desirable to mention that governments who supply nuclear materials, plant, andtechnology
should assume a special responsibility. Rigorous safeguards will be necessary, so that international coopera-
tion in the field of peaceful utilization does not become a channel for the proliferation of nuclear weapons.
This is not a commercial problem, but one of politics and safety. It is well known that a group of 15 supplier-
countries of nuclear materials, plant, and technology have implemented guiding principles for nuclear export.
At the conference of suppliers held in Sept. 1977 in London, understanding was reached to inform IAEA through
its General Director concerning the policy followed by them for nuclear export control.
The guiding principles are intended as an obligatory condition for the granting to nonnuclear countries of
export services, the official assurance of the government of the recipient-country that the imported nuclear
materials, plant, and technology enumerated in the reference list supplied by the exporter-countries will not
be used for the creation or production of nuclear explosive devices. The guiding principles require from the
recipient assurances for the physical protection of the articles of the references list received, if it accepts
the safeguards (monitoring) of the IAEA not only on the transferred items, but also on the materials and plant
produced by means of the items received. The guiding principles provide for special IAEA control in the case
of export of facilities, plant, and technology for the enrichment of uranium, and reexport regulations, in which
cases reexport may be effected only with the agreement of the original exporter and in the same conditions of
initial supply, and other regulations including sanctions inthe event of violation by the recipient of the conditions
of the guiding principles for nuclear export. In addition to this, the exporters have been obliged to render ac-
tive assistance for improving and increasing the effectiveness of the monitoring (control) activities of the IAEA.
The task of-intensifying control measures during report will be continued. With regard to the Soviet
Union, it will subsequently strive for the acceptance of a principle of -total control as a condition of supply of
any materials, plant, and technology included in the agreed reference list.
Being a specialized international organization, the IAEA reacts tactfully to the political changes in the
world. The scientific-technical direction of this organization is subjected to the influence of these political
problems which stand before mankind. An example of this is the activities of the IAEA in consolidating the
conditions for the nonproliferation of nuclear weapons, etc. It is important that concern about the assurance
of peace on earth and the safety of mankind from a nuclear catastrophe are the initiating elements in the activi-
ties of the IAEA, and here the words of the salutory address of L. I. Brezhnev at the Twenty-First Jubilee
Session of the IAEA General Conference are pertinent: "The Soviet Union, for its part, will even further render
total cooperation to the IAEA inthe achievement of the noble aims, which stand before this authoritative inter-
national organization.
1. IAEA Statute, 1963.
2. L. I. Brezhnev, Address to the Participants in the Twenty-First Session of the International Atomic
Energy Agency's General Conference [in Russian], Pravda, Sept. 29, 1977.
3. Agency Program in 1977-1982 and Budget in 1977 [in Russian], GC(XX) 567.
4. U. Panitkov, Forecast of World Nuclear Activity, Vienna, IAEA/STR-40 (1974).
5. I. D. Morokhov, R. M. Temirbaev, M. N. Ryzhov, and V. P. Kuchinov. International Safeguards for
the Nonproliferation of Nuclear Weapons. Report to the International Conference onNuclear Power Gener-
ation and Its Fuel Cycle [in Russian], Salzburg, May 2-13, 1977, IAEA.-CN-36/340.
6. I. D. Morokhov, Statement of the head of the Soviet delegation in general discussion at the Twenty-First
Session of the IAEA General Conference, Sept.27, 1977, Vienna [in Russian], IAEA., GC(XXI)/OP. 194.
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PROSPECTS FOR THE DEVELOPMENT OF CHEMICAL
TECHNOLOGY OF FACTORIES OF THE
NUCLEAR-POWER GENERATION FUEL CYCLE
B.
N.
Laskorin, A. K. Kruglov,
UDC 621.039.54
D.
B.
A.
I.
A.
M.
Skorovarov, V.
Chumachenko,
Babenko, and
F.
E.
E.
Semenov,
A. Filippov,
P. Vlasov
The necessity for the development of the nuclear industry in the Soviet Union [1] is conditioned by the in-
creasing demand for power, the continually expanding use of radioactive isotopes for the intensification of tech-
nological processes in chemistry, control and automation of the various branches of industry, the use of the
achievements of nuclear science and technology in agriculture, medicine, geology, and for controllable contami-
nation of the atmosphere caused by concentrated sources of energy. All this is accompanied by an increase of
the role of chemical and radiochemical processes in the treatment of natural uranium raw material and the re-
generation of spent fuel, in the production of new types of fissile material and in other factories of the nuclear-
power generation fuel cycle. Let us consider the achievements and future prospects for the development of
these processes.
System Analysis and Mathematical Modeling of
Production Development
In a nuclear-power generating complex, the decisive circuit is that of the fuel cycle, representing an
assembly of interrelated different plants. The fuel cycle consists of four stages of the total technological pro-
cess, each of which includes one or several plants.
The first stage is the manufacture of the nuclear fuel: extraction of uranium or thorium, concentration,
production of uranium concentrate and uranium hexafluoride, isotope separation, fuel component manufacture,
and fuel elements.
The second stage' is the combustion of the nuclear fuel in reactors.
The third stage is the cooling of the spent fuel and its transportation to the reprocessing site.
The fourth stage is the reprocessing of the spent nuclear fuel (in closed cycles); extraction of valuable
components, manufacture of uranium-plutonium fuel, reprocessing and storage of waste. The following
plants occur in the structural layout of fuel cycles: structural materials for nuclear power station reactor
cores, specialized plant, instruments for monitoring radioactive materials, and also spent fuel-element stor-
age and production tailings during isotope separation.
In order to determine the prospects in detail of the second alternatives of production development of a
nuclear-power generating complex, taking account of new technology and types of reactors, system analysis
of the fuel cycle structure is of great importance. System analysis permits one:
to establish the mutual effect of the plants entering into the fuel cycle;
to show the technicoeconomic significance of each plant from the point of view of the long-term develop-
ment of nuclear power generation;
to reveal the varied development factors of each plant and to establish their interrelation;
to determine the system of limitations when considering different alternatives and to select optimization
criteria for production development.
Translated from Atomnaya Energiya, Vol. 44, No. 2, pp. 118-125, February, 1978. Original article sub-
mitted August 11, 1977.
0038-531X/78/4402-012,'3$07.50 ?1978 Plenum Publishing Corporation 123
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The analysis shows that the multivariability of production development is determined by the type of reac-
tor, the form of the nuclear fuel (uranium, thorium, uranium-thorium, uranium-plutonium, etc.), the regen-
eration technology of the spent fuel elements, and also the treatment of the natural raw material, the structure
of the capacities of the separation plant, which is characterized by the feasibility of using different physico-
chemical methods for the separation of uranium isotopes, and other factors.
These factors of the alternative development of individual production plants govern the fuel cycles. The
choice of alternatives for their implementation is determined by the econimic competitiveness of each alterna-
tive, the balanceability of operation of the fuel cycle plants, the supply of raw materials and materials in short
supply, and the readiness of industry for ensuring the production of the fuel cycle with the necessary facilities.
In order to investigate and optimize the alternatives for the development of the nuclear-power generating
fuel cycle and to choose from them the best in different countries, including also the Soviet Union, mathemati-
cal models have been developed [2-6]. According to their nature, they are subdivided into optimization and
simulation models. The first of these permits an all-round analysis of the effect of various factors in their
interrelation in the development of the nuclear-power generation system (taking into account both the inherent
special development features of nuclear power generation, and also interrelation with the fuel-power generat-
ing economy of the country). By means of simulation models, the effect of individual factors on the develop-
ment of nuclear power generation can be investigated.
At present, it is advantageous to construct an interrelated set of mathematical models. Such a combina-
tion of models makes it possible:
to consider the large number of alternatives for the production development of the fuel cycle;
to compare acceptable alternatives and, taking into account their limitations, to recommend the best of
them according to the chosen critiera;
to allow for the large number of influencing factors;
to carry out complex and laborious calculations for forecasting and estimating long-term development
alternatives;
to operatively correct previously made calculations in proportion to the accuracy of the starting data
(technological parameters of a different kind, technicoeconomic indices and restrictions, etc.), and to
change the production structure; '
to plan effective paths of scientific-technical progress and improvement of the fuel nuclear-power
generating cycle.
Taking all this into account, it should.be mentioned that for processing in detail of complicated valid
decisions for determining the prospects of development of the nuclear-power generation fuel cycle like a large
production-economic system, it will be necessary to use methods of program-objective planning and system-
mathematical analysis. This approach allows a dynamic model of planning and control to be established in
a development process and the introduction of new industrial technology and nuclear reactors, with an assess-
ment of the long-term direct and indirect consequences of the solutions used.
It will be interesting to consider the prospects of development of certain production plants of the nuclear-
power generation fuel cycle, taking account of the achievements of nuclear technology both in the Soviet Union
and also abroad.
Processing of Uranium Raw Material
Forecasts of the development of nuclear power generation indicate a significant increase of capacity
during the next decades; because of this, the requirement on uranium increases with every year. Therefore,
the importance of bringing into line the scale of possible extraction and processing with the known natural
resources increases.
In the Soviet Union, the most diverse problems have been solved successfully in the processing of ura-
nium rawmaterial and the prevention during processing of contamination of the environment. Theoretical cal-
culations, laboratory investigations, semi-industrial tests and industrial practice substantiate the effective
application of radiometric concentration to the majority of low-grade uranium ores [7]. Further reduction of
the cost of sorting and improvement of the technological indices are possible by the use of new, higher-output
separators and the utilization of methods based on the use of the artificial radioactivity of the ores.
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Low-grade, resistant, and complex ores are processed by using autoclave processes. The use of ele-
vated temperatures and pressures, together with the cheapest of oxidants (atmospheric oxygen), permits a
profitable processing of the ore raw material to be organized, and permits a high uranium extraction to be ob-
tained with a reduction of the consumption of reagents (e.g., acids) and a reduction of power costs (steam).
Commercially manufactured autoclave equipment provides for carrying out oxidized leaching processes of
uranium over a wide range of temperatures, pressures, and reagent concentrations.
In recent years, uranium from low-grade ores is being extracted on even greater scales by leaching out
the useful component at the site of the ore deposit. The uranium in this case is extracted from the depths to
the surface in the form of a solution. Soviet scientists reported for the first time on these investigations at the
International Conference on the Processing of Low-Grade Uranium Ores, held in 1966 in Vienna. Underground
leaching at present has been fashioned into a self-sustaining chemicotechnological process [8]. A technology
has been developed which is intended for the recovery of uranium from hard (massif) rock and from sedimen-
tary ores, deposited in stratified conditions (horizontal strata). Underground leaching has permitted the capi-
tal costs on production organization to be reduced, the cost of uranium production to be reduced, and the work-
ing conditions to be improved considerably. Moreover, the possibility has been given of processing local
small-scale ore deposits, to include in the processing compensated ores treated by the usual method of me-
chanical extraction, and also deposits lying in complex mining-geological conditions. Experience in industrial
operation shows that different low-grade uranium ores can be processed by this method.
New possibilities in the processing of low-grade and complex uranium ores are opened up by sorption
processes [9].
The irrefutable advantages of these processes are due both to the aggregation state of granular ion-ex-
changers, which permit the separation process to be conducted easily, and also to the high exchange capacity
of the majority of resin types. This, even at the beginning of the 1950s, permitted sorption processes from
pulp to be carried out, which are predominant in the uranium industry of the Soviet Union.
The development of a filtrationless sorption method has led to the development of leaching and extraction
desorption processes, which intensifies the uranium ore recovery processes and considerably improves the
technicoeconomic indices, due to the elimination of laborious operations of repeated filtration and repulping
of the filter cakes. The method has made it possible to include lower-grade uranium raw material in the pro-
cessing, and simultaneously to separate valuable components: molybdenum, vanadium, rare-earth elements,
scandium, and phosphorus [10].
Industrial experience has been built up of sorption from dense pulps up to solid:liquid = 1:1, which has
led to an increase of productivity of the operative plants by a factor of 1.5-3, an increase of uranium extrac-
tion by 5-10o, an increase of work productivity of the basic workers, and a reduction of the consumption of
chemicals, auxiliary materials, electric power and steam by a factor of several. In essence, an efficient
technology, continuous in all its links, has been created with total and complex automation of the process,
high-productivity equipment of large unit capacity with mechanical and pneumatic mixing for high-density pulp,
and also equipment for the continuous regeneration of saturated sorbent.
The ionites manufactured in the Soviet Union with weakly acid and strongly basic exchange groups can
be used for almost any (including even complex salt) systems. The production of granular ionites with high
kinetics properties, sorption capability and a high mechanical strength, has expanded the use of ion-exchange
processes. The production of new types of ionites-ampholites has permitted simultaneously the extraction of
attendant elements.
In the Soviet Union ion-exchange resins have been produced for sorption from pulp and solutions and,
especially the production of strongly basic anionites of helium structure AM, AMP, VP-1A, VP-3A, macro-
porous AMp, AMPp and VP-lAp, bi- and polyfunctional anionites of the type AM-2B, medium-basic AM-3
and VP-1p, and also extremely promising carboxyl ampholites AMK, AMK-2, VPK and various phosphorus-
and phosphorus- nitrogen- containing ionites (ampholites A.FI-5, AFI-7, VPF-1, and VPF-2). These ionites
have a high mechanical strength, which ensures. minimum losses of resin under the most rigorous operating
conditions [11].
When processing low-grade ores by underground and mound leaching, solutions are obtained with a low
uranium content. An even lower uranium content is characteristic for natural and mine water. In order to
extract uranium from the large volumes of solutions with a low concentration, an equipment has been designed
which makes it possible to carry out the process at a high linear flow-rate of the solutions.
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It is well known that extraction with organic solvents, from the point of view of physical chemistry, is
similar to sorption with solid ionites. The most efficient and optimum regions of application of each of these
processes has been determined from a comparative assessment. Sorption extraction from pulp usually is com-
bined with extraction processing of the. desorption solutions. Depending on the salt content of the solutions and
the problem of supply, a suitable extractant can be used. For reextraction, it is most advantageous to obtain
a uranium salt directly from the organic phase.
With the development of high-capacity equipment,- it has become possible to carry out extraction directly
from ore solutions. Two types of extractors, as usual, occupy the predominant position in the equipment lay-
out of the processes - mixer-settlers and columns. The main trend of the future improvement of extractors
consists in the search for optimum mixing conditions.
A considerable reduction of capital and operating costs can be achieved with extraction directly from
dense pulps and nonaqueous-leaching. However, these operations have still not emerged from the semi-indus-
trial stage and test-rig experiments.
New possibilities in hydrometallurgy are being opened up by the creation of processes which combine the
advantages of sorption and extraction methods. Sucn methods are the impregnation with organic solvents of
porous granules, and desorption of uranium from solid ionites with acids or with neutral extractants.
Considerable research has been undertaken by Soviet scientists on the extraction of uranium from natural
water with granular sorbents .[12]. In the process of investigation of selective sorbents, more than 400 dif-
ferent ionite samples have been tested. The most efficient were found to be certain strongly basic anionites,
with a capacity amounting to 2.5-5.3 mg/g. The regenerates, obtained by desorption of the anionites saturated
with uranium, are reprocessed by extraction or sorption concentration.
The scientific-technical level achieved at the present time will permit the most diverse problems in the
field of uranium raw material processing to be solved and will prevent contamination of the environment.
Wide possibilities in the inclusion of low-grade uranium-containing raw material in processing are being
opened up by the extraction-of uranium as a by-product or as a joint product in combination with other useful
components.
The Soviet Union has available great production experience in the extraction of uranium and other valuable
components from phosphate raw. material, and also in the complex utilization of uranium-molybdenum ores.
