Soviet Atomic Energy Volume 17, No. 5

Body: -111111101111111 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Volume 17" No 5 rc ( 1964 SOVIET ATOMIC ENERGY ATOMHAFI (ATOMNAYA iNERGIYA) TRANSLATED FROM RUSSIAN CONSULTANTS BUREAU Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 VOLUME 9 Advances in Cryogenic Engineering Proceedings of the 1963 Cryogenic Engineering Conference August 19-21 at Boulder, Colorado The 69 papers contained in this volume (the latest in a series recognized as the most authoritative' compilation for low tempera, ttFe engineering) examine such widespread areas as: insulation, heat transfer,/ mechani- cal properties, seals, fluid phenomena, engi- neering aspects of superconductivity, therma- dynamics, phase equilibria, safety, instrumen- tation, cryopumping, and various develop- ments in processes and equipment. Of par- ticular importance are the chapters describ- ing mechanical properties, which will be of interest to all those in materials' research (especially .for the new alloy 7039) the sec- tions on superconducting materials and prop- erties for all those working with solid-state devices, and papers concentrating on thermo- dynamic properties for designer's of cryo- genic systems.. 598 pages Edited by K:D. Tirnmerhaus ? The information on recent developments contained in this ,volume will be a: valuable ' aid to all low temperature engineers in the resolution of such practical problems as heat transfer in cryogenic storage and in flight, space-simulation chamber design and opera- tion, fluid transfer of large cryogenic flows, mechanical property selection, ? seal evalua- tion for cryogenic applications, and many others. The equipment and procedures used. by 'various investigators in performing the work reported in this volume?varying from such complex and elaborate apparatus as a space rocket, to intricate and special testing equipment?indicate the breadth and scope of the topics covered. An author and a cumula- tive subject ,index (Volumes 1-9) are in- cluded. Every paper accepted for publica- tion in ADVANCES IN CRYOGENIC ENGINEERING is being published for the first time. Complete contents of Voldmes 1-8 will be sent on request. $17.50 PLENUM PRESS 227 W. 17th St., New York, N.Y. 10011 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 I Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 ATOMNAYA kN_ERGIYA EDITORIAL BOARD A. I. Alikhanov A. A. Bochvar N. A. Dollezhal' K. E. Erglis V. S. Fursov I. N. Golovin V. F. Kalinin N. A. Kolokol'tsov (Assistant Editor) A. IC: Krasin I. F. Kvartskhava A. V. Lebedinskii A. I. Leipunskii M. G. Meshcheryakov M. D. Millionshchikov (Editor-in-Chief) I. I. Novikov V. B. Shevchenko A. P. Vinogradov N. A. Vlasov (Assistant Editor) M. V. Yakutovich A. P. Zefirov SOVIET ATOMIC ENERGY A translation of ATOMNAYA iNERGIYA A publication of the Academy of Sciences of the USSR ? 1965 CONSULTANTS BUREAU ENTERPRISES, INC. 227 West 17th Street, New York 11, N. Y. Vol. 17, No. 5 November, 1964 CONTENTS PAGE ENG. RUSS. The Third International Geneva Conference?A. M. Petros'yants 1065 323 High-Temperature Reactor-Converter "Romashka " ?M. D. Millionshchikov, I. G. Gverdtsiteli, A. S. Abramov, L. V. Gorlov, Yu. D. Gubanov, A. A. Efremov, V. F. Zhukov, V. E. Ivanov, V. K. Kovyrzin, E. A. Koptelov, V. G. Kosovskii, N. E. Kukharkin, R. Ya. Kucherov, S. P. Lalykin, V. I. Merkin, Yu. A. Nechaev, B. S. Pozdnyakov, N. N. Ponamarev-Stepnoi, E. N. Samarin, V. Ya. Serov, V. A.Usov, V. G. Fedin, V. V. Yakovlev, M. V. Yakutovich, V. A. Khodakov, and G. V. Kompaniets. 1071 329 Development of Superheating Power Reactors of the Beloyarsk Nuclear Power Station Type ?N. A. Dollezhal', I. Ya. Emel'yanov, P.1. Aleshchenkov, A. D. Zhirnov , G. A. Zvereva, N. G. Morgunov, Yu. I. Mityaev, G. D. Knyazeva, K. A. Kryukov, V. N. Smolin, L. I. Lunina, V. I. Kononov, and V. A. Petrov 1078 335 Novo-Voronezh Nuclear Power Station ?In Operation?N. M. Sinev 1088 ? Sodium-Cooled Fast Reactors?A. I. Leipunskii, 0. D. Kazachkovskii, I. I. Afrikantov, M. S. Pinkhasik, N. V. Krasnoyarov, and M. S. Poido 1090 345 Operating Experience with the Nuclear Propulsion Plant on the Icebreaker "Lenin" ?I. I. Afrikantov, N. M. Mordvinov, P. D. Novikov, B. G. Pologikh, A. K. Sledzyuk, N. S. Khlopkin, and N. M. Tsarev 1094 349 Experience in Operating the First Nuclear Power Station as an Experimental Facility ?G. N. Ushakov, L. A. Kochetkov, V. G. Konochkin, V. S. Sever'yanov, V. Ya. Kozlov, 0. A. Sudnitsyn, N. T. Belinskaya, P. N. Slyusarev, and V. A. Ivanov 1105 359 Containment of Plasma in a Trap with Combined Magnetic Field?M. S. Ioffe and R. I. Sobolev 1112 366 Calculation of Water-Moderated Water-Cooled Reactors?E. A. Garusov and Yu. V. Petrov 1121 375 The Burn-Up of Natural Uranium When it is Moved Axially in a Reactor?V. Bartosek and V. Lelek 1126 380 Column Packings Used in Isotope Separation?Yu. R. Akopov, I. G. Gverdtsiteli, V. A. Kaminskii, and G. L. Partsakhashvili 1133 384 Some Characteristics of Radiolysis Under the Influence of a Pulsed Beam of Fast Electrons ?V. L. Tal'roze and V. E. Skurat 1142 393 Producing High-Purity Tantalum?O. P. Kolchin and I. K. Berlin 1150 400 LETTERS TO THE EDITOR Thermodynamic Calculation of the Reaction Between Sodium and Water for a Sodium?Water Type Steam Heater?N. N. Ivanovskii and F. A. Kozlov 1155 406 Annual Subscription: $ 95 Single Issue: $30 Single Article: $15 All rights reserved. No article contained herein may be reproduced for any purpose what- soever without permission of the publisher. Permission ?nay be obtained from Consultants Bureau Enterprises, Inc., 227 West 17th Street. New York City, United States. of America. Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 CONTENTS (continued) The Critical Heat Fluxes in Tubes Carrying Monoisopropyldiphenyl, Heated Below PAGE ERG. I RUSS. the Saturation Temperature?F. F. Boganov 1159 408 Errors in the Calibration of y-Dosimeters with a Collimated Beam?F. F. Garapov, Yu. N. Gryaznov, and G. A. Dorofeev 1162 .410 SCIENCE AND ENGINEERING NEWS Interaction of Neutrons and Nuclei in the 1 eV-100 keV Range?L. B. Pikel'ner 1165 413 Symposium on Control Rod Physics and Control Rod Materials?I. R. 1168 414 A Symposium on Assay of Human Body Burden?Yu. V. Sivintsev 1170 415 A Polish Whole-Body Counter?Yu. V. Sivintsev 1173 417 New Device Unpacks Irradiated Targets?B. G. Cbistov 1175 419 News Item 1177 420 BIBLIOGRAPHY 1178 421 l'Two page insert facing page 336. The Russian date "Podpisano k pechati" of this issue was 10/16/1964 . This is equivalent to "approved for printing." Publication did not occur prior to this date, but must be assumed to have taken place reasonably soon thereafter. Publisher Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 THE THIRD INTERNATIONAL GENEVA CONFERENCE A. M. Petros'yants Chairman USSR State Committee for the Utilization of Atomic Energy Translated from Atomnaya gnergiya, Vol. 17, No. 5, pp. 323-328, November, 1964 In August and September of this year, the third UN international scientific and technical conference on the peaceful use of atomic energy, the largest meeting of the past decade, was held in Geneva. This conference was the most representative of all those held. Where 35 countries participated in the first Geneva conference, about 2000 scientists from 71 countries participated in thq third conference. The principal countries participating (38 countries) were represented by 750 reports including more than 100 from the USSR. About 350 reports were heard and discussed in 7 plenary and 35 sectional sessions. Whereas mainly scientific work was reported at the first conference in 1955, the main significance lay in the lifting of secrecy from some research and in the reestablishment of ties between the scientists and spe- cialists of the world ,which were broken in the course of the Second World War. A great deal of experimental work was made public at the second conference in 1958 which was of value in commercial applications, and direct con- tacts between countries and scientists working in the fields of atomic energy and nuclear physics were expanded. The third conference in 1964 summed up the experience with commercial application of atomic power instal- lations accumulated in the intervening years, emphasized confidence in the economic efficiency of atomic power stations in the immediate future, and stressed the inevitability of the widespread use of atomic power stations in many regions and countries of the globe. In some countries, they have come to the conclusion that it is now impossible to get along without the use of nuclear reactors for power purposes and that atomic energy will become one of the important factors in long-range economic development; in the immediate future, atomic energy will make it possible to satisfy the continually growing requirements for electric power in those regions and countries of the world where the reserves of fossile fuels are either exhausted or are close to exhaustion. Among the participants in this great international forum of scientists and engineers, were not only representa- tives of the highly developed industrial countries where atomic energy finds application in various fields of science and engineering, but also representatives of the countries of Asia, Africa, and Latin America which are just taking the first steps in the utilization of atomic energy, and of nuclear power in particular, or which are preparing to pro- ceed to its use. The main task of the third conference was consideration and discussion of questions connected with the devel- opment of nuclear power. UN Secretary General U. Thant noted in his speech that "the problems of nuclear power are key problems in the future development of the greater part of the world." The report of Prof. H. Bhabha (India) noted the need for the use of nuclear energy, especially in developing countries:"The principal part of the world population lives in the developing countries of the world, i.e., 2200 mil- lion people out of 3069 million or 71.8%, but the power requirement in these countries in only 20.6% of power pro- duced and electric power production is only 14.8%. If one assumes that the yearly increase in electric power re- quirements is 5%, the total power requirements of the world in 1960, which amounted to 4200 million tons of coal (equivalent per year), will increase to almost 30,000 million tons in the year 2000." Assuming that the growth in power requirements will be approximately 51 per year, H. Bhabha concludes that the world reserves of coal will be used up in approximately 75 years. One need not agree with these figures, but one ought to recognize that the posi- tion with regard to power resources in the world is not bright at all. In their report, the representatives of the Japanese Atomic Energy Commission said that "Japan has great in- terest in the peaceful use of atomic energy ..." The Japanese program for the development of nuclear power states: 1065 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 For the introduction of new Power production, the use of hydroelectric power in Japan is limited, arid therefore the main efforts will be directed toward the construction of thermal (ordinary) stations. A large part of the fuel for ther- mal stations must be. imported., The development of new power sources is reqUired. This consideration has led the Japanese economic authorities to realize necessity fOrthe production of atomic power." The representatives of a number of other governrrients also emphasized in their reports that nuclear power is; ??? for Many countries, a realistic and comPletely dePenciable means for Obtaining power. . . The third Geneva conference took place. under very favorable circumstances for international cooperation ? , among scientists of all countries. The signing of the Moscow agreement on August 5; 1963 banning the testing ,of nuclear weapons in three media, the annOuncernent,by,the three nuclear powers T.theSOviet Union, the United States, And biedi Britain ?concerning the reduction in production Of fissionable Materials for military prirposes, all these ? ? ? , ? . I ? ? ? ? ? created broad possibilities for Using the power of the atom for the good of Mankind and facilitated a Successful dis- cussion at the cOnferefice cOnceiiiing ihe problems of Using intrantidear energY for peaceful purposes. ? ?? ? .1., ? ? . ? , All the Main trends in the peaceful Use Of atomic energy were considered and discussed at the conference, . sorrie in greater 'detail; other's in 'snm?marY. However, MOst attention ,-Nis paid io aterhic power stations. Reports and papers on various problems were heard and discussed : atomic power stations and the economics of nuclear power, research reactors ; the physics of nuclear reactors, the application of radioactive isotopes and radia- tions in science and industry., Also discussed were problems directly connected With nuclear power such as problems . in the chemical processing Of rnieleOr fuel, the problems of materials for atomic techndlogy, the question of nuclear , safety and the medical and biological aspects of the use of atomic energy. New methods for utilizing atomic energy ? I ? were discussed : various methods for the direct conversion of nuclear and thermal energy to electricity,. the use of atOrnic reactors for desalinization of sea water; the problems in achieving controlled thermonkleat reactions, and many others. In the six years. which have gone by since the second international conference, nuclear power has made great strides foiward. In 1955, at the time Of the first Geneva conference, only the first atomic power stition, also the first commercial type of station in the USSR (Obninsk), was operating at a power of 5 MW > by 1958; several atomic power stations With a ,CaPacity over 180 MW were operating throughout the World; arid by the end of 1964, ihe total capac- ity of Oil the atomic Power stations in the World will reach almost 5000 MW. . - Sufficient practical experience in the Operation of atomic power stations was accumulated by the time of the third conference. Now, more than 35 atomic power stations throughout the .world "are Producing electric power, and , . there are about 30 atomic Power Stations in various stages of construction it is well known that there are operating commercial atomic power stations in the USSR, the United States; Great Britain, France, and ItOly. Atomic stations are operating, oi are in the process of construction, in Canada, japan, Czechoslovakia, East Germany, and other countries. Thus, by the time of the Third ihternOtiOnal Conference of Atomic Scientists,niiclear Power had been prOdUced ? . On a corhMercial scale, and a spirit of optimism prevailed With regard to the prospects for the development of Com- mercial nuclear power as a decisive tactor in the generation of electric power in a nurnber of countries. In accordance with their specialties and with local conditions, a number of countries have developed charac- teristic national Methods and programs for the development of nuclear power. These plans and programs Were re- ported at the conference, .Some of them provide for an increase in the capacity of atomic power stations to tens of milliOns of kilOwatis during the coming decade arid for considerable decrease in the cost of electric power produced by atomic power stations. Various countries are working on different types of reactors for atomic power stations depending on the partic- ular conditions. which obtain in a given country in the economic and industrial fields. For example, in the United States; where there are plants for the production of uranium highly enriched in the isotope U235, a great deal of ex- perience has been accumulated concerning reactor systems using pressurized water and boiling water.. These are the fundamental types of reactors that have been developed in the United States; they are the foundation for the de- velopment of nuclear power in the United States in the coming years. Great Britain is chiefly developing reactors with graphite moderation and gas cooling. France has also been working on gaS-graphite reactors lot a long tithe. 