Isotope Separation
At the present time, requirements for enriched uranium are met by gaseous diffusion plants, which are'
linked with a large demand on electric power [13]. With the development of nuclear power generation, interest
is increasing with the realization of the possibilities of centrifuging, which is characterized by a significantly
lower power requirement. The Federal Republic of Germany, Holland, and Great Britain have concluded an
agreement on joint cooperation of separation plants with centrifuges. Investigations on the technology of cen-
trifuging are being carried out in Japan.
The first work on the chemical and ion-exchange separation of uranium isotopes is related to the end of
the 1940s. In 1953 a report appeared onthe enrichment of uranium inthe light isotope up to 2.8%by the ion mobility
method. The separation of isotopes by the precipitation of oxalates with countercurrent migration is described.
Ion-exchange chromatography, carried out by the use of anionites and cationites, occupies a special place
(including solutions of phosphorus- and nitrogen-containing complex-forming agents), and also water and or-
ganic solutions, including hydrochloric, sulfuric, nitric, or chloric acid solutions of uranium (VI), uranium
(IV) or their mixture. The isotope separation factor varied from 1.00006 to 1.0004. In the majority of cases,
the results of the work on the separation of uranium isotopes are only satisfactory, the latter associated with
an insignificant coefficient not exceeding 1.001. It is true that Japanese scientists have achieved an increase of
the 235U content over a single cycle by a factor of 1.017, by filtration of a solution through a sulfocationite JRA-
120B in a column of height 1 m and with a cross section of 1 cm2. Fractions enriched in 235U emerged pri-
marily from the column [14].
On the whole, sorption processes for uranium isotope separation have been widely investigated in the
U.S.A., Yugoslavia, France, and the UAR.
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The kinetics of the electron exchange of 235U and 238U, in the four- and six-valent states is being studied
in aqueous or organic solutions (TOA. and TBP*) in the presence of cationites and anionites. The purpose of
these investigations is the achievement of a maximum rate of exchange for the subsequent use of suitable sys-
tems in ion-exchange fractionation of isotopes.
Extraction processes for the separation of uranium isotopes have been less studied. The achievement
of a single separation factor of 1.002-1.00006 has been reported. In recent years, scientists in France and
other countries have published the results of investigations into the separation of isotopes by extraction, which
allow the advisability of further exploration in this field to be judged.
Ion-exchange and extraction methods of 235U and 238U separation could play an important role in the crea-
tion of a single water cycle for the regeneration of reactor fuel elements of low-enrichment uranium fuel. The
solution of the problem of increasing the rate of electron exchange between isotopes in the ionite phase and the
development of a high-capacity continuous chromographic process is imminent. The well-known extraction
systems do not yet provide acceptable uranium isotope separation factors, although they are characterized by
a high speed of attaining equilibrium. Efficient organic complex-forming agents and new principles for the
organization of phase flows in the stripping and enrichment lines will be necessary.
Since 1970 reports have been appearing about the separation of isotopes by laser. Great attention was
paid to this at the Eighth International Conference on Quantum Electronics (San Francisco, 1974) and at the
International Conference on Uranium Enrichment Methods (London, 1975).
Laser separation of isotopes includes the stages: introduction of the starting material into the system,
selective excitation, and extraction. As the starting material, at present the vapor of a mixture of isotopes
in atomic and molecular forms is being used; there are indications, however, of the possibility of using start-
ing materials in both the liquid and solid states [15].
Isotope separation is effected by laser by means of the selective excitation of the isotopes. In the inter-
action of radiation with a mixture of two isotopes, one of them is resonantly excited, while the other remains in
the ground state. The excited isotope can be extracted by various physicochemical methods (photon ionization,
photodissociation of the molecules, spatial separation of an atomic beam, chemical reactions). The most
widely developed method is photon ionization. The extraction of laser-excited isotopes by means of chemical
reactions is considered to be the most promising for industrial application.
The laser method is characterized by a high separation factor, which permits the same degree of enrich-
ment to be achieved with a considerably smaller number of process stages; the degree of enrichment is sharply
increased and the content of 235U in the tailings is reduced to 0.03% [16]. In addition, with laser technology, .
the required power is proportional to the quantity of separated isotope, and not the starting material, as in the
case of the methods being used at present, and therefore it is the least energy-consuming. The power consump-
tion in the separation of a single atom of 235U by different methods confirms this: gaseous diffusion, 3000 keV;
centrifuge, 300 keV, and laser 100 keV.
The economic efficiency of the isotopes of certain metals (zirconium, iron,. etc.) obtained by laser tech-
nology, which have a low neutron absorption cross section, for the manufacture of fuel-element claddings
should be mentioned; this leads to a significant improvement in the use of neutrons and to a reduction of the
requirements on the degree of uranium enrichment. Moreover, 232U and 233U can be separated by laser, which
increases the efficiency of the uranium -thorium cycle, as it permits the use of 232U as a radioisotopic source
of heat. The industrial achievement of uranium isotope separation by laser. technology is possible by 1985 [17].
Information has appeared on the possibility of commercial laser isotopic separation of plutonium earlier
than for uranium isotopes. According to estimates of different researchers, laser technology is the most eco-
nomical for the separation of highly radioactive 238Pu from its other isotopes in spent nuclear fuel (the cost of
1 g of high-quality 238Pu is reduced from 1300 to 125-250 dollars) [18]. 238Pu is used as a compact energy
source, e.g., for satellites and cardiological devices.
The introduction into production of laser technology will permit not only the utilization of uranium in ther-
mal reactors to be improved, but will also permit optimization of the isotopic content of the fuel of breeder-
reactors. Thus, 240Pu undoubtedly can be more usefully used in breeder-reactors than in thermal reactors.
*Trioctylamine and tributylphosphate.
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In investigations in this direction, great attention has been paid to consideration and comparison of the
various technological factors which determine the quality of the pelleted cores of uranium oxide, and discus-
sion of methods of producing and using granulated oxide fuel in vibropacked fuel elements [19].
Thermal reactors of nuclear power stations, operating on enriched (2-4%) uranium, usefully use only
- 1% of the required natural uranium.. Therefore, in order to increase the utilization of natural uranium in
the period preceding the bringing into operation of fast reactors, there is considerable interest in the conver-
sion of thermal reactors partially to a plutonium fuel cycle, which provides for the repeated utilization of plu-
?toniumbasedon a mixed uranium-plutonium fuel. Numerous investigations of the technological and economic
aspects of this process [20-26] have shown promise for the development of a cycle with this fuel in thermal
reactors:
In the Soviet Union, as in other countries, investigations are being carried out on the design of fuel ele-
ments based on uranium-plutonium fuel for fast reactors [3, 27], including experiments on the irradiation of
these fuel elements up to high burnups in the reactor [28]. Different technological schemes for obtaining a
mixed oxide fuel for fast reactors are being analyzed.
In connection with the development of high-temperature, gas-cooled reactors and fast reactors, carbide,
nitride, phosphide and other fuel compositions have been investigated [29, 30]. Many papers are devoted to
the study of the physicomechanical, radiation, thermodynamic and other properties of these refractory uranium
and plutonium compounds. In the investigations, an important place is being assigned to the manufacture of fuel
elements based on microparticles (uranium dioxide or dicarbide) with multilayered protective coatings of
graphite (of different density) and silicon carbide. The microfuel elements, when inserted in a graphite matrix,
are grouped into elements of different geometry (rods, plates, and spheres) [30], and are characterized by a
high degree of fission product retention (up to 1300-1400?C) [31].
The buildup at present of experience in the technology. of manufacture of microfuel elements with coat-
ings makes it possible even now to obtain coolant gas temperatures in nuclear reactors at 1000?C and some-
what higher. The production technology of microfuel elements is being advanced continuously. Although at
the present-day stage of their production they are coating more than rod-type fuel elements, there is a basis
for hoping that future improvement in the, technology of manufacture of coated particles will bring their cost
near to the cost of fuel-element rods. This permits microfuel elements with coatings to be considered as ex-
tremely promising fuel for future nuclear power stations.
Carbonitride fuel is considered to be the most promising for fast reactors. Possibilities are being de-
veloped for improving the technology for the production of carbides from oxides, and the design of continuous
technological processes for obtaining carbonitride fuel, including also in granulated form, is promising.
Regeneration of Spent Fuel
In the nuclear fuel cycle, its regeneration is one of the most complex and most important technical prob-
lems. Regeneration remains one of the tightest points in the fuel cycle, from the point of view of guaranteeing
production capacities essential for satisfying the requirements in the bulk production of fuel for nuclear power
stations.
The industrial method of reprocessing the fuel from thermal reactors, which is unique in world practice,
independently of its composition and degree of irradiation, is the continuous counterflow extraction of uranium
and plutonium with solutions of tributylphosphate into diluents. The differences in the individual extraction
schemes consist in the number of cycles of extraction purification, in the separation of uranium and plutonium
in the first or second extraction cycle, in the method of separation, operations for the intercycle treatment of
the uranium solutions, the presence of a nodal point in the final purification of the uranium (on silica gel,
titanium phosphate, etc.), methods of concentration and refining of the plutonium.
The number of extraction cycles depends on the activity of the starting solution, which is determined by
the type of fuel, depth of burnup, and cooling time. With approximately equal conditions, the decisive factor
is the level of development of technology in a given factory, consisting in the correct choice of the optimum
influence of factors which affect purification from fission products, such as the degree of saturation of the ex-
tractant with uranium, the acidity of the eluted solutions, temperature, the use of complexing agents, time of
contact between phases in the extraction plants, chemical and radiation stability of the extractant and diluent,
and the removal of certain fission products in preparatory operations.
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Technicoeconomic requirements are met most completely by extraction schemes which ensure the follow-
ing basic indices [32]:
Purification factor:
uranium from plutonium .... ....... ................ 107
plutonium from uranium ........................... 106
uranium from fission products ....................... 107
in the first cycle ............ ................ ... 2 ? 104
in the second cycle. ........... ................ 5. 102
plutonium from fission products ...................... 10$
in the first cycle ....... ............... ....... 2- 104
in the second cycle .............................. 2. 103
during anion exchange ............................ 3
Extraction of uranium and plutonium, % ............ . . .. . . . 99.9
Degree of regeneration, %
nitric acid . .......................... ...... ... 95
extractant ..... ................. ............. 99.7
Nonaqueous methods of regeneration of spent nuclear fuel (sublimation, pyrometallurgical processes,
etc.), although quite well studied, at present have not reached the stage of industrial application.
Processes combining both aqueous and nonaqueous methods, e.g., the aqua-fluor-process, have proved
to be interesting. The most important difference in the known variations of the process consists in the pre-
cipitation and separation of the valuable components. The aqua-fluor-process, with the extraction cycle for
the combined extraction of the actinides into the organic phase and their purification from fission products at
the head of the technological scheme, has the advantage over other alternatives [33]. The preliminary separa-
tion of the fission fragment elements from uranium and the transuranic elements considerably simplifies the
direction and control of the entire process. Control is simplified in the operations for correcting and stabiliz-
ing the valence forms of plutonium and neptunium, and the solution of problems of the volatility of ruthenium
in the zone of dehydration of the uranium product after removal from it of plutonium and neptunium is not elimi-
nated, but is considerably facilitated. There is no need in the plant for any additional measures due to the
buildup in the fluoride and separation zones of fluorides of the main mass of fission products and the origina-
tion of heat release as a consequence. The fission products are removed in the aqueous raffinate, which can
be subjected to direct thermal concentration. The total purification factor from fission product elements
in the extraction cycle amounts to 103-106.
The aqua-fluor-process permits spent nuclear fuel of any type to be regenerated: metallic (uranium,
plutonium, thorium, or their alloys), oxides, carbides, nitrides, silicides, etc. However, the prospects for
its industrial utilization are doubtful, because it is inferior to extraction methods in its technological indices
and it leads to the formation of additional solid radioactive wastes.
In connection with the planned program of nuclear power generation development, the Soviet Union has
worked out the principles for locating the establishments for regenerating the spent fuel from nuclear power
stations, storage and transportation of the burnt fuel elements, and protection of the environment [34]. Tech-
nological schemes for the combined regeneration of spent fuel elements from nuclear power stations-with ther-
mal and fast reactors provide for the use of extraction [34, 35] and sorption operations [36] during regeneration
of uranium, plutonium, and also neptunium, americium, and other valuable elements. In this case, consider-
able attention is being paid to the dissolution, radiation chemistry of aqueous and organic solutions, extraction
and ion-exchange separation of macroquantities of plutonium and uranium, and the use of water-soluble neutron
absorbers.
Extraction processes have been studied for the extraction, separation, and purification of uranium, plu-
tonium, and neptunium in different valence states, using tertiary aliphatic phosphine and arsine oxides [37-39],
amides of carbonaeous and phosphoric acids, phosphazo compounds [40], and phosphazines [41]. Thus, the
investigation of the extraction capability of normal and isomeric tertiary aliphatic esters of phosphoric acid
showed a higher chemical stability of tri-isobutylphosphate, a thermoselectivity of trialkylphosphates in the
extraction of uranium and the transuranic elements from nitric acid solutions, and an inversion of the reaction
capability of the transuranic elements in extraction equilibrium states [42].
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As a result of a number of improvements, it became possible to achieve high indices in aqueous pro-
cesses of spent fuel regeneration. Suffice to say, that in a single extraction cycle for regenerating the fuel
elements of water-cooled/water-moderated power reactors, almost the complete separation of uranium, plu-
tonium, and neptunium has been achieved, and the purification factor of uranium from fission products
amounted to 5. 105-106.
The fluoride method has been improved considerably, which was demonstrated by the experimental re-
generation of the spent uranium fuel from the BOR-60 core with a burnup in excess of 101/o, and a cooling time
of 3-6 months. Stripping of the.BOR-60 and BR-5 fuel with alloyed cladding has been carried out; the distribu-
tion of uranium, plutonium,-and fission product elements throughout the plant of the technological circuit has
been studied.
However, despite this, -the unique methods-of regeneration that have received widespread recognition are
extraction using a system based on tri-n-butylphosphate..:(pyrex-process) and sorption based on strongly basic
anionites in the refining.
The main problem at present is the further increase of the economy and effectiveness of technological
schemes of operating, under-construction, and planned factories. Scientific-research development should be
directed at increasing the purification of the valuable elements from fission products, the choice of the opti-
mum ratio between extraction and sorption operations during regeneration, determination of the resources of
the possible operation of extraction and sorption systems without their replacement or regeneration, and also
on increasing the purification of plutonium and neptunium in the final refining operations and the use of fire,
explosion, and nuclear safety systems. The latter is especially important, since emergency situations [43] in
the majority of cases have been determined by the properties of the extraction and sorption systems in opera-
tion.
1. Data from the Twenty-Fifth Congress of the Communist Party of the Soviet Union [in Russian], Politiz-
dat, Moscow (1976).
2.
3.
4.
N.
A.
Dollezhal' et al., At. Energ., 31, No.3, 187
(1971).
M.
P.
Dergachev et al., At. Energ., 43, No. 5, 365
(1977).
V.
F.
Semenov et al., in: Handling of Nuclear Information.
Proceedings of Symposium, Vienna, IAEA,
279 (1970).
5. V. V. Batov and Yu. I. Koryakin, Economics of Nuclear Power Generation [in Russian], Atomizdat,
Moscow (1969).
6. V. N. Bobolovich, At. Tekh. Rubezhom, No. 3, 3 (1974).
7. M. L. Skrinichenko et al., Report at the International Conference of IAEA on Nuclear Power Generation
and Its Fuel Cycle [in Russian], Salzburg, May 2-13, 1977, IAEA=CN-36/321.
8. A. P. Zefirov et al., Fourth Geneva Conference, Soviet Report No. 459 [in Russian] (1971).
9. G. A. Kovda, B. N. Laskorin, and B. V. Nevskii, in: Soviet Nuclear Science and Technology [in Rus-
sian], Atomizdat, Moscow (1967).