1066 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14 : CIA-RDP10-02196R000600120001-5 In Canada, the basic type has turned out to be the heavy water power reactor operating with natural uranium, which can be explained to a considerable degree by the lack of plants for the production of enriched uranium in Canada. In the USSR, large reactors of the water-cooled, water-moderated (light water) type have been, and are being, constructed as well as reactors with graphite moderator and water coolant. In the Soviet Union, the I. V. Kurchatov atomic power station was built at Beloyarsk in the Urals, and a commercial power reactor operates there with nuclear superheating of steam directly in the 100-MW electrical output reactor, the first in use in the world. The construc- tion of a second such reactor at the Beloyarsk atomic power station with an electrical output of 200 MW is going ahead at full speed ?the assembly of the reactor proper will be completed this year. Construction and assembly of the first section of the Novo-Voronezh atomic power station with a capacity of 210 MW has been completed; it supplies current to the Voronezh power system. Construction has begun on a second section of this station with an electrical output of 365 MW. In the Ul'yanovsk region, at the Scientific Research Institute for Atomic Reactors in the Melekess area, con- struction of an atomic power station with an output of 50 MW has been complete. Its reactor, of the water-cooled, water-moderated "boiling" type, operates with ordinary water in a single-loop circuit and direct_transmission of steam to the turbine. As has been reported in the literature, .a dual-purpose atomic power station with an electrical output of 100 MW is operating in Siberia, and the present total output of all the dual-purpose atomic reactors operating in Siberia is about 600 MW. As reports and statements at the third Geneva conference have indicated, the present trend in the development of nuclear power reactors of various types is toward an increase in their individual outputs to powers of 500 MW (el) and higher. There existplanned power reactor developments in several countries, including the USSR, for outputs up to 1000 and even to 2000 MW (el). Such an increase in the individual output of a single unit of an atomic power station is explained by the attempt to reduce the cost of electric power production by reducing the cost per installed kilowatt. However, excessive increase in the output of an individual unit is not always advisable. The associated power systems must surely have to be of very high capacity in order to avoid system "breakdown" in the case of an emergency situation at such a large station. At the present time, one can consider as a most acceptable output that of an atomic power unit with a capacity of 500-800 MW (el). In addition to the increase in the specific output of an atomic power station, the struggle for competitive posi- tion compels engineers and builders to look for means of reducing the cost of building and equipping an atomic power station. One of the steps in this direction is an increase in steam parameters. An improvement in the characteristics of the steam furnished the turbines makes it possible for an atomic power station to use mass-produced turbines of high output (200-300 MW and higher). This situation should play no small part in the improvement of the economic characteristics of atomic power stations. A big problem for physicists and atomic power scientists is the increase in the amount of average burnup of nuclear fuel. This also has a direct effect, in a favorable way, on the economic characteristics of atomic power station operation. No less important a problem, as was made clear at the conference, is the reduction in the cost of nuclear fuel (fuel elements). In a number of cases, they are so expensive that they amount to as much as 30 or even 50% of the cost of the entire power unit. All these problems, and many others, which determine the economics of atomic power station operation were well presented and discussed at the third Geneva conference. As already pointed out, the various countries which are carrying out development and planning of power reac- tors have chosen for themselves more or less definite types of reactors with which they proposed to solve the problem of electric power production. At the conference, the overwhelming majority of delegates was of the opinion that the future belonged to fast reactors since they, by their very nature, make it possible to produce more fissionable material than they themselves require, and in that way, increase the utilization of existing supplies of nuclear fuel by 10 or even 100 times because they bring U238 and thorium jnto the cycle. However, a number of delegates, chiefly those from the United States, feel that the development of nuclear power must go through three stages. The first of these is the stage which was reached in recent years and which is based on slow reactors (gas coolant and graphite moderator; heavy water cool- 1067 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 ant and moderator; light water coolant and moderator). The second stage is the use of improved reactors of the converter type. The third stage is the further development of reactor technology with the use of breeder reactors. A prolonged existence for the second stage is assumed by this group. The chairman of the American delegation, Prof. G. Seaborg, remarked in his report, "considering the number of shortcomings associated with the economic outlook for the development of breeder reactors, the construction of improved converter reactors would possibly seem extremely reasonable for many countries. . ." And, further, ". . . we shall never see the widespread use of breeder reactors if the technology for their production does not achieve an economically favorable stage." As is well 'known, the Soviet consensus with regard to the development of nuclear power supposes a more rapid transition to the construction of breeder reactors as the general trend in nuclear power although it is clear to us, of course, that fast breeder reactors, being the most promising of the new reactor types (keeping commercial applica- tion in mind), still require a great deal of creative engineering work. The Canadian atomic power scientists feel that breeder reactors will not be needed for many years. This is reflected in the long-range Canadian plans for the construction of atomic power stations. We, together with the representatives of Great Britain, France, Germany, and a number of other countries, cannot agree with this position. It should be noted that extensive work on the production of economical fast reactors is already being carried on in many countries. The Soviet Union has performed a large amount of research work with the BR-1 and BR-5 reactors and with the experimental setup at the BFS which has made it possible to proceed to the construction of a large commercial fast reactor with an output of 300-35'0 MW (el) in the Caspian Sea region. Great Britain has ac- cumulated a large deal of experience with the breeder reactor at Dounreay. The United States has recently pro- ceeded to the power mode for the EBR-II reactor. The operation of the Enrico Fermi atomic power station using fast neutrons will facilitate the accumulation of experience. The French Atomic Energy Commission is completing the construction of a fast reactor. Several other countries have also joined in these efforts. At the Geneva conference, the reports of this work were heard with a great deal of interest, and they evoked lively discussions. Along with work on large atomic stations, the results of work on the production of atomic power stations of low and medium capacities were discussed at the conference. Interest was expressed in the reports of Soviet scientists about constructiOn in the USSR of the modular atomic power station "Arbus" with an output of 750 kW containing an organic-cooled, organic-moderated reactor in which hydrostabilized diesel fuel acts as coolant. A distinguishing feature of this reactor is the special regenerative apparatus which maintains the coolant in a specified condition. Great interest was aroused at the conference by reports concerning the operation of the first experimental model of the transportable caterpillar driven TES-3 atomic power station with an electrical output of 1500 kW. This mobile power station has already been operating for three years and more at the Physics and Power Engineering In- stitute of the USSR State Committee for the Utilization of Atomic Energy. A report on the experience in the operation of the flagship of the USSR icebreaker fleet, the atomic ship "Lenin," was highly regarded by the specialists at the conference. As is well known, the first atomic icebreaker in the world, the "Lenin," has already successfully carried out five successive voyages, and has successfully coped with its difficult responsibilities in the rugged Arctic conditions of the icy northern seas; it has actively assisted our freighter fleet by guiding ships carrying lumber and other cargos. This unique ship has practically an unlimited sail- ing range without fuel loading, and sails in any Arctic region, breaking ice two to three meters thick. The atomic ship "Lenin" conclusively demonstrates all the advantages of using atomic energy in a commercial marine fleet. The American atomic passenger-freighter "Savannah," having this year made its first trip across the Atlantic to the shores of Europe, has also demonstrated its excellent seagoing capabilities. At the time of the conference, the USAEC organized a trip to Sweden aboard the atomic ship "Savannah" for representatives from the delegations of several countries. An acquaintance with the 'Savannah" reveals that it is a comfortable and cozy ship for ocean travel. At present, the "Savannah" is a demonstration ship propagandizing for the creation of an atomic freighter fleet. The "Savannah" only partly fulfills its function as a freight and passenger ship because of the lack of confi- dence in it by commercial firms and passengers. In comparison with the difficult, great, and honorable service ren- dered by our famed atomic icebreaker "Lenin,' the "Savannah" is being used insufficiently. The experience in operating the atomic ship "Lenin" and the discussions of the Soviet report at the Geneva conference indicated that the production of marine atomic installations in a commercial fleet, along with the de- velopment of nuclear power, is one of the promising trends for the use of atomic energy. 1068 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Several other countries are also working on the production of atomic ship installations. Thus, the atomic commercial cargo vessel "Otto Hahn," whose hull was launched recently, is being built in Germany. In Japan, which is a country of ship navigators and ship builders, they are proceeding to the development of a technology for manu- facture of ship reactors. In 1963, a special Japanese agency was created for the development of atomic ships which, in particular, was given the responsibility for the construction of an atomic oceanographic ship. A pressurized water reactor with an output of 35 MW will be installed in the ship. Construction of the ship should be completed in 1968. Thus, atomic steam-producing marine equipment is starting gradually to earn a place in the civilian commer- cial marine fleet. In every new field and particularly in such a rapidly developing one as the use of atomic energy, nuclear power engineering follows not only previously laid out paths but also completely new ones whose scientific develop- ment is only just beginning. For the first time in engineering history, there were reported at the third Geneva conference by Soviet scien- tists the practical results of research and experimental work on the direct conversion of nuclear energy to electricity. A high-temperature fast reactor with silicon-germanium converters which was built in the USSR at the I. V. Kur- chatov Atomic Energy Institute was brought up to power on August' 14, 1964, operated at a power of 500 W for about 2500 h, and continues to operate successfully. This apparatus bears the poetic name of "Rmashka" (Daisy). Inci- dentally, it reflects an actual situation since the highly enriched U235 dicarbide is disposed in the form of flower petals. The report of our scientists and a film on this new type of converter-reactor aroused great interest at the con- ference. Now, the Russian word "Romashka" appears in all specialized technical, and ordinary, publications, along with reports on the construction of a new form of power source without the use of any sort of auxiliary equipment in the form of turbines, pumps, etc. The manufacture of similar equipment, with a small number of moving parts, to be sure (pump, coolant), is being completed in the United States (SNAP-10A). It is intended for installation in space vehicles (satellites). American scientists read a report on this equipment and showed films of its construction. Apparently, it will be operating under ground-level conditions in a short time. There were also reports at the conference about low-power installations in which the heat source is a radio- active isotope. Ordinarily, they are very simple, their useful power is small (of the order of 5-100 W), and they operate with direct conversion of heat to electricity by means of semiconductors.j.Soviet scientists have reported on such equipment, in particular, on the "Beta-1" apparatus. The 13-radioactive isotope Ce144 is used in it, and it has now been in service for a year as the power supply for a standard automatic radiometeorological station provid- ing power for a transmitter with an output of about 150 W by means of an accumulator. Isotopic current sources for various purposes have been manufactured, and are being operated, in the the United States also. At the scientific and technical exhibition in Geneva, the Americans demonstrated a model of a barge- mounted meteorological station with a 60-W isotopic current source operating with Sr". Nuclear power holds a more distant prospect ?controlled thermonuclear fusion. The thorough and extremely complex scientific investigations have a completely clear and concrete purpose ?the creation of thermonuclear power stations. The reports which were presented and discussed at the conference indicate that, during the six years since the second Geneva conference, definite successes have been achieved in this field. Although scientists have not yet arrived at a complete solution of this important power problem, there have been successes in the area of finding the field. The Soviet section of the exhibits, which occupied 1000 m2, occupied a central position both with respect to location and to the large number of