10. B. N. Laskorin et al. , At. Energ., 43, No. 6, 477 (1977). _
11. B. N. Laskorin et.al., At. Energ., 43, No. 6, 472 (1977).
12. B. N. Laskorin, Tsvetnye Met., No.8, 15 (1975).
13. K. Khigasi, Uranium Enrichment, Short translation into Russian from Japanese, Atomizdat, Moscow
(1976).
14. B. N. Laskorin et al., Usp. Khim., No. 5, 761 (1975).
15. A. A. Sazykin et al., At. Tekh. Rubezhom, No. 3, 19 (1977).
16. Sci. News, 105, No.25, 396 (1974).
17. Nucl. Week, 15, No. 44, 2 (1974). -
18. Laser Focus, 12, No. 14, 26 (1976).
19. F. T. Reshetnikov, At. Energ., 43, No. 5, 408 (1977).
20. D. Deonigi, Nucl. Technol. , 18, No. 2, 80 (1973)..
21. D. Brite, Nucl. Technol., 18, No.2, 87 (1973).
22. Energia Nucl., 15, No. 1, 60 (1973). -
23. R. Smith et al., Nucl. Technol., 18., No. 5, 97 (1973).
24. C. Brown et al., Nucl. Technol., 18, No. 5, 109 (1973).
25. V. M. Abramov et al., At. Energ., 36, No.2, 113 (1974).
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26.
V.
M. Abramov et al., At. Energ., 36, No. 3, 163 (1974).
27.
A.
K.
Kruglov, At. Energ., 40, No.2, 103 (1976).
28.
I.
S.
Golovnin, At. Energ., 43, No. 5, 412 (1977).
29.
T.
S.
Men'shikova et al. , Fourth Geneva Conference, Soviet Report No. 454 [in Russian] (1971).
30.
V. Ya. Novikov et al., At. Tekh. Rubezhom, No. 6, 14 (1974).
31.
N. A. Dollezhal' andYu.I. Koryakin, At. Energ., 40, No.2, 133 (1976).
32.
Plutonium, Handbook, O. Vik (editor) [in Russian], Atomizdat, Moscow (1971).
33.
O.
Erlandson and B. Judson, U.S.A., Patent No. 3374068 (1968).
34.
V.
V. Fomin et al., At. Energ., 43, No. 6, 481 (1976).
35.
P.
I. Ivanov et al., Report at the International Conference on Nuclear Power Generation and Its Fuel
Cycle [in Russian], Salzburg, May 2-13, 1977, IA.EA CN-36/318.
36.
V. I. Anisimov et al., At. Energ., 42, No. 3, 191 (1977).
37.
B. N.
Laskorin et al., Fourth Geneva Conference 1971, Soviet Report No. 443 [in Russian].
38.
B.
Laskorin et al., J. Radioanal. Chem., 21, 65 (1974).
39.
B.
N. Laskorin et al., At. Energ., 28, No. 5, 383 (1970).
40.
D.
I. Skorovarov et al., Radiochemistry. Abstracts of Reports No. 1 [in Russian], Nauka
Moscow (1975)
,
,
p. 246.
41.
D. I. Skorovarov et al., Radiokhimiya, 18, No. 1, 29 (1976).
42.
E. A. Filippov et al., Dokl. Akad. Nauk SSSR, 234, No. 1, 117 (1977).
43.
F. Mi
lest, Isotopes Radia. Technol., 6, No. 4, 428 (1969).
NUCLEAR SUPERHEATING OF STEAM, RESULTS
AND PROSPECTS AT THE PRESENT STAGE
B. B. Baturov, G. A. Zvereva,
Yu. I. Mityaev, and V. I. Mikhan
Testing of the extended operation of the superheating channels (SC) of the Beloyarskaya Atomic Electric
Power Plant (BAEPP) has shown convincingly the economy of nuclear superheating of steam. The channels
being operated at the BAEPP with a steam temperature up to 565?C at the exit confirmed their high reliability
with a fuel depletion of 35 kg/ton and a calendar term of service of 6-7 years. These data allow acceptable
economy to be obtained for an atomic electric power plant (AEPP) in comparison with a thermal electric power
plant (TEPP), notwithstanding the relatively large number of neutron absorbers in the active zone.
The use of SC with fuel-element rods in which the amount of steel per unit mass of uranium is reduced
but the catalyst is excluded from the fuel composition permits improving the engineering-economic charac-
teristics of the channel reactor when nuclear superheating of steam is produced in it.
The results of the operation of the AEPP have been supplied in a report, and the prospects for nuclear
superheating have been discussed as an example of the sectional-modular high-power reactor (RBMKP), in the
design of which problems of this type in energy reactor construction, which is important from the standpoint of
saving uranium and significant reduction of thermal discharge, have been solved most completely.
The idea of obtaining superheated steam directly in a nuclear reactor attracted attention in the very first
stages of energy reactor development. Already in 1950 during discussion of possible alternatives to the reactor of
the first AE PP in the world (Obninsk) an alternative with nuclear superheating of steam was considered [1], but it was
postponed as technically insufficiently prepared. The successful start-up in 1954 and operational test of the reactor
of the first AEPP served as the basis for realization of the idea of nuclear superheating of steam having high
parameters in the most powerful energy reactors. Great interest in nuclear superheating was exhibited in the
USA, West Germany, England, Sweden, and other countries; however, the long-term test of the operation of
the I. V. Kurchatov BAEPP is the most impressive in the industrial sense.
Translated from Atomnaya Energiya, Vol. 44, No. 2, pp. 126-131, February, 1978.
0038-531X/78/4402- 0131$07.50 ?1978 Plenum Publishing Corporation 131
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Fig. 2
Fig. 1. Arrangement of the fuel channels in a (a) circular and (b) rectangular active
zone: 1, 2) evaporative and superheating sections; 3) reflector units.
Fig. 2. Engineering layout of the unit: 1) reactor; 2) evaporative channel; 3) superheat-
ing channel; 4) separator; 5) turbine unit; 6) condenser; 7) condensing pump; 8) conden-
sation purifier; 9) low-pressure preheater; 10) deaerator; 11) feed pump; 12) high-pres-
sure preheater; 13) superheating regulator; and 14) circulation pump.
The most suitable reactor in the constructional sense for obtaining high-parameter superheated steam is
the channel-type, in which the separate organization of the evaporative and superheating zones, which should
in the general case have different physicostructural characteristics and operating properties, is solved more
simply in comparison with reactor vessels. These zones should provide, in particular, the necessary ratio
of power to evaporation and steam superheating.
Nuclear superheating in connection with the use'of a single-circuit layout with direct supply of steam to a
turbine and the operation of thermomechanical equipment on active steam determined the advisability of the use
of tubular-type fuel elements as a first step in reactors with nuclear superheating; such elements have already
shown reliability in the operation of the reactor of the first AEPP. The standard parameters of traditional
power engineering for steam were selected, viz., 510?C and 90 kgfkm2.
Construction of the channel-type water-graphite reactor which was adopted for design studies corres-
ponded to the greatest extent to the problem posed, with the past experience and the possible outlook taken into
account.
Peculiarities of Nuclear Superheating of Steam. Nuclear superheating of steam has a number of positive
qualities. Nuclear superheating, together with the. possibility of the use of standard thermomechanical equip-
ment, provides a high thermodynamic efficiency to a facility, which lowers the consumption of nuclear fuel
and the discharge of heat per unit of generated electrical energy and reduces the thermal emission into the
environment. The latter fact takes on especially important meaning in connection with a significant increase
in the total energy production and an increase in the concentration of AEPP in industrially developed regions,
in particular in connection with estimating the possible ecological consequences resulting from the effect of
the heat discharge on the temperature conditions of the environment. This effect is still difficult to measure
in financial terms, but its significance increases in proportion to the growth of the energy supply, and it is
impossible to disregard it:
The choice of a water-graphite channel reactor permits providing:
freedom of installation in the reactor of fuel channels of various purposes and differentiated action on
the physical and heat-engineering characteristics of the active zone (Fig. 1);
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0 1,9 V 49 12,2 143 18,4 21,4 24,5 27,6 30.2 33.8
MW ? days/channel
Fig. 3. State of the steam-superheating
channels of the second unit on Jan. 1, 1976:
1) channels operating in the reactor; 2)
channels removed from the reactor due to
the absence of reactivity; and 3) channels
removed from the reactor for defects in
the fuel elements.
through-channel overloading for more effective use of the fuel in the case of a sufficiently good equaliza-
tion of power distribution throughout the active zone;
the use of various designs of the fuel (removable and nonremovable) channels, sleeve and rod fuel ele-
ments (see Fig. 1);
the use of a progressive single-circuit engineering layout with the input of steam from the reactor to a
turbine (Fig. 2); and
enlargement of the individual power capacities of reactors on the basis of standard elements without
fundamental restrictions from above both from technical reasons and from the point of view of safety.
The operating possibilities of this type of reactor are distinguished by great flexibility. The output of a
reactor with nuclear superheating of steam into the energy cycle can be accomplished without the use of out-
side heat sources.
The existing objective tendency towards reconsolidation of the energy supply diagram can increase the
requirements on the adjustability of the energy units. The engineering and economic characteristics of reac-
tors with nuclear superheating permit considering them as potential semipeak energy sources [2].
The introduction of nuclear superheating is positively expressed in the characteristics of the heat engi-
neering portion of the unit, since the reliability of turbine operation is increased due to the elimination of the
possibility of moist steam entering it. In this connection the layout of the turbine unit is also simplified due to
rejection of intermediate separators and superheaters. The use of high-speed turbines (3000 rpm) in connec-
tion with the enlargement of the individual capacities of the turbine units to 1.2-2.0 million kW, as well as
tapping the heat for central heating and industrial needs, has turned out to be theoretically possible.
Principal Problems of Organization of Nuclear Superheating of Steam. The most important scientific-
engineering problem in creating a reactor with nuclear superheating is the development of fuel elements which
would permit producing steam at a temperature of 500-540?C, a pressure of 90-130 kgf/cm2, and thermal loads
up to 1 ? 106 kcal/m2 ? h with acceptable neutron-physics characteristics and an economically practical depletion
of the uranium.
The physical problems of creating such a reactor, in addition to providing for uranium depletion (when
significant unproductive neutron absorption in the SC is present) acceptable on economic grounds, are included
in the maintenance of an equalized energy distribution and the ratio of capacities for producing and superheating
steam necessary for a thermal balance. In this connection the physical characteristics of the reactor should
provide for safety of the transition and start-up modes, in particular, an acceptable reactivity effect upon con-
version of a SC from water cooling to steam, and vice versa.
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OL .,CE-
420 440 460 480 500 520 540 560
Temp. of superheated steam, ?C
Fig. 4. Temperature of steam at output from the superheating channels:
Date of measurement
Electrical capacity
of the unit, MW
Pressure in steam
pipe, kgf/cmZ
Average superheating
temp., ?C
a b
C
Feb. 27, 1975 Aug. 2, 1972
Sept: 17, 1972
196
170
172
75
73
72
515
496
497
An important problem is providing for reliable operation of the reactor, fuel channels, and fuel ele-
'ments in steady-state and transitional operational modes under variable-load conditions as well as for accept-
able reliability of the main subassemblies and systems of the reactor based on a 20-30-yr useful life.
Nuclear superheating affiliated with the single-circuit thermal layout has determined the high level of
requirements on provision of radiation safety for the staff, in particular, for the machine room when the tur-
bines are operating on radioactive steam.
I. V. Kurchatov Beloyarskaya AEPP (BAEPP). The problems noted for nuclear superheating have es-
sentially been successfully solved in the designs and upon the construction of the first reactors of the BAEPP.
The experimental checking of the most important elements of the reactor, physical characteristics, ther-
mal hydraulic processes, and transitional engineering conditions was conducted on special test stands and in
the experimental loops of the Obninsk AEPP [1, 3].
The powering-up of the first reactor with nuclear superheating and an electrical capacity of 100 MW oc-
curred in 1964, followed in 1967 by a second reactor with a capacity of 200 MW; the gross efficiency of both
units was 37-38%. The reactors are identical in the structural sense, and they differ only in the capacity and
the external engineering layout. Up to now the total working time of both installations amounts to - 21 reactor-
years with an acceptable installed capacity usage coefficient of 62-77% and time coefficient of .75-91% [4].
One should note that the supplying of steam (20 Gcal/h) for heating a settlement located several kilometers
from the power station is accomplished at the BAEPP along with the production of electrical energy.
Results of Operation of the BAEPP. The experience of extended operation of industrial reactors with nu-
clear superheating of steam is unique, and the data accumulated during their operation are the basis for crea-
tion of the next generation of reactors. Let us note the main results of the operation of the BA.EPP,reactors-.
Replaceable SC are used in the BAEPP in whose fuel elements, having stainless-steel jackets, uranium dioxide
is used, which is enriched up to 5.0-6.5%o in uranium and dispersed in a heat-conducting matrix alloy. The allowable
temperature of the fuel element jackets is 630-650?C, which provides for superheating of steam up to 565'C. in
the channels.
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Fig. 5. Production (-) and cost (-?-) of
electrical energy in the second unit.
From the moment it was started up right up to the present more than 700 SC have been operated in the
reactors. The average energy production of the unloaded SC is ~26 MW ? days/kg, and their useful life in the
reactor is 5-6 years. However, the characteristics cited are not the limit. A group of channels is operating
with an energy production of ^'35 MW ? days/kg, which it has been decided to bring up to 37-40 MW ? days/kg in
them.
During the operation 30 SC were extracted prematurely from reactor 1 for monitoring inspections and
checks by virtue of putting the channels out of service and for other reasons, and only eight channels were ex-
tracted from reactor 2 in connection with disturbances in the operating conditions or for monitoring inspections
(Fig. 3). During the entire period of operation of the superheating fuel elements no case of their being put out
of service due to radiation impairment and incompatibility of materials was observed [5].
Thanks to the high reliability of the channels, the physical characteristics of the BAEPP reactors (ura-
nium enrichment, reactivity) provide fora satisfactory amount of the fuel component in the cost of the electri-
cal energy, notwithstanding the significant unproductive neutron absorption in the SC.
Evaluation of the fuel component in the cost of electrical energy permits confirming that at an average
depletion of 34 MW ? days/kg and with maintenance of the existing technology and the cost of preparation of the
fuel elements and channels one can expect values of the fuel component of R0.3 kopecks/kWh, which makes
nuclear superheating competitive in regions with a price level of 20-22 rubles/ton for organic fuel [6].
The operating experience with the Beloyarskaya reactors confirmed a rather stable equalization of the
energy distribution. A reduction in the capacity of maximally loaded channels and a practically constant ratio
of the total capacities of the evaporative and superheating circuits, as well as a negligible scatter in the tem-
perature of the steam at the output from the SC (Fig. 4), are a consequence of this. Regulation of the tempera-
ture of the superheated steam, the average over the reactor and at the output from individual SC, presents no
complication. The temperature of the steam at the output from the channels is stable in time, and its oscilla-
tions are negligible (2-3?C). Fluctuations on the ratio of power for evaporation and superheating of steam did
not exceed 1%. When necessary, e.g., during start-up, one can vary this ratio by altering the radial energy
distribution with the regulating rods.
The designers of the BAEPP reactors strived for the minimum effects possible of reactivity associated
with variation in the operating conditions of the AEPP, in particular, a variation in the amount of water in the
active zone under different operating conditions of the units, especially when starting and stopping them. The
operation of both reactors of the BAEPP has confirmed their weak sensitivity to the amount of water in the
zone. The greatest effect of reactivity in the BAEPP reactors is connected with emptying or filling the SC
with water during the start-ups and shutdowns of the units. This effect changes significantly during the operat-
ing process, which is explained, e.g., by its dependence on fuel depletion; however, it does not exceed 0.41/o in
absolute magnitude. The variation of the reactivity during the start-up of the reactor is easily compensated
by the regulation system.
Operational experiments with regard to the physics of channel reactors with nuclear superheating has
shown that nuclear-physical characteristics can be selected in this type of reactor which completely satisfy
both the nuclear safety requirements and the specific heat engineering requirements for nuclear superheating
while simultaneously providing for an acceptable amount of the fuel component, notwithstanding the use of steel
in the fuel channels and the additional neutron loss in the SC.
The production and cost of electrical energy during 1971-1975 in the second unit of the BAEPP are shown
in Fig. 5.
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TABLE 1. Principal Characteristics of Reac-
tors with Nuclear Superheating of Steam
Character- BAEPP
istic 1
BAEPP
2
Supercritical
parameters of
steam
Electrical
200
capacity,
Thermal
1820
pacity,
Fuel charge,
67
50
59,8
80
4
293
2
tons
Av. deple-
13,7/23
13,7/23
33/33
,
34/38
,
.
19/19
tion (EC/
SC), MW
days/kg
Enrichment
of urani-
u.m,
No. of evap-
389
429
orative .
than. (EC);
pieces
No. of su-
1304
1264
peteating
channels,
pieces
Steam temp.
540/540
540/540 *
450
before the
Crbine,
Steam pres-
240
240
sure before
the turbine,
kgf/cm2
*Turbine with intermediate superheating of steam.
The design of the fuel channels provides for an appreciable reserve with regard to the number of per-
missible heat-exchange cycles in the channel during the operating period of the fuel with rapid load variation.
The number of such cycles during 6 years is about 200, and the actual maximum rate of change in the steam
temperature was 20-40 deg C/min and in the pressure, X0.7 kgf/cm2 in 1 min. The reliability of operation of
the basic equipment is characterized by the readiness coefficient of the main circulating pumps (0.997-0.999)
and the feed pumps (0.993-0.995) [7].
The radiation environment of the AEPP site, and in particular next to the turbine during operation and
in connection with the maintenance of the process equipment during shutdown of the units, does not prevent
carrying out the maintenance operations. The deposition of radioactive corrosion products on the inner sur-
faces of the turbine are negligible. The radiation intensity at the high-pressure cylinder is 1.0-10 AR/sec and
at the low-pressure cylinder 0.2-8.0 ?R/sec. The strength of the radiation doses is 0.05-0.10 -?R/sec in con-
tinuously occupied places, 0.3-12.0 ?R/sec in places occupied part of the. time, 15-20 tsR/sec next to the equip-
ment of the superheating circuit of the first unit, and 5-50 pR/sec near the equipment of the condenser-feed
line of the second unit [8]. The ejection of radioactive products into the atmosphere under conditions of nor-
mal operation is less by a factor of 5-10 than the permissible health standards [9].
Prospects for the Development of Nuclear Superheating of Steam. A water-graphite reactor with nuclear
superheating to supercritical parameters of steam can be used for the indicated purposes under conditions of
the increasing need for energy systems in subpeak energy units and of the need for operation of AE PP accord-
ing to a dispatcher load diagram. Design studies of such a reactor are being conducted in the USSR. Accord-
ing to the expenditures cited, a specialist atomic unit will be competitive with similar units using organic fuel
for a comparable power of 800 MW in the utilization range of installed capacity of 3500-5000 h/yr [2].
The existing tendency of enlarging the individual capacities of reactors and turbogenerators makes the
combining of nuclear superheating with the application of low-absorbing construction materials continually
more urgent. The design of the RBMKP-2400 reactor, in which the superheating of steam to 450?C at a pres-
sure of 65 kgf/cm2 is provided [10], is promising in this direction; the zirconium alloys already mastered in
reactor technology are being used, and stainless steel will be used only for the casings of fuel element rods
made of uranium dioxide [11].
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Prototypes of the superheating channels of the RBMKP-2400 reactor are presently undergoing resource
tests on the BAE PP. Improvement of the engineering-economic indices of nuclear superheating is expected in
the RBMKP-2400 reactor due to an increase in the specific power of the fuel, the use of more favorable con-
struction materials for the active zone in the neutron physics respect, the application of nonremovable-chan-
nel design, etc.
The principle of sectional-modular preparation has been realized in the_ RBMKP reactor, which .improves
the engineering-economic indices of the AEPP, simplifies operations of bringing about a reactor on-line, and
permits regulation of the temperature of the superheated steam with the help of a system for controlling and -
regulating the energy distribution. The reactor is discussed in more detail in [10, 11]. The principal techni-
cal characteristics of reactors with nuclear superheating of steam are given in Table 1.
CONCLUSIONS
The operating experience of the BAEPP reactors has confirmed the possibility of the industrial realization
of nuclear superheating of steam right up to 510-540?C, sufficient reliability, and the safety of reactors of this
class.
The introduction of nuclear superheating is economically justified: when steam is superheated to 500?C
and higher with the use of stainless steels as the construction material in the active zone and the use of re-
movable and sleeve fuel elements; when zirconium alloys are used in the active zone and the steam tempera-
ture is ^-450?C, and when rod fuel elements, nonremovable channels, and the appropriate organization of steam
in the channel is used.
Reactors with nuclear superheating of steam permit operation under variable conditions and at atomic
heat and electric power plants with channeling of the heat to domestic and industrial needs.
Channel-type reactors with nuclear superheating permit enlarging capacity on the basis of standard units
and the use of high-speed turbine units having large capacity, and they significantly reduce the thermal emis-
sions into the environment.
LITERATURE CITED
1.
I. D. Morokhov et al. (editors), To Atomic Power of the 20th Century [in Russian], Atomizdat, Moscow
(1974).
2.
3.
P. I. Aleshenkov et al., in: Operating Experience of AEPP and Ways to Further Develop Atomic Power
[in Russian], Vol. 2, Izd. FEI, Obninsk (1974), p. 99.
I. K. Emel'yanov et al., At. Energ., 33, No.,3, 729 (1972).
4.
N.
A. Dollezhal' et al., At. Energ., 36, No. 6, 432 (1974).
5.
6.
A.
N.
G. Samoilov, A. V. Pozdnyakova, and V. S. Volkov, At. Energ., 40, No.5, 371 (1976).
A. Dollezhal' et al., in: Operating Experience of AEPP and Ways to Further Develop Atomic Power
[in Russian], Vol. 1, Izd. FEI, Obninsk (1974), p. 149.
7.
I.
Ya.
Emel'yanov, B. B. Baturov, and A. I. Klemin, ibid., p. 33.
8.
A.
P.
Veselkin et al., At. Energ., 30, No. 2, 144 (1971).
9.
A..
M.
Petros'yants, Atomic Power [in Russian], Nauka, Moscow (1976).
10.
A.
P.
Aleksandrov, Lecture at the International Atomic Energy Agency International Conference on
Nuclear Power and Its Fuel Cycle, Salzburg, May 2-3,1977, IAEA-CN-36/586.
11.
N. A. Dollezhal' et al.,. in: Operating Experience of AEPP and Ways to Further Develop Atomic Power
[in Russian], Vol. 1, Izd. FEI, Obninsk (1974), p. 233.
137
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 Declassified and Approved For Release 2013/03/07: CIA-RDP10-02196R000700110002-4
THE PR,INCIPA.L TECHNICAL PROBLEMS AND
PROSPECTS FOR THE CREATION OF
GA.S-COOLED FAST REACTORS WITH A.
POWER O.F 1200-1500 MW USING A.
DISSOCIATING COOLANT
A.
K.
Krasin, V. -B. Nesterenko,
B.
E.
Tverkovkin, V. F. Zelenskii,
V.
A.
Naumov, V. P. `Gol'tsev,
S.
D.
Kova.lev, and L. I. Kolykhan
Preliminary engineering-economic characteristics of atomic electric power plants (AEPP) with a fast
reactor of 1200-1500-MW electrical capacity were determined on the basis of neutron physics, thermal hy-
draulic, and engineering calculations and design studies of the reactor and the main equipment of an AEPP in,
which dissociating nitrogen tetroxide (N204) is used as the coolant (BRGD-1200-1500).
The advantages of such an AEPP are a decrease in the amount.of equipment due to the use of a single-
circuit layout of heat conversion and a reduction in the metal content.of the equipment by virtue of peculiarities
of the thermophysical properties of N2O4, as well as high yield rates of secondary nuclear fuel. This permits
one to predict the attainment of specific investments in AEPP of the BRGD-1200-1500 type up to the level of
investments in AEPP with water coolant.
Reactors of 1000-MW electrical capacity based on N2O4 can, according to the computational data, yield
up to 500-900 kg/yr of plutonium. These same reactors permit yields of up to 1400 kg/yr when operated as re-
processors [1].
A large number of alternatives wereconsideredinthe course of the design studies of fast reactors based
on N204, and they differ among themselves in the gas exit temperature of 2800-570?C, the pressure in the circuit
of 80-160 bars, the construction of the fuel elements (rod and spherical), and the type of fuel composition (ma-
trix fuel based on uranium dioxide and nitrides in Nichrome or chrome matrices with 30-4010 by volume [2];
low-alloy metallic fuel with double protection from possible interactions of N204 with the fuel; and carbide fuel
[3] with a carbon-silicon casing for spherical microfuel elements).
All the alternatives discussed essentially satisfy contemporary requirements on the yield rate of secon-
dary nuclear fuel. Investigations of the fuel cycles of the growing nuclear power show that the consumption of
natural uranium in a nuclear power system can be reduced by 45-5010 upon the introduction (in 5 years) of
fast reactors based on Na and N204 in comparison with thermal and fast reactors based on Na (the external
cycle time is T = 0.5-1 year).. In addition, in ^r 30 years the system under discussion will develop into the
mode of providing its own plutonium [8].
The principal thermal hydraulic and physical characteristics of breeder reactors and reprocessors of
the BRGD type with a matrix fuel based on uranium dioxide and plutonium in the active zone (1500-MW electri-
cal capacity) are given in Table 1. At a gas exit temperature from the reactor of 450?C and a pressure of 150
bars, amaximum temperature of the fuel-element casings of 650-680?C, and with heating of the gas in the reac-
tor to 230-270?C one can achieve a heat release rate of 800-1000 kW/liter of the active zone, having obtained
a doubling time of 5-6 years with a plutonium yield of 500-900 kgf/yr for breeder reactors and up to 1400 kg/yr
for reprocessors.
The best characteristics of the BRGD-type reactor are produced by: the high energy release rate; the
rigid spectrum of the neutrons (especially in the case of the use of a chrome matrix) (a.9O =0 - 248 and (a)N211
z a
a =
0.260; and the large contribution of the shields to-the reproduction of fuel by virtue of the high leakage of neu-
trons from the active zone and the use of metallic uranium in the shields as the source material.
Translated from Atomnaya Energiya, Vol. 44, No. 2, pp. 131-136, February, 1978.
138 0038-531X/78/4402- 0138 $07.50 ?1978 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/03/07: CIA-RDP10-02196R000700110002-4
150 200 250 300 350 400 450 500 1?C
Fig. 1. Experimental data on the specific heat of N204, kcal/(kg ? deg): A) 116; A) 130;
0) 150; ?) 170.
Alternatives with matrix fuel based on uranium dioxide were also considered along with the use of metal-
lic uranium in the shields. The possibility of increasing the diameter of the fuel elements in the shields in
comparison with metallic uranium serves as a prerequisite for this by virtue of the relatively high thermal
conductivity and the higher limiting temperature of the matrix fuel.
Investigations have shown that a significant advantage is maintained in the doubling time of a reactor with
metallic uranium in the shields when the matrix content in the fuel of the shields.is > 401/o. When the matrix
content is - 300, the doubling time of reactors with matrix and metallic fuel in the shields (with optimal shield
thickness) is approximately identical.
Double protection of the fuel is advisable for increasing the reliability, excluding contact of the coolant
with the fuel, and reducing the outflow of gaseous fission fragments. The use of matrix-type fuel [2], which
consists of particles of uranium dioxide and plutonium covered by metal and contained in a casing made of
stainless steel, is explained by this increase in reliability. The role of backup protection is filled by the
metallic matrix with fuel particles dispersed in it.
The use of matrix fuel reduces, on the one hand, the amount of uranium dioxide and plutonium and in-
creases the contribution of the construction materials, which increases the parasitic capture of neutrons. On
the other hand, it is possible to raise the energy release rate of the active zone by a factor of 1.5-2 in com-
parison with pure uranium dioxide due to the better thermal conductivity of the matrix fuel (higher by approxi-
mately a factor of 5-6 than that of pure uranium dioxide). This circumstance allows the creation of a more
compact active zone and production of satisfactory physical characteristics. In addition, the matrix material,
playing the role of an additional casing, reduces the outflow of gaseous fragments into the circuit.
Microfuel elements based on UO2 and UC2 with a coating of pyrocarbon, silicon carbide, and chrome are
proving to be very promising for application. Such coverings not only protect the fuel from the corrosive ef-
fect of the coolant but also retain the radioactive fragments very effectively [3].
Another distinctive feature of reactors based on N204 is the comparatively low level of the coolant tem-
perature, which offers the prospect of using metallic uranium in the reactor shields.
The dissociating nitrogen tetroxide proposed for use as a coolant and working medium has the following
interesting characteristics.
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TABLE 1. Calculated Thermal Hydraulic and Physical Characteristics of Breeder
Reactors and Reprocessors Operating on N204 -
Rod fueel ele-
ments
Reactor characteristic
Thermal capacity of the reactor, MW
Temp. (exit/ entrance). 'C
Gas pressure (exit/ entrance), bars
Gas consump., kg/sec -
Fuel in the active zone
Equivalent diam., m
Height, m
Vol:,-liters
Av. heat release rate of active zone, kW/liter
Cassette size, mm
Triangular lattice step; mm
Fuel density, g/cros
Fuel element diam.. mm -
Material of fuel element casing
Casing thickness, mm
Total critical charge of .. ,9 235
Pu and U, kg.
.
Av. enrichment, To
Reproduction coeff. of active zone
Fuel of shields
Yield of excess plutonium per year
Doubling time (years) -when T -0.5 year
Total reproduction coeff.
*For breeder reactors.
tFor reprocessor reactors.
30% I 40%
5100
451/183
154/169
5670
2,67
1,02
5,3
Reactors with mi-
crofuel elements'
pressure, bars'
150 I 80
5100:
451/183
150/80
5670
Rod fuel elementst
chrome matrix
30% 40%
5100 5910
451/183
154/169'. 84/99
5670 7260
U02 + matrix,
2,67 2,53
1,02 0,98
5,3 4,94
864 926
92X2
94
10
6
2,53
0,98
4,94
926
864
92X2
94
10
6
1,8
.4,6
1000
1,92:
5,62
0,4
1900
12,8
0,90
690
4,6
1,51
1780
14,5
0,80:
620
5,2
1,40
172X2
187
10
6
Stainless steel
0,4
1537 I 1794
15,62 15,43
0,73 0,75
Metallic uranium
860 940
4,0 4,5.
1,62 1,58
TABLE 2. Values of the Mean Integral
Specific Heats of Various Coolants, kcal/
kg deg
150-280
200-500
1
x204 I C02
1,5-1,6
0,7-0,8
0,28
0,275
0,31
0,306
1,243
1,243
0,4
2600 2460
17,6 20
0,45 0,38
1400 1380
0,92 0,88
1. The significant size of the thermal effect of the chemical reactions of dissociation upon heating and
recombination upon cooling, N204 2NO2 (-149 kcal/kg) 2N0 + 02 (-293 kca1A g), permits organizing in-
tense heat removal in the active zone of the reactor and heat exchange equipment.