excellently executed models, exhibits, and photographs. Visitors and conference participants were particularly interested in models of the icebreaker "Lenin" and the Beloyarsk atomic power station, and two working thermonuclear laboratory machines from the Novosibirsk Nuclear Physics Institute. Quite good exhibits were arranged by other countries. The exhibits, on the whole, realized their purpose, and helped direct contact between scientists of various countries very successfully. LSome countries prepared interesting films on particular problems in the use of atomic energy. The Soviet Union presented 24 films which were shown continuously while the exhibit was open. 1069 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 The Geneva conference showed that science is an excellent means for intercourse between all the peoples of the world, particularly when there is a desire for it. The spirit of international cooperation was clearly evident at this conference. All the work of the conference was carried on in an efficient and friendly atmosphere. But, speaking of peace and friendship, of cooperation between scientists and specialists, between all the peo- ple of the world, we cannot forget that the military atom hinders the thorough development of the peaceful atom. This thought occupies the minds of many scientists and many progressive people in all countries. The peaceful use of nuclear fuel on a widespread scale is only possible when the path to thermonuclear war is completely closed to the military atom by the efforts of the people of the world. 1070 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 HIGH-TEMPERATURE REACTOR-CONVERTER "ROMASHKA"* M. D. Millionshchikov, L. V. Gorlov, Yu. D. G V. E. Ivanov, V. K. Ko N. E. Kukharkin, R. Ya Yu. A. Nechaev, B. S. E. N. Samarin, V. Ya. V. V. Yakovlev, M. V. and G. V. Kompaniets Translated from Atomnaya gnergiya, Vol. 17, No. 5, pp. 329-335, November, 1964 I. G. Gverdtsiteli, A. S. Abramov, ubanov, A. A. Efremov, V. F. Zhukov, vyrzin, E. A. Koptelov, V. G. Ko-i-ovskii, . Kucherov, S. P. Lalykin, V. I. Merkin, Pozdnyakov, N. N. Ponamarev-Stepnoi, Serov, V. A. Usov, V. G. Fedin, Yakutovich, V. A. Khodakov, The effective use of atomic energy in various kinds of apparatus is bound up with the quest for new methods of transforming energy and their study. The development of converters with the direct transformation of thermal into electrical energy is of special interest. One such system is the experimental power reactor "Romashka" in the I. V. Kurchatov Institute of Atomic Energy. This apparatus uses one of the most structurally-simple and operationally-reliable systems, based on a reactor and converter combined into a single unit, in which the heat generated in the active zone of the reactor is transferred to a thermoelectric converter situated at the outer surface of the reflector, by way of the thermal con- duction of the materials. The reactor uses fuel elements based on uranium carbide, which by virtue of its properties (high-working tem- perature and fairly high-thermal conductivity) is a promising material for this purpose. Good thermophysical and neutronophysical parameters are ensured for the reactor by using metallic beryllium as reflector material and graph- ite as structural material for the active zone. Use of these materials in the reactor makes is possible to employ a high-temperature converter based on semiconductors consisting of a silicon-germanium alloy. Description of the Reactor-Converter The nuclear reactor (Fig. 1) constitutes a neutron-physical system operating with fast neutrons. This serves as a source of thermal energy, which is converted into electrical energy by means of the thermoelements. The heat evolved as a result of the fission of U235 in the active zone of the reactor is transferred in the radial direction by thermal conduction to the reflector, and then from the side surface of this to the semiconductor converter *Report No. 873 presented by the USSR to the Third International Conference on the Peaceful Uses of Atomic En- ergy, Geneva, 1964. The converter "Romashka," the first nuclear-power system in the world with direct energy conversion, began delivering power on August 14, 1964. Since then the system has operated under optimum load with a temperature of 1770?C in the center of the active zone and 1000?C at the surface of the reactor. The electric current taken from the converter with thermoelement groups connected in parallel reaches 88 A. According to the situation on November 18, more than 1000 kWh electrical power have been produced. Tests have enabled us to draw one important conclusion regarding the characteristics of "Romashka," namely, the stability of its operation. Although automatic control is provided, it has not proved necessary to set this in motion, since the system requires little regulating. After setting in the nominal condition, the power and tempera- ture are maintained to a high accuracy at the assigned level on account of the self-regulation of the reactor, which has a negative temperature coefficient of reactivity. The results of the tests confirm the high efficiency of the silicon-germanium-alloy semiconductor thermoelements in the radiation fields of a fast-neutron reactor. After 2500 h operation, changes were found in the main characteristics of the thermoelements. Neither the electrical power nor the emf of the converter had changed appreciably. The results of the tests are being studied. These should lead to a series of important conclusions for use in future investigations and the development of new converters. 