2. The values of the specific heat are high over a wide range of temperatures and pressures. Experi-
mental data on the specific heat- [4] are given in Fig. 1. The comparison of the mean integral specific heat
(Table 2) was carried out on an isobar at the identical distance from the critical one (it > P/Pcrit = 1.1-1.2).
3. Obtaining heat-transfer coefficients which are enhanced in comparison with the inert gases [5, 6] due
to concentration diffusion; thus, average heat-transfer coefficients of ^90,000 kcal/(m2 - h ? deg) are obtained in
tests for the conditions of the active zone of the BR?-1500 reactor with respect to temperature, pressure,
and thermal flux. This fact permits creating simultaneously a more compact and lower - metal - content heat
exchange device.
4. The relatively small amount (^'100 kcal/kg) of hidden heat of vaporization and the parameters on the
saturation line permit realizing a subcritical gas-liquid cycle in which complete evaporation and superheating
of the coolant occur due to regeneration; the reactor is purely gas-cooled in this case.
5. On account of the lower expansion ratio in the cycle (for equal initial parameters and identical cooling
conditions of the terminal heat exchange) a turbine operating in N204 will have first stages significantly larger,
and for a density of N204 behind the turbine 30-40 times larger than that of water vapor, the last stages will be
2-3 times smaller than for a turbine (of same capacity) operating on water vapor. This exerts a favorable
effect on the efficiency, and the total number of stages is reduced by a factor of 4-5, which decreases the
overall size and metal content of the turbine.
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 Declassified and Approved For Release 2013/03/07: CIA-RDP10-02196R000700110002-4
TABLE 3. Main Characteristics of Experimental-Industrial AEPP
Electrical capacity, MW
300
300
300
Thermal capacity, MW
955
955
955
Cycle
Gas-liquid
No. of circuits, pieces
1
No. of reactor cooling loops, pieces
3
No. of emergency cooling loops, pieces
2
Main coolant parameters :
flow rate through reactor,
/sec
1064
1064
1064
1
temp. at entrance-exit of eactor, ?C
196/480
196/480
196/480
pressure at entrance-exit of reactor, bars
165/152
165/152
165/152
temp. at entrance-exit of condenser, ?C
60/31
60/31
60/31
ressure at entrance-exit of condenser, bars.
2,25/2
2,25/2
2,25/2
Net eff. of AEPP,
31,4
31,4
31,4
Fuel comp. in active zone
U02+30%Cr
U02+Pu02+30%Cr
Diam., m
1,3444
1,3444
1,3444
Height. m
0,74
0,74
0,74
Vol., liters
1050
1050
1050
Diam. of fuel elements x casing thickness. mm
6,2X0,4
6,2x0,4
6,2x0,4
v. energy release rate (max.), kW/liter
4
835/1250
800/1225
825/1220
sU-239Pu charge, tons
0,935
0,591
574
0
Size of cassette under key?, mm
142
142
,
142
Ooerating period, effective years*
0616
0,602
0,602
Shield material
Umet
Umet
U02+30%Cr
Total repro. coeff.
1,020
1,644
1,466
Amt. of excess plutonium unloaded from the reactor,
381
210
160
kg/eff ? yr.
Doubling time for Pu content (1%) in the unloaded uranium of
the reproduction zone and T = 0.5 year
*Net operating time at nominal capacity.
T, ?0
500
45 47 0,9 1,1 1,3
Op, kcal/kg- deg
Fig. 2. Basic thermal layout and cycle
of the BRGD-1500 AEPP: I) reactor; II)
high-pressure turbine; III) regenerator;
IV) low-pressure turbine;. V) condenser;
VI) booster pump; VII) feed pump; VIII)
"dry" water-cooling tower.
6. Obtaining high thermal efficiencies of gas-liquid cycles over a wide range of temperatures and pres-
sures.
7. The presence of resistant construction materials in an N204 medium.
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8. The rather high radiation resistance of the coolant. Thus the radiation losses for the BRGD-1500
AEPP amount to 10-5-10-6 in the case of a gas temperature at the reactor exit of 450?C and a pressure of 150
bars. Irreversible decomposition is higher by approximately a factor of 1.5 and 2 for H2O and CO2 under com-
parable conditions. The thermal decomposition of N204 is negligibly small for these parameters.
9. The intrinsic activity of the N204 coolant is rather low (5. 10-5 Ci/g for N204 based on 16N, and 10-$ Ci/
g for that based on 17N).
The corrosion activity amounts to 2.10-7 Ci/g when a decontamination system with a coolant flow rate
of 1% the total flow rate is present. The activity of the sodium circuit amounts to 50 Ci/liter as a. rule for 24Na.
A. single-circuit thermal layout with heat regeneration at an intermediate pressure was selected for the
BRGD-1500 AEPP [7]. The thermal layout includes three reactor cooling loops with turbines of 500 MW each.
The main equipment of the AEPP also contains heat exchange devices - regenerators, a condenser, and feed
pumps.
The basic thermal layout of 'an AEPP with the cycle is presented in Fig. 2. The coolant i's -compressed
in the liquid phase in the booster VI and feed VII pumps up to the maximum supercritical pressure in the cycle;
then it is superheated in the regenerator III to the gaseous state due to the heat of the gas entering from the
high-pressure turbine II. The coolant is heated up to the maximum temperature of the cycle in the reactor I,
from which the gas enters the high-pressure turbine; the final expansion of the gastakes place after theregener-
ator III in the low-pressure turbine IV. The cycle is completed in the condenser V, where the coolant is
cooled to the minimum parameters of the cycle. Taking account of the coolant parameters on the saturation
line, it is possible to use effectively "dry" air water-cooling towers of the Heller type VIII in the condenser
(pressure - 2 bars, and entrance- exit temperature is equal to 60-31?C for N2O4).
Engineering-economic calculations have shown that effective plutonium yield and electrical energy produc-
tion at acceptable cost can be accomplished in two economically equivalent optimal areas of maximum gas tem-
perature behind a 430-480?C reactor for an oxide fuel and behind a 250-320?C one for a metallic fuel.
The main characteristics of the BRGD-1500 AEPP are given below for characteristic regions of the
parameters.
Thermal capacity, MW .............................. 5100
Coolant flow rate, kg/sec ... .. . ......... . ......... 5670
No. of cooling loops, pieces ......................... 3
No. of emergency cooling loops, pieces ........ .... ...... 2
Parameters of coolant prior to the turbine:
temp., ?C .... ... ........................ 450
pressure, bars .... .......................... 154
Parameters of coolant at the condenser entrance:
temp., 'C ................................... 66
pressure, bars ............................. 2.36
Parameters of coolant prior to the pump:
temp., ?C ..... 31
pressure, bars............ .................. 2.02
Net eff., % ................. 34
Cycle .......................................... Gas-liquid
The theoretical possibility of creating an AEPP at the exit from a reactor with a subcritical pressure of
80-90 bars inthe circuit has also been evaluated at the Nuclear Power Institute of the Beloyarskaya SSR A cad-
emy of Sciences. In this case the physical characteristics and thermal-hydraulic characteristics of the equip-
ment deteriorated. .
Nitrogen tetroxide is a. product of the chemical industry (it is used in the production of nitric acid and be-
longs to the group of nonexplosive noncombustible materials).
The operating experience accumulated in the chemical industry in connection with the production of N204
and the extended operation of semi-industrial test stands at the Nuclear Power Institute of the Beloyarskaya
SSR Academy of Sciences (from 1965) permit drawing a conclusion as to the mastering of the coolant and the
advisability of proceeding to larger-scale tests.
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A large complex of experimental-industrial test stands has been created at the Nuclear Power Institute
of the Beloyarskaya SSR Academy of Sciences on which tests are conducted of mock-up models of heat-ex-
change devices, scram system units, the turbine, pumps, and engineering instrumentation, ,materials study
research is conducted, etc.
Countries which are members of the Council for Mutual Economic Aid are participating extensively in
a program within the framework of the Programof Scientific-Engineering Cooperation of scientific-research
work on mastering the dissociating coolant. Thus, loop and ampoule apparatus for materials study tests of
construction and fuel materials.are being produced at the Nuclear Research Institute of the Polish People's
Republic, and a high-pressure apparatus for conducting thermal-hydraulic tests of mock-up equipment and
engineering research is being developed in Hungary at the Budapest Technical University.
The concepts, scientific-engineering data, and basic design characteristics of the BRGD-1200- 1500 AEPP
outlined above show that the proposed AEPP satisfies contemporary requirements with respect to its engineer-
ing and economic characteristics, and the distinctive features of the engineering layout and the thermodynamic
cycle (gas-liquid) permit reliably guaranteeing the safe operation of the AEPP. The program of theoretical
and experimental research being conducted into the basis of the design and the mastering of the technology of
the coolant operation in the required range of parameters provide every reason for posing the problem of
creating an experimental-industrial AEPP based on N2O4.
The main characteristics worked out for an experimental-industrial AEPP with a fast reactor operating
on N204 are given in Table 3.
The plant is intended for finishing up the research on a gas-cooled fast reactor operating on N204 and the
study of a single-circuit layout for heat conversion, as well as for finishing up all the engineering solutions
associated with the specific properties of the coolant being used.
The thermal capacity of 1000 MW was. selected from considerations of the possibility of using this reac-
tor after its completion and introduction as an independent power unit capable of providing for the production
of low-potential heat and electrical energy together with the production of secondary nuclear fuel.
LITERATURE CITED
1. A. K. Krasin et al., in: Dissociating Gases as Coolants and Working Media of Power Plants [in Russian],
Part I, A. K. Krasin et al. (editors), Inst. Teplo- iMassoobmena Akad. Nauk BSSR, Minsk (1967),
p. 25.
2. H. Bumm et al., Proceedings of the International Conference, "Fast Reactor Fuel and Fuel Elements,"
Karlsruhe, Sept. 1970.
3.
4.
5.
6.
7.
I.
Sayers, in: Proceedings of the IAEA: "Gas-cooled fast reactors," Minsk, July 1972.
A.
V.
K.
B.
Krasin et al., op. cit., p. 113.
Nesterenko et al., Teploenergetika, 11, 72
(1974).
A.
K.
Krasin et al., op. cit., p.42.
V.
P.
Bubnov and V. B. Nesterenko, Schemes for Conversion of the Heat of an AEPP Operating on
Dissociating Gases [in Russian], Nauka i Tekhnika, Minsk (1975).
8. V. B. Nesterenko and V. Ya. Tsirikhova, in: Dissociating Gases as Coolants and Working Media of Power
Plants [in Russian], Part III, A. K. Krasin et al. (editors), Izd. Inst. Teplo- i Masoobmena Akad. Nauk
BSSR, Minsk (1976), p. 25.
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PHYSICOTECHNICAL ASPECTS OF NUCLEAR AND
CHEMICAL SAFETY OF POWER PLANTS WITH
GA.S-COOLED FAST N204 REACTORS
V. B. Nesterenko, G. A. Sharovarov,
S. D. Kovalev, and V. P. Trubnikov
The provision of safety is one of the most controversial problems in the design of new atomic power
plants, especially of those based on fast reactors.
If atomic power plants are to be competitive with conventional plants with respect to safety, the prob-
ability of occurrence of serious emergency situations in them must be at least as low as in conventional plants
at any given location. The fact that atomic power plants possess better thermoeconomic characteristics than
thermal plants is presently not sufficient. Thus, all atomic plants currently being designed or constructed
must be provided notionly with conventional safety devices used in thermal plants burning organic fuel but also
with special devices and systems that ensure security in case of specific accidents involving the danger of reac-
tor core melting or discharge of the coolant into the environment.
By using dissociating nitrogen tetroxide (N2O4) as a coolant and process. medium, it i& possible to design
high-power atomic plants with fast reactors having, as indicated by preliminary studies, promising technico-
economic characteristics [1-3].
The main advantage of such plants is the possibility of using a single-loop heat-conversion circuit based
on the gas-liquid cycle [4]. The heat circuit includes three reactor cooling loops with turbines of 500 MW each
[1]. As follows from the description of the heat circuit of such a power plant, its reactor is purely gas cooled.
It is well known that the use of gaseous media for core cooling creates certain safety problems. These
problems are associated with the high thermal stresses existing in the reactor core, the low storage capacity
of the coolant, and the correspondingly fast rate of development of emergency situations.
The basic condition for safe operation is thus the assurance of reliable circulation of the cooling agent
in emergency conditions and during reactor cooldown. The most dangerous emergency situations include the
development of leaks at various points of the technological circuit, power cutoff in the entire plant and to the
supply pumps, accidental changes of reactivity, etc.
Various possible emergency situations have been analyzed in course .of the design of a high-power atomic
plant (BRGD-1500) and an experimental industrial 300-MW electric power plant (BRIG-300).
One of the most dangerous failures in multiple-loop atomic power plants with gas or liquid metal coolants
is power cutoff either to the entire plant or to the supply pumps and compressors, since in such cases it be-
comes very difficult to secure continuous circulation of the coolant. Such a continuity can be prolonged for a
certain time by means of pressure equalization, provided there is a certain amount of coolant on'the high-
pressure side.
The use of a gas-liquid thermodynamic cycle has an advantage over purely gas or liquid metal systems
because of the large amount of coolant in the loop in proportion to the flow rate, and because of the great dif-
ference between maximum and minimum pressure in the loop. With a constant flow rate in the technological
circuit, the ratio of the coolant mass M to the flow rate is given by the factor Km=(M/G) dl/W, equal to the.
Jo
time for which the section 1 can maintain the rated coolant flow rate. Depending on the construction and cycle
parameters, this time can vary between 20 and 60 sec from section to section, whereas in purely gas cooled
circuits using, e.g., helium, this time does not exceed 2-7 sec [5].
Translated from Atomnaya Energiya,, Vol. 44, No. 2, pp. 137-140, February, 1978.
144 0038-531X/78/4402- 0144 $07.50 ?1978 Plenum Publishing Corporation
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 Declassified and Approved For Release 2013/03/07: CIA-RDP10-02196R000700110002-4
P
0,8
48
44
0
Fig. 1. Variation of flow rate, pressure, and temperature of coolant
at the reactor inlet in case of power cutoff to the main circulation pumps
and the entire power plant..
Fig. 2. Variation of coolant pressure and flow rate at the reactor inlet
in case a break in the main steam line behind the turbine.
The difference between maximum and minimum pressure in the loop determines the driving forces which
cause natural coolant flow. For an atomic power plant with a fast reactor and a gas -liquid cycle using a dis-
sociating coolant, the pressure increase factor amounts to about 70, and the pressure difference to 160 bars.
This figure is considerably lower for other gas coolants. These facts offer good conditions for unaided flow
of large amounts of coolant in case of power failure and shutdown of the main supply pumps [6]. For example,
Fig. 1 shows the changes in coolant flow rate G, pressure P, and temperature T at the inlet to the reactor in
case of power failure. It has been assumed that no pump coasting takes place (worst case conditions). Figure
1 indicates that a considerable coolant flow is maintained for a time sufficient to connect emergency power
sources (the fuel-element temperature did not exceed the maximum permissible value for nearly 5 sec).
Studies of the loss of containment indicate that a break in the main steam line behind the turbine results
in a gradual fall of coolant flow rate from nearly its nominal value (Fig. 2). The rate of fall does not exceed
2.2 bars/sec,theflowrate decreasing to about 5d'loof its nominal value after 20 sec.
Figure 3 shows the variation of coolant parameters at the reactor inlet in case of a, break in the main
steam line between the reactor and turbine. The initial rapid rise of the flow rate is followed by a fast drop
to about 35% of the nominal flow rate after 20 sec; the temperature of fuel and of the fuel-element jackets ini-
tially decreases. In this case it is desirable to provide flow rate restrictors with a diameter one-half as great
as the diameter of the break in order to prevent deformation and damage of the structural elements of the reac-
tor.
The most dangerous emergency situation involves the loss of containment in the main loop at the reactor
inlet which causes an initial drastic drop of coolant flow through the reactor and reverse circulation. Circula-
tion reversal is practically instantaneous, so that the fluctuations of temperature and pressure taking place in
a very short time interval are of great importance. Circulation reversal can be prevented by increasing the
number reactor feed lines so that a break in any one line does not cause reversal and an intolarable drop of
circulation.
Thus, due to the accumulation of coolant in the reactor circuit, power failure or loss of containment in
the main steam lines do not immediately interrupt circulation which continues for a time sufficiently long for
restoring power or for shutting down the reactor and starting the cooling procedure.
Of great importance in operating safety are the self-regulation properties of the reactor. A comparison
with a sodium-cooled reactor proved that to cause the same power "overshoot," an N2O4 reactor needs a reac-
tivity twice as high. The effect of density in large reactors is negligible within the operating range. Even with
total removal of coolant from the core and breakdown of the power control system, the available protective
devices are capable of shutting down the reactor without increasing the power above the permissible level.
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Fig. 3. Variation of coolant pressure
and flow rate at the reactor inlet in
case of a break in the steam line at the
reactor outlet with (---) and without
(-) flow restriction.
To improve safety, atomic power plants using N204 cooling are provided with the following additional
measures: duplication of supply pumps, several reactor cooling loops (at least three), turbine bypassing in
case of generator power failure, the use of two emergency cooling systems based on two independent and dif-
ferent operating principles.
To improve radiation safety, it has been decided to use double protection of the fuel from interacting
with N204. The matrix fuel consists of uranium dioxide or plutonium particles covered by metal and placed
in a stainless-steel jacket. The operation of fuel compositions in N204 media will be studied in Poland in the
loop system of the "Mariya" reactor.
The use of chemically active N204 as a coolant and active medium of an atomic power plant, as proposed
by the Institute of Nuclear Power Engineering of the Academy of Sciences of the Belorussian SSR,-places addi-
tional demands on safety associated with the fact that any discharge of coolant can cause contamination of the
surrounding space with nitrogen oxides.
Nitrogen tetroxide is toxic and strict norms are imposed on the amount of nitrogen oxides and of the
products of their chemical interaction released into air. Thus, their total amount must not exceed 5 mg/m3
in hot laboratories, and 0.085 mg/m3 in populated areas. Accordingly, atomic power plants using N204 as a
coolant and active medium must be provided, in addition to systems ensuring radiation safety, with systems
securing chemical safety in normal operation and in emergency situations.
Nitrogen tetroxide as a process medium is quite familiar in the chemical industry (in the production of
nitrogen fertilizers and other similar products). Experimental test stands operating for more than 10 years
at the Institute of Nuclear Power Engineering of the Belorussian Academy of Sciences indicate that there is a
practical possibility of designing constructions capable of withstanding pressure up to 160 bars and coolanttem-
peratures of 500-600?C, i.e., over the entire range needed in atomic power plants. Accidental leaks of N204
vapors from the plant can be reduced to a few kilograms per hour and subsequently removed by the ventilation
system. Chemical; devices for monitoring the presence of nitrogen oxides in air allow early detection of cool-
ant leakage. The points of leakage can be rapidly determined with the aid of special leak detectors as well as
by color (yellow) and odor.
Calculations indicate that a ventilation pipe 125 m high together with "dry" cooling tower (Heller tower)
can expel into the atmosphere more than one ton of coolant vapors per hour. Dispersal of such large dis-
charges is made possible by a current of warm air 15?C hotter than the environment ascending at a speed of
about 5 m/sec and having a diameter of 90 m at the top section of the tower.
Thus, with such a ventilation system the permissible discharge of coolant will be limited not by the sum
of purposeful and accidental leaks but the material balance of the coolant and the observation of proper radia-
tion norms.
The sources of prearranged leaks are blowoffs of gases from the main-loop condenser to remove the
incondensible products of radiation and thermal decay and from the end seals of the turbine shaft. The major
fraction of coolant vapors from these leaks is returned to the main circuit after separation and purification of
the accompanying gases.
To localize any significant discharge of coolant vapors the power plant canbe enclosed inside a sealed protec-
tive cover. The volume of the cover should be large enough so that the largest possible discharge of coolant
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vapors does not raise the pressure under the cover above 4 bars. For stilltighter localization, the pressure
is lowered by means of special technological equipment that condenses the coolant vapors and collects the con-
densate in special reservoirs. After aging and radiochemical decontamination (if necessary), the condensed
coolant is put back into circulation.
Penetration of the coolant to the environment through leaks in the condenser is prevented by using a "dry"
cooling tower with an intermediate water-cooling circuit. At the same time, provisions have been made to al-
low the power plant to operate with a certain. leakage from the condenser. The produced nitrous acid is re-
moved from the cooling water by ion exchange and from the coolant, by rectification.
The following conclusions can thus be drawn from the above discussion: the specific.properties of the
technological scheme and of the gas -liquid cycle of N204 allow continuous circulation and initial heat removal
when the protective system operates in case of emergencies; physicochemical properties of the dissociating
coolant ensure that in the cool-down procedure of a gas-cooled reactor the coolant will be in a liquid phase;
structural measures applied to fuel elements, control systems, cool-down procedures, power plant cooling,
outside cover including devices for localization and elimination of the consequences of accidental discharge
of coolant ensure safe operation of atomic power plants using N204.
LITERATURE CITED
1. A. K. Krasin et al., in: Experience of Operating Atomic Power Plants and Further Development of
Atomic Power Engineering [in Russian], Vol. 1, Obninsk (1974), p. 262.
2. V. B. Nesterenko, Physicotechnical Principles of the Application of Dissociating Gases as the Coolant
and Working Medium of Atomic Power Plants [in Russian], Nauka i Tekhnika, Minsk (1971).
3. In: Dissociating Gases as Coolants and Working Media of Atomic Power Plants [in Russian], Part 1,
Inst. Teplo- i Massoobmena, Akad. Nauk BSSR, Minsk (1976), p. 85.
4. V. P. Bubnov and V. B. Nesterenko, Heat Conversion Circuits of Atomic Power Plants Using Dissociat-
ing Gases [in Russian], Nauka i Tekhnika, Minsk (1975).
5. J. Brit. Nucl. Energy Soc., 8, 3 (1969).
6. G. A. Sharovarov et al., in: Dissociating Gases as Coolants and Working Media of Atomic Power Plants
[in Russian], Part 1, Inst. Teplo- i Massoobmena, Akad. Nauk BSSR, Minsk (1976), p. 72.
PHYSICAL PROPERTIES OF FAST POWER REACTOR
FUELS AND THEIR EFFECT ON THE FUEL CYCLE
0. D. Bakumenko, E. M. Ikhlov,
M. Ya. Kulakovskii,. B. G. Romashkin,
M. F. Troyanov, and A. G. Tsikunov
The article discusses the principal properties of spent fuel: isotopic composition, activity of fission
products and steel, and residual heat release.. Data are cited on the accumulation of transplutonium elements
in spent fuel and on the effect of these elements on activity and residual heat release.
The effect of higher plutonium isotopes on the natural activity of fuel and on the radiation environment
in handling plutonium fuel is analyzed. The necessary degree of decontamination of fuel of fission products
in chemical processing has been determined by analyzing the natural activity of fuel.
The effect of shortening the cooling time of spent fuel on the dynamic changes of fission products activity
and residual heat release is investigated. The effect of storage time of fuel between chemical reprocessing
and loading into the reactor on the magnitude of residual heat release and neutron activity of the unloaded fuel
is estimated.
Engineering problems arising in connection with the reduction of the cooling time of spent fuel in the fuel
cycle system are discussed on the basis of the obtained information.
Translated from Atomnaya Energiya, Vol. 44, No. 2, pp. 140-145, February, 1978.
0038-531X/78/4402-0147$07.50 ?1978 Plenum Publishing Corporation 147
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TABLE 1. Accumulation of Transpluto-
nium Elements as a Function of Storage
Time, kg/ton
, I
1 I
2
I
238pp
0,9
1
1,05
1,2
1,3
241Am
0;6
0
8
1
1.,4
1,8
242mAm
0,03
,
0,06
0,08
0,12
0,16
243Am
0,5
0,5
0,5
0,5 .
0,5
242Ku
0,11
0,18
0,24
0,37
0,49
2t3Ku
0,01
0,02
0,03
0,05
0,07
244Ku
0,1
0,1
0,1
0,1
0,1
TABLE 2. Total Activity of Fission
Products
Unit of ac- Cooling time, years
tiVity 0 I 0,25 I 0,5 1 1 2,5
Ci/ton
7,6.107
2,2.107
1,3.107
7,7.106
2,7.1081
(U+Pu)02
g-eq/ton
2.10
3,8.106
2,2.106
9.105
4,1.105
(U+Pu)02
Fast power reactors are called upon to provide the required space of development of future nuclear power
engineering. The solution of this problem depends in a considerable measure on fast turnover of spent fuel in
the external fuel cycle which involves cooling, transportation, and storage of fuel, chemical reprocessing,
preparation of new fuel elements, etc. The external fuel cycle time, together with the conversion ratio and
the burnup fraction, is one of the most important factors that determine the breeding rate of nuclear fuel in
fast reactors. For example, reduction of the external cycle from two years to one year has the same effect
on the rate of plutonium breeding as a twofold increase of allowable fuel burnup.
Most of the external fuel cycle time is occupied by cooling the fuel after its exposure in the reactor.
Various specialists give different figures for the optimal cooling time for fast reactor fuel. One of the main
reasons for such different estimates is the fact that the properties and specific features of fuel have not been
adequately studied. An important stage in such studies is a comprehensive experimental investigation of fuel
properties. This stage has not yet received sufficient attention since the experience gained in operation of
fast reactors is still insufficient. Nevertheless, quantitative analysis based on experimental nuclear-physics
data makes it possible to reveal the basic features of the external fuel cycle which are pertinent for this or
that stage of the cycle [5].
The characteristics of the BN-1500 reactor calculated on the basis of a fast sodium-cooled power reac-
tor of 1500 MW (electrical) are listed below.
Thermal power ...... ........ ....................
Fuel....... ............. ....... ...........
Pu enrichment ....... ...........................
4000 MW
(U + Pu)O2
14-18%
Duration of run ((p = 0.8). .................. ... .......
Average fuel burnup ... ..................... ..... .....
No. of fuel assemblies loaded (unloaded)
core. ............ ............... .........
side breeder blanket. ...... ..................
No. of elements unloaded from assemblies
480 days
70 kg/ton fuel
320/yr
.120/yr
core .............. .. .............. .........
90 , 00 0 /yr
side breeder blanket .............................
7000/yr
Mass of fuel in assemblies
core (U + Pu)O2 .................. .. .. ..........
70 kg
end breeder blanket (UO2) .........................
50 kg
side breeder blanket (U02) . .. .................. ..
190 kg
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TABLE 3. Activity of 95(Zr+Nb), 103Ru,
and 106Ru in Ci/ton (U + Pu)02
0,25 I 0,5 j 1 I
2,5
957r
2.106
8.105
1,1.105
2.102
3
95Nb
3,3.106
1,5.106
2,4.105
,
7,4.102
103Ru
2,6.106
5.105.
2.104
106Ru
2,6.106
2,2.106
1,6.106
6.105
Mass of steel in assemblies
core and end breeder blanket .......................... 60 kg
side breeder blanket................................. 40 kg
Amount of plutonium dioxide (isotopic mixture)
unloaded from:
core ......................................... 3.5 tons/yr
end and side breeder blankets ....................... 0.8 ton/yr
Amount of PU02 loaded per fuel cycle as a function of external fuel cycle time:
T, years..... ... 0
0.5
1
2
3
PU02, tons ........ 5
6.9
8.8
12.6
16.4
Isotopic Composition of Plutonium. Accumulation of
At the first stage of development, fast reactors are intended to use plutonium fuel generated in thermal
reactors. Plutonium obtained in thermal reactors under high burnup conditions [1, 2] has approximately the
following isotopic compositions: 10-5% 236Pu and 1, 58, 23, 14, and 4% of 238-242pu, respectively.
If such fuel were loaded into the BN-1500 reactor, plutonium unloaded from the reactor core would be
composed of the nearly same amount of 236pu, 239-240pu, and 242Pu and approximately one-half as much of 238pu
and 241pu
The computed isotopic composition of plutonium in fuel unloaded from the different enrichment regions
is: (5-6) . 10-6% 236pu, 0.7-0.5% 238Pu, 62-65% 239pu, 8-7 0 241Pu, and 6-5% 242pu.
The isotopic composition of the unloaded plutonium is close to the composition of equilibrium plutonium
generated in fast breeder reactors with multiple recirculation of fuel and plutonium make-up from the shield.
According to the author's estimates the equilibrium composition of such plutonium is 5 ? 10-6, 0.2, 62, 27, and
6% for 236P and 238-242Pu, respectively.
Transpiutonium elements produced in irradiation of plutonium fuel affect both the activity and residual
heat release of fuel. The content of transplutonium elements in plutonium fuel in the different enrichment re-
gions of the core is given below in kg per ton (U + Pu)O2:
2a;P
ii,^_
242Cm-0.13-0.1
241An1
0.9-0.4
243Cm - 0.012-0:006
242-Am - 0.05-0.03
244Cm.., 0.1
243Am 0.5
These data were obtained under the assumption that the original loaded fuel contains no 241A.m.
In addition, irradiation causes production of 7-day 237U with a final content of about 0.01kgperton(U+
Pu) 02.
The decay of 241Pu in storage after chemical reprocessing causes production of 241Am at the rate of 4.8%
of 241Pu content per year of storage, and of 237U which comes into equilibrium with 241Pu after 1 month in
storage.
The loss of nuclear fuel in decay of 241Pu and accumulation of 241Am is the greater the longer the external
fuel cycle. This and also the "parasitic" capture of 241Am neutrons should be accounted for in calculating the
rate of fuel increment or the doubling time.
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TABLE 4. Activity of Volatile and Gaseous
Fission Products in 'Ci/ton (U + PuO2
1291
1311
134CS
136C6
1370'6
85Kr
131Xe
131Xe
0,25 I 0,5 I 1
I 2,5
0,065
0,065
0,065
0,065
2;4;103
1,1
1;9.105
1,8:105
1,5.105
9.104
3,5.103
2,2.105
2,2.105
2,2.105
2,1.105
1,21.104
1,19.104
1,16.104
1,05.104
4,7.102
5,2.101
If the regenerated fuel is stored for some time and then irradiated in the reactor, the amount of accu-
mulated transplutonium elements increases as a result of the initial presence of 241Am. Table 1 shows the
effect of storage time of regenerated. fuel on the accumulation of transplutonium elements in plutonium fuel
unloaded from the intermediate enrichment zone in the reactor.
The content of 241Am and 242Ku in the unloaded fuel increases 3 and 5 times,, respectively, when the
storage time of the regenerated fuel is increased from 0 to 3 years. The increase of 242Ku content will sig-
nificantly change the neutron and a activities and the residual heat release in the unloaded fuel.
Activity of Fission Products. Residual Heat Release
The activity of fission products must be known for evaluating both the degree of decontamination of fuel
in chemical reprocessing, and the radiation environment and residual fuel release in transport and technologi-
cal operations within the reactor and in the transportation of spent fuel to the reprocessing plant.
Table 2 shows the total activity of solid fission products (T1/2 > 5 days) of spent fuel from the reactor
core as a function of cooling time after unloading. After a cooling time-of one-half year, most of the fission
products activity comes from (95Zr + 9Nb), 106Ru, and 144Ce nuclides. With increasing cooling time, the activity
of fission products is governed by.106Ru, 144Ce, 147Pm, 134Cs, and 137Cs nuclides. 134Cs is produced by activa-
tion of the stable cesium isotope.
"Sr and 131Cs.
During the first half year of cooling, the total y activity of radium fission products decreases?by a factor
of 10, in the next half year the activity decreases by half so that from this point of view prolonged storage of
fuel is not advisable.
To secure nuclear safety, a certain amount of fissionable material is loaded into the dissolver appara-
tus.. As a result, the concentration of fission products in dissolved plutonium fuel of the fast reactor will be
after chemical reprocessing lower than in dissolved fuel of thermal reactors.
For cooling times up to 3 years the most difficult to purify in chemical reprocessing are the nuclides
"(Zr + Nb), 103Ru, and 106Ru [3]. Table 3 shows the activity of these nuclides as a function of fuel cooling
time.
Of considerable importance in the external fuel cycle technology are volatile and gaseous fission prod-
ucts. Table 4 shows the activity of. such fission products of plutonium fuel as a function of storage time after
unloading from the reactor.
With the exception of 10-yr 85Kr with an activity of 10-12 Ci/ton of fuel and the long-lived 129I, whose ac-
tivity after 6 months is comparable to the activity of 1311, only an insignificant amount of volatile and gaseous
fission -products remains after storage of 6 months or more. For. cooling times of 6 months or less, the ac-
tivity of 1311 . increases drastically and requires a special decontamination system.
In addition, the fuel of fast reactors contains radioactive 3H and 14C. Tritium forms in the fuel elements
in the course of irradiation in the reactor; spent fuel contains 0.5-5% of the total amount of tritium producing
up to 70 Ci/ton of fuel. A. conference of the IAEA established that the removal of tritium does not present special
difficulties [4].
After prolonged storage (>3 years) the activity of fission products is determined chiefly by the nuclides
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TABLE 5. Neutron Activity of Spent Fuel
in 109 neutrons/sec -ton (U + Pu)02
0 10.251 0-,51 1 1 2
242KU
2,8
1,9
1,3
0,6
0
1
2441(u
1,2
1,2
1,2
1,1
,
1,1
1,1
Total activity
4
3,1
2,5
1,7
1,2
1,1
14C is formed in the reaction 14N(n, p)14C, and with a nitrogen concentration of 0.1% in the oxide fuel the
activity of 14C amounts to about 10 Ci/ton of fuel. According to [4], the escape of 14C into the atmosphere during
reprocessing can act as a source of population irradiation. The problem of 14C is still not clear and requires
further studies. The activity of the structural materials of fuel elements and thermal assemblies is governed
by 54Mn and 58'60Co isotopes and after 6 months of cooling amounts to 3 ? 105 g-equiv/ton of steel. The activity
of steel drops by a factor of 2.5 after 1 year of storage and by a factor of nearly 15 after 3 years.
Neutron activity of spent fuel is determined chiefly by the content of 242Ku during the first year of storage
and by the content of 244Ku afterwards. Table 5 shows data on the neutron activity of fuel unloaded from the
intermediate enrichment zone as a function of storage time, the storage time of the regenerated fuel before
irradiation being zero.
As the storage time of regenerated fuel increases, the neutron activity of spent fuel increases as a re-
sult of increasing 242Ku content. Because of the high neutron activity of spent fuel, the shields of transporta-
tion containers should include hydrogenous materials since .heavy materials such as steel and lead used as y-
radiation shields are not effective against neutrons.
After a cooling time of 0, 0.25, 0.5, 1, and 2.5 years the residual heat release is 42.5, 5.6, 3.5, 2.0,
and 0.7 kW per fuel assembly respectively, decreasing by a factor of 12 after the first 6 months and by a fac-
tor less than 2 in each succeeding 6-month period.
The contribution of 242Ku into residual heat release is 2, 11, 12, 10, and 3% after 0, 0.25, 0.5, 1, and
2.5 years of storage, respectively.
After 1 year of storage of regenerated fuel the contribution of 242Ku into residual, heat release can reach
25% (6 months cooling time).
Natural Activity of Regenerated Fuel and Radiation
One of the components of natural activity of plutonium fuel is a activity. According to [1, 2] the total a
activity of plutonium fuel with an isotopic composition as obtained in a thermal reactor is about 230 Ci/kg PuO2
and is determined chiefly by 238Pu. The a activity of fast-reactor plutonium of equilibrium composition is ap-
proximately one-half of the a activity of thermal-reactor plutonium. High a activity necessitates hermeti-
cally sealed technological equipment for handling unshielded plutonium.
The natural neutron and 'y fuel activity governs the degree of exposure of personnel in the course of.
production of fuel elements and heat-generating assemblies.
Neutron activity of fuel is due to spontaneous fission of 238pu, 240pu, 242pu and to the (a, n) reaction with
oxygen.
Neutron activity of thermal-reactor plutonium amounts to about 4. 105 neutrons/sec ? kg of PuO2 [1, 2].
The contributions of spontaneous fission and of the (a, n) reaction are approximately the same. Neutron acti-
vity of equilibrium plutonium from a fast reactor is equal to the neutron activity of plutonium from a thermal
reactor. The neutron activity due to the (a, n) reaction in light-element impurities (Be, B, F) is comparable
to natural neutron activity of plutonium fuel when the amount of impurities exceeds 102 mass %.
Gamma activity of plutonium fuel is determined by "soft" characteristic radiation (13-17 keV) and amounts
to about 1012 y-quantum ? sec/kg PU02 for thermal-reactor plutonium and about 4. 1011 y-quantuni ?sec/kgPuO2 for
equilibrium plutonium from a thermal reactor. The y radiation intensity of 241Pu (in equilibrium with 237U) and
241A.m isotopes exceeds the intensity of soft characteristic radiation.
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TABLE 6.. Decontamination Factor for Fis-
sion Products
Cooling time,
years ..
96(Zr + Nb) rosRu
0,25
5,3.107
1,3.106
6,5.106
0,5
2,3.107
2,5.105
5,5.106
1,0
3,5.106
1.104
4.106
2,5
1.104
1,5.106
In the absence of 241Am, the dose rate from unshielded plutonium-fuel of a thermal reactor is deter-
mined by 238Pu radiation and can reach up to about 3000 MR/sec on the surface of plutonium dioxide powder;
the dose rate of equilibrium plutonium of fast reactors is 1000 ?R/sec. After 1 year of storage of thermal-
reactor plutonium, the accumulation of 241Am raises the dose rate by about 1500 ?R/sec.
The dose rate on the surface of thermal-reactor fuel pellets with 18% plutonium enrichment can reach up
to ^'400t.4LR/sec and decreases rapidly with distance so that manual handling of such pellets is safe.
The y-radiation dose rate of fuel elements and heat-generating assemblies is due mainly to decay prod-
ucts of 241Pu 237U, and 241Am. On the surface of fuel elements (thermal-reactor plutonium) the dose rate is
- 50 ?R/sec and is equal to the maximum permissible dose rate for manual handling at a distance of 5 cm from
the surface. The dose rate of equilibrium plutonium of fast reactors is approximately one-half as high. After
1 year of storage the dose rate of regenerated fuel increases by a factor of nearly 2 as a result of 241Am ac-
cumulation.
The dose rate of heat-generating assemblies usingthermal-reactor plutonium is also determined by y
and neutron radiation and can reach up to 60-80 irem/sec ontheassembly surface depending on the degree of
fuel enrichment. The dose rate is not affected significantly by accumulation of 241Am in regenerated plutonium
during storage (5% increase per one storage year). The contribution of 236Pu decay products into the dose
rate is small even after 10 years of storage.
To ensure radiation safety in. fuel assembly operations, a 15-cm-thick shield of a hydrogenous material
should be mounted at the level of the active part of the stack providing free access to the cap and stem of the assem-
bly. The dose rate from an unshielded heat-generating assembly does not exceed the maximum permissible
value at a distance 2 in from its surface.
The necessary decontamination factors can be evaluated considering natural activity, the activity of plu-
tonium fuel fission products (see Table 3), and the permissible concentration of fission products which is diffi-
cult to purify in chemical reprocessing. Assuming that the contribution of these fission products into the sur-
face dose rate of heat-generating assemblies does not exceed 10% (for each isotope), the permissible content
of these nuclides in regenerated plutonium fuel has been calculated as follows: 95(Zr + Nb) 0.1 mCi/kg, 103Ru
2 mCi/kg, and 106Ru 0.4 mCi/kg of (U +Pu)O2.
Table 6 lists the necessary decontamination factors for these nuclides.
An analysis of fast-reactor fuel reveals certain specific features that are important in planning the entire
fuel cycle technology.
The principal drop of activity of spent fuel takes place during the first 6 months.
The problems of gaseous activity must be analyzed if the cooling time is less than, 6 years.
The problem of 241Am can have a significant effect on fuel cycle technology. Increasing the duration of
the entire fuel cycle causes partial loss of 241Pu as a result of its decay into 241Am. Increasing the duration
of the external cycle interval between. chemical reprocessing and mounting the heat-generating assemblies in
the reactor results in an increased accumulation of 242Ku which increases the neutron activity of fuel and the
heat release.
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The natural fuel activity should be given much attention at such stages as the preparation of fuel elements
and heat-generating assemblies, transportation of "virgin" fuel, and initial inspection in the reactor. Manual
operations can be carried out despite significant levels of natural radiations.
The transportation and handling equipment should be capable of unloading and storage of spent fuel after
a short cooling time.
High activity and heat release complicate transportation and distribution of spent fuel.
All stages of fuel cycle are interrelated and necessitate a comprehensive technological and economical
analysis which should take into account all specific features of fuel discussed above.
LITERATURE CITED
1. R. Noyes et al., Nucl. Technol., 26, 460 (1975).
2. L. Faust et al., Nucl. Technol., 15, 249 (1972).
3. V. B. Shevchenko (editor), Chemical Technology of Irradiated Nuclear Fuel [in Russian], Atomizdat,
Moscow (1971).
4. Decisions of the Conference of Experts of the IAEA on the Regeneration of Fuel for Fast Reactors,.
Leningrad, May 17-21, 1976.
5. O. D. Bakumenko et al., Report at the Conference of Experts of the IAEA on the Regeneration of Fuel
for Fast Reactors [in Russian], Leningrad, May 17-21, 1976.
ATMOSPHERIC RELEASE OF VOLATILE FISSION
PRODUCTS FROM OPERATION OF NUCLEAR
POWER REACTORS AND SPENT FUEL
REPROCESSING FACILITIES AND
PROSPECTS FOR EXTRACTING THE PRODUCTS
B. Ya. Galkin, L. I. Gedeonov,
N. N. Demidovich, R. I. Lyubtsev,
I. V. Petryanov, B. F. Sadovskii,
V. N. Sokolov,. and A. M. Trofimov
One of the most important scientific-technical problems in the development of nuclear power is that of
ensuring a high level of extraction of radioactive elements in gaseous emissions from reactors and radiochemi-
cal reprocessing plants. The importance and urgency of this problem are evident at present, in spite of the
fact that, among all types of industrial and agricultural human activity, the safety of the personnel, the popula-
tion and the surroundings is highest in the atomic and the nuclear power industry. Here the main concern is
in removing relatively long-lived volatile fission products from the gaseous emissions, primarily tritium,
iodine isotopes and radioactive inert gases (RIG), in addition to collecting aerosols.
As a rule, short-lived isotopes of iodine and RIGs are found in the local radioactive makeup of the
atmosphere and constitute a radiation constraint only near nuclear objects. Because of their rapid decay,
their contribution to the radiation dosage over large areas is insignificant.
The long-lived nuclides 3He, 85Kr and 129I propagate in the atmosphere on a global scale, and further de-
velopment of nuclear power requires a comprehensive examination of possible effects associated with con-
tamination of the surrounding medium by these nuclides. The accumulation of gaseous fission products in
nuclear fuel depends on the burn-up level and the type of fuel (and reactor). For example, the tritium content
in thermal and fast reactor fuel is 200 and 2000 Ci/ton, respectively, for 85Kr it is 5000 and 20,000 Ci/ton,
and for 129I it is from 10 to 15 Ci/ton. When one considers that in 2000 years the total reactor power will reach
4000-5000 million kW, one can calculate the expected accumulation at that time of tritium, iodine, and krypton.
Translated from Atomnaya Energiya, Vol. 44, No. 2, pp. 145-149, February, 1978.
0038-531X/78/4402- 0153 $07.50 ?1978 Plenum Publishing Corporation 153
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At present in nuclear fuel cycle operations the 85Kr is practically ejected, into the' atmosphere. The main
source for -its accumulation in the surrounding medium is reclamation plants for spent nuclear fuel.
Krypton is absorbed to an extremely small degree. by dry rock, is absorbed very little by the oceans,
and is not assimilated by living organisms. As a result, the chief process whereby 85Kr is eliminated from
the atmosphere is its radioactive decay (T1/2 =10.6 years).
In the world today most of the- 85Kr is distributed mainly in the troposphere of the northern hemisphere.
Its concentration in air at ground level is ^?20 pCi/m3. If nuclear power develops at the estimated rate, the
radiological capacity of the atmosphere in regard to 85Kr, equal to 1.5. 104 MCi; will be exceeded before the
year 2000; the annual emission of krypton at that time will be 1.7. 103 MCi, and the specific activity of air in
regard to this nuclide will reach 10-9 Ci/m3. Under unfavorable annual conditions the local contamination of
the atmosphere in a fuel reprocessing locality may exceed the limiting allowable concentration by a factor of
10-100. Accumulation of 85Kr in the atmosphere leads to deterioration of the environment and is a problem in
air-separator plants, where by the year 2000 the radiation dosage received by the workers may be 10-15 mrem/
yr: a user of stable krypton will obtain up to 15 mCi of 85Kr with each standard bottle.
While the technical capability presently exists to seal fuel rods and thereby localize the main mass of
85Kr in the spent fuel elements, and to concentrate the removal of gaseous emissions from it at the reproces-
sing plant, for tritium, however, it is a different matter. Since it has an extremely high permeability, it can
diffuse through the fuel envelope during reactor operation. and enter the surrounding medium as liquid and
gaseous emissions, both at the power station and in the fuel reprocessing plant. Existing data indicate that
the residual content of tritium, e.g., in the spent fuel of fast reactors, enclosed in a stainless-steel envelope,
constitutes only a few percent of the theoretical accumulation. It follows from this that to prevent the emis-
sion of tritium to the atmosphere one must purify the emissions, not only in the reprocessing plant, but also
in the power station.
The tritium resulting from power plant operation propagates globally. However, in contrast with RIG it
is relatively easily oxidized and forms a compound with water as the HTO molecule. The comparatively rapid
removal of tritium from the atmosphere and its scattering in the surrounding medium as liquid debris leads to
its nonuniform geographic distribution. For example, with a background concentration of tritium in atmos-
pheric precipitation in 1975, its concentration in water at the lower levels of the Danube was from 130 to 200
TE. In the same period in the surface water of the Black Sea the tritium concentration was 30TE, and inthe
water of the Baltic Sea it was 60 TE. In atmospheric precipitation water over the USSR, the tritium content
in 1972 was 90-230 TE; in the Leningrad region during 1974=1976 it was 30-60 TE. In spite of the nonuniform
distribution of tritium, it is assumed that finally it will accumulate mainly in the mixing layer of the ocean
water.
Information on the presence of 129I in the surrounding medium is presently limited mainly to data from
the immediate vicinity of nuclear plant and experimental nuclear test ranges. For example, its concentration
in air at various points of the U. S. A. ranges from 104 to 109 atoms/m3 of air, and in air above the oceans it
ranges from 105 to 3. 106 atoms/rn3.
Analysis of the propagation of volatile radionuclides around emission sources shows that, prior to global
mixing, they will exist in concentration exceeding the background at a distance upto several hundreds of kilo-
meters from the source. For an expected average distance on the order of 100-120 km between nuclear plants
in some regions there will be a field which adds the emissions from individual plants, and thus, there will be a
danger of regional contributions to the surrounding medium and the population.
The purification of gaseous emissions is universally recognized to be an inherent facet of the technology,
of spent fuel reprocessing. Recently, both nationally and internationally, a trend has appeared towards formu-
lating standards for emissions. For example, in the U. S. A. it has been established that the standards for
emission of 85Kr in that country may be limited to 50,000 Ci/yr' 1000 MW. This means that a 10-20-fold puri-
fication of gaseous emissions from krypton is required.
Many factors determine the fraction of gaseous fission products (GFP) released in different reprocessing
operations, and therefore an absolute value for each stage cannot be predetermined for all cases. Practical
experience shows that, e.g., in the process of mechanical cutting of fuel elements from water-cooled -water-
moderated reactors, about 5-10% of the iodine, up to 20% of the tritium, and" 40-50% of the krypton are carried
away with the air flow. The remaining part is retained by the fuel, goes into solution with the fuel, and is sub-
sequently distributed between the gaseous and aqueous phases.
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The variety of forms in which iodine occurs makes the problem of removing it a very complex one. Us-
ually the following forms are found: molecular iodine, iodides, iodates, and iodo-alkali (as a rule, methyl
iodide). Some fraction of the iodine comprising part of the initial solutions can be retained through the extrac-
tion operations and is then distributed between the aqueous and the organic phases, thereby setting up condi-
tions for contaminating both the final and the intermediate products. This last fact makes for accumulation of
iodine in the washing solutions and subsequent unmonitored emission in the low-activity separation line.
All that has been said confirms that it is desirable to remove iodine prior to the extraction operations.
A very effective method of removing iodine from solutions after the fuel is dissolved is to blow it away with a
stream of air (both in the molecular form and in the form of iodo-alkalis), and subsequently to absorb it from
the gas-air stream by means of solid absorbers or solutions. Many kinds of solid absorber, based on inor-
ganic and organic substances, are well known, containing different impregnators (nitric acid or silver iodide,
etc.), which remove iodine efficiently: in element form up to 99.95%, and in metal iodide form up to 99.90% A.
negative facet of solid absorbers is their low specific capacity and the difficulty of regeneration (for repeated
use), which is particularly important in regenerating fast reactor fuel, containing a high iodine content. In
this respect liquid absorbers deserve attention, e.g., solutions of inorganic salts, e.g., mercury nitrate, and
organic solvents and have significantly higher specific capacity with respect to iodine. However, they are
inferior to solid absorbers in regard to the purification of gases from iodine vapor. Therefore, it is more
promising to employ purification systems containing two successive operations: absorption by liquid absorbers
and a final cleaning with solids. Any method of cleaning unavoidably meets the problem of safe containment
of residues containing 1291.
A no less important task is that of complete purification and subsequent burial of tritium. Its distribu-
tion between the liquid and gas phases leads to the situation where one must create different purification sys-
tems. While the vapor phase can be collected in a small volume, after condensation in cold traps, liquid tri-
tium-containing products (condensates from chilling the first cycle refined products) are distinguished by hav-
ing a large volume, and for this reason their burial is difficult. At present intensive studies are in progress
to assess possible recycling of tritium-containing condensates (and nitric acid) in the technological process.
A possible method of removing tritium being considered is volumetric oxidation of the fuel (voloxidation),
followed by removal of the RIG in a minimum volume. This method is being studied also from the point of view
of removing other volatile products, iodine, RIG, and possibly 14C, although the problem of 14C is less clear,
both because of the sources of its formation, and also because of the uncertainty in determining it.
The single inert gas isotope which is a radiation hazard is 85Kr. Until now existing plants for reclaiming
fuel have not established systems for eliminating krypton, at least systems designed for full-plant power. How-
ever, in the not too distant future such systems will apparently be commissioned at all the plants.
The principles which are being followed in setting up facilities for purifying gaseous emissions from RIG
at the reactors and in the reclamation plant differ substantially, due to differences in the chemical and isotopic
composition of the emissions. The purification systems at the reactors are intended to eliminate short-lived
nuclides, which, in turn, allows short-duration gas containment systems to be used as the main technological
agent, set up in the emissions process line. A. system which would ensure operation of such a process must
act for a long period without the need for periodic regeneration of the cleaning agent. Here the short-lived
RIG nuclides decay and are localized in the cleaning system itself, without reducing its efficiency of opera-
tion. In the practical operation of reactors of water-cooled-water-moderated reactors such systems are
based on absorption of RIG from gaseous emissions by activated charcoal. An example of this kind of system
is the facility operating at the Kolsk reactor which reduces the RIG activity in the emission gas stream by a
factor of 200, a level which fully meets the safety requirements.
To purify gaseous emissions of regeneration plant one must use low-temperature rectification pro-
cesses, low-temperature absorption, and also a method for selective absorption of fluorocarbons. All the
residual well-known methods for RIG absorption drop from consideration, when they are evaluated for an in-
dustrial scale.
Low-temperature rectification and absorption processes offer not only efficient cleaning, but also sepa-
rate out the krypton and xenon. Here it should be noted that the gaseous plant emissions from reclamation of
nuclear fuel can be a raw-material source for obtaining stable xenon, enriched with xenon to a factor of about
104 greater than air. In a single plant, reprocessing 5 tons/dayof nuclear fuel, one can obtain up to 1000 m3/day
of high-cost stable xenon, for which the requirements are continuously increasing. The utilization of stable
xenon, obtained from the gaseous emissions, can compensate to some extent for the cost of creating such a
plant to clean and separate the inert gases.
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 Declassified and Approved For Release 2013/03/07: CIA-RDP10-02196R000700110002-4
Among the methods for retaining 85Kr, removed during purification of gases, it is profitable to consider
storing it in special underground vaults (geological beds), or in tanks under pressure, and also converting it
to clathrate compounds.
A problem very closely related to that of cleaning RIG is purifying the gas-air emissions from radio-
active aerosols, which are formed in great quantity in reprocessing plant. In nature these aerosols may be
both condensed and dispersed. The main mass of aerosols takes the form of salt mists, and also mists of
acids, and the mass concentration of the dispersed substance usually amounts to tens of milligrams, and in
some cases up to several grams, per 1 m3 of gas,. The basic radioactivity of the mists comes from particles
of size no less than 1.0 ?.
To shield the surrounding environment and the safety zones of plant from aerosol contamination one needs
a reliable continuously acting purifying system, with purification factors suitable for the allowable atmospheric
conditions.
Experience of radioactive mist cleaning has shown that the most reliable method is to use fiber-type fil-
ters, which give the necessary degree of cleaning and do not require special treatment.
At present fiber self-cleaning filters have been developed and used successfully; one can subdivide them
into preliminary and high-purification filters. The process of filter self-cleaning is one where liquid particles
precipitating on the fibers migrate to the filter layer under the action of different forces, and are removed as
a result. Thus the filter characteristics remain constant with time.
The external and internal filtration mechanisms, both for coarse and for fine fiber filters, including all
the changes in the filtering layer, which occur in the presence of liquid-in the filter (e. g., formation of secon-
dary drops on ultrafine fibers), have been thoroughly investigated and described in the work of Soviet special-
ists. On the basis of this work, self-cleaning filters of fine glass fiber have been constructed which do not re-
quire force removal of liquid. To overcome capillary forces which oppose the efflux of liquid, the filtering
layer is located vertically in the fine filters, and its height is significantly larger than the height of capillary
rise of the liquid in the filter, layer. Such filters are produced in different modifications (cassette-wedge and
cylindrical) with filtering layer area of from 1.5 to 5.6 m2, and are used as technical equipment for long-term
application.
The fuel reprocessing and fuel element manufacture processes include operations accompanied by the
formation of a large amount of dry dust, and -in some cases this dust is a valuable product, which must be re-
turned to the technical system. In recent years filters based on metallized cloth with a high efficiency for re-
moving solid particles have been increasingly used to purify gases containing dust, and they operate very well
at both low and high temperatures (up to 500?C). Metal-cloth filters have high strength and corrosion resis-
tance, they are simple to manufacture, and, in contrast with metal-ceramic materials, have considerably
less aerodynamic drag.
In the Soviet Union metal-cloth filters are widely used to extract dry radioactive dust from gaseous
emissions. The efficiency of filters under self-filtering conditions, for particles with mean diameter of less
than 1 g, is 95-99%, and it is 99.5-99.97o for particles of 2.5-3 ?. The usual filtering material is a mesh of
heat-treated wire (stainless steel) of two sizes: diameter 0. 09/0.055 mm and 0.064/0.032 mm. The allowable
dust concentration is up to 40-50 g/m3.
Operating experience indicates that metal-cloth filters can be used satisfactorily as a first stage of gas
purification for high dust concentration in the gaseous emissions.
The scientific- engineering level of development achieved at present allows us to predict confidently that
a combination of existing systems for purifying gases from aerosols and dust, and complex schemes for re-
moving volatile fission products from the gases will allow one in coming years to reliably protect the surround-
ing environment and the population from contamination- associated with operating reactors, atomic facilities
and reactor fuel reprocessing plant,. and that this undoubtedly will promote the growth of nuclear energy at the
predicted rate.
1. R. Clarke, in: Proceedings of the IAEA Symposium: "Population dose evaluation and standards for man
and his environment," Vienna, May, 1974, Rep. No. 184.
2. D. Beninson, ibid., Rep. No. 102. -
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 Declassified and Approved For Release 2013/03/07: CIA-RDP10-02196R000700110002-4
3. Radioactivity of the Environment, Dokl. NKDAR OON, 2, A/AC. 82/R (1971).
4. Yu. I. Koryakin, At. Tekh. Rubezhom, No. 1, 3 (1975)
5. L. Ottendorfer, IAEA Advisory Group to study questions of mutual cooperation between countries in the
Danube catchment area, Belgrade, NAG-41 (1975).
6. P. Bryant and J. Jones, Rep. No. RPB/R8 (1972).
7. W. Schell, G. Sauzay, and B. Payne, in: Proceedings of the IAEA Symposium: "Population dose evalua-
tion and standards for man and his environment," Vienna, May, 1974, Rep. No. 34.
8. F. Brauer, H. Rieck, and J. Hooper, in: Proceedings of the IAEA Symposium: "Physical behavior of
radioactive contaminants in the. atmosphere," Vienna, Rep. No. 6, Nov. (1973).
9. A. T. Ageenkov et al., At. Energ., 41, No. 1, 23 (1976).
10. Yu. V. Sivintsev, At. Tekh. Rubezhom, No. 12, 25 (1963).
11. T. I. Smolkina and A. A. Chubakov, At. Energ., 18, No. 3, 298 (1965).
12. B. F. Sadovskii et al., Dokl. Akad. Nauk SSSR, 199, No. 1, 154 (1971).
13. B. F. Sadovskii and I. V. Petryanov, Protecting the Atmosphere from Contamination: Part 2: Methods
of Determining Atmospheric Contamination, Sb. Akad. Nauk Lit. SSR [in Russian], Izd. Inst. Fiz. Mat.,
Vilnius (1974).
14. 1. V. Petryanov and B. F. Sadovskii, Cloth and Fiber Filters: Theory, Methods of Investigation and
Operation, Coll. Papers of Symp., P. Akad. Nauk, Visla-Partechnik (1975).
15.
16.
17.
18.
19.
20.
B. F. Sadovskii et al., Dokl. Akad. Nauk SSSR, 202, No.4, 886 (1972).
B. F. Sadovskii, Prom. San. Ochist. Gazov, No. 3, 14 (1975).
A.
S.
Mandriko and I.
L. Peisakhov, Tsvet. Met., No. 11, 51 (1970).
A.
S.
Mandriko and I.
L. Peisakhov, Nauchn. Tr. Giredmeta, 40, 74 (1972).,
I.
E.
Nakhutin et al.,
Fourth Geneva Conference, Dokl. SSSR, No. 49/R/70 (1971).
V.
B.
Shevchenko, I.
E. Nakhutin, and E. V. Renard, "Problems in using thepyrexprocessinregener-
ating fast neutron reactor fuel," Paper at the meeting of IAEA experts to discuss problems in regenerat-
ing fast reactor fuel," Leningrad, May 17-21 (1976).
21. W. Burch, O. Varbro, and W. Grocnier, "Retention of gaseous fission products in reprocessing LMFBR
fuels," ibid.
22. L. Baetsle, "Head-end and nuclear gas purification research on LMFBR fuel reprocessing in Belgium,"
ibid.
23. J. Sauteron et al.," Le retraitement des combustibles "rapides" en France," ibid.
24. L. I. Gedeonov, L. N. Lazarev, andA. N. Suprunenko, "Environmental protection in regard to the repro-
cessing of fast reactor fuel," ibid.
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 Declassified and Approved For Release 2013/03/07: CIA-RDP10-02196R000700110002-4
PROBLEMS IN TRANSPORTING REPROCESSED
NUCLEAR- FUEL
A. N. Kondrat'ev, Yu. A. Kosarev,
and E. I. Yulikov
At present the total nuclear reactor power in the USSR is - 7.9 million W. According to the main trends
in the development of the industry in 1976-1980, the total nuclear reactor power in 1980 will be 13-15 million
W. In the future the rate of introduction of new nuclear reactors must increase. Basically nuclear power
stations with thermal reactors will be built, and fast-neutron reactors are proposed for construction after
1985, and the rate of construction will depend on the amount of plutonium generated in the thermal reactors.
The individual power levels of the main types of reactors in the period 1986-1990 will be 1-1.5 million W.
Some features of the fuel of existing and future reactors 'are shown in Table- 1.
The reprocessed fuel of the reactors constructed in member-countries of the Council of MutualEconomic
Aid (CMEA) will be transported for reprocessing to the Soviet Union. The total reactor power in these coun-
tries by 1980 will increase to 7.3 million kW, which will require a corresponding increase in the number of
transport operations to handle this fuel. Analysis shows that the most convenient and economical transport
is by rail, since this form of transport is widely developed in the USSR and the other member-countries of
CMEA: there are rail lines at almost all the reactors; and the. cost is lower than automobile transportation.
A single railway train can carry the annual output of spent reactor. fuel from an installed electrical power of
1 million W.
Specially developed and constructed wagon containers (Fig. 1) are used for rail transport, in which the
containers are located vertically or horizontally, depending,on the size of the fuel bundles. Bundles less than
3.5 m in length are transported in vertical containers, and the others - in horizontal containers. Consider-
ably more fuel is held in the vertical containers: The wagons have movable sections in the top, which makes
it easy to load and unload. The wagon sizes are standardized. In the USSR the maximum height and width of
wagons is, respectively, 5300 and 3700 mm, and in member-countries of the CMEA, it is 4650 and 3150 mm.
The allowable rail loading for railways in the USSR and CMEA. member-countries is 22 and 18 tons per wheel
pair. Because the railway gauge in the USSR and other CMEA member-countries differs (1520 and 1435 mm)
the container wagons are equipped with both pairs of wheels, which are changed at the border stations.
At present in the USSR single-layer steel containers (Table 2) are being developed and manufactured,
lined internally with stainless steel. The outer surface of the container has welded to it fins and special
TABLE 1. Fuel Characteristics of USSR
Power Reactors
-
Reactor
Elec-
trical
power,
MW
Fuel
charge,
UO2, tons
Av, burn
up level,
MW
day/ton
No.
of
bun-'
dies
Bundle di-
mensions,
S XL', mm
VV E R-210
210
44
13
349
144 X 3200
VV R-365
365
44
28
349
144 X 3200
VV R-440
440
44
28
349
144 X 3200
VV R-1000
1000
72
41
151
238X4665
RBMK