1071 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Fig. 1. General view of the reactor. mechanism, electrically driven. The drive mechanism adjacent thereto. The reactor is of cylindrical form, comprising an active zone and a reflector; it is dis- posed vertically. The active zone of the reactor is made up of fuel elements, each of which comprises a graphite body and uranium dicarbide plates with 90% enrichment with respect to U235. The total weight of U235 in the active zone is 49 kg. The radial reflector is compos-,d of coaxially-disposed beryllium and graphite elements. The end reflectors are made of metallic beryllium. To reduce the leakage of heat through the ends of the reac- tor, thermal insulation is applied. The control system of the reactor consists of four rods, set in the radial beryllium reflector, and the lower end reflector. An AC (automatic control) rod, con- sisting of beryllium and beryllium oxide in a stainless- steel shell, effects the automatic control of the reactor. Hand control is effected by the motion of HC (hand control) rods comprising a scattering section based on beryllium oxide and an absorbing section based on a boron-containing alloy. The temperature effect is compensated by motion of the lower end reflector. For emergency protection of the reactor, two scram rods (SR) are used; these are set in the radial reflector and the lower end reflector. In construction the SR are similar to the HC rods. All the control devices, apart from the automatic-control AC rod, are driven by a hydraulic system. The AC rod is moved by a servo- for the control and protection devices is disposed under- neath the reactor body. The thermoelectric converter used in the system comprises thermoelements based on a germanium-silicon alloy. The thermoelements comprise two thermopiles with n- and p-type conductivity connected on the hot side by a switching plate. On the cold side the individual couples of the thermopiles are switched into a single circuit. The whole thermoelectric converter is divided into four groups of thermoelements, each of which has independent power leads. Thus the construction of the converter part of the system enables us to study the characteristics of individual groups as well as the whole converter for either series or parallel connection of the groups. Inside each of the four groups of the converter, the thermoelements are switched in series into four parallel circuits. The general view of the converter appears in Fig. 2. Below we present the main parameters of the reactor-converter: Thermal-Energy Parameters of the System Electrical power, kW 1 0.50-0.80 Total thermal power, kW 2 40 Maximum temperature of beryllium reflector, ?C 1200 Maximum temperature of outer surface of Be reflector, ?C 980 Mean temperature of base of radiating fins, ?C 550 Maximum temperature of UC2 fuel elements, ?C 1900 Neutron-Physical Characteristics of the Reactor Charge of U235, kg 49 1Depending on the temperature conditions. 2Allowing for leakage. 1072 IDeclassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Efficiency, /o: AC rods 0.2 HC rods 0.4 Scram rods 0.4 All control rods 1.4 Movable end reflector. 3.5 Total neutron flux, neutrons/cm2. sec: In center of active zone 10" At active-zone boundary 7. 1012 Leakage of neutrons from reactor, neutrons/cm2 ? sec 3 ? 10" Choice of Parameters for the System The power potentialities of a reactor-converter without a heat-carrier are determined by the limiting char- acteristics of the materials used, the dimensions of its main elements and their structural arrangement. The close interconnection between these parameters demanded the execution of a wide range of theoretical computations and experimental work on thermophysical and neutron-physical aspects and properties of matter, all directed towards finding the optimum characteristics of the apparatus and a basis for the efficiency of its components. Thermal -Energy Calculations. The electrical power of the system is ultimately determined by the thermal conditions of the converter, its structural parameters, and the physical properties of the materials. An ex- pression describing the electrical power may be written in the form rs S 2X3S3 1(Q1+ Q2) s (xi +x2) 2v3,2 r- . TH I m+1 TH?TC 2 (M+1)2 I 1 -q nms (Qt + Q2) (x1+ x2) 71s E i=1 h=1 (xi +x2)2 (Qt i -I-Q2i) where Qi is the thermal flux through the i-th zone, QT the total thermal flux through the converter, x the thermal conductivity, p the specific electrical resistance, a the thermo-emf coefficient,/ the length of a thermopile, s the Fig. 2. General view of thermoelectric converter. 1073 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 cross section of the thermopiles, s3 the cross section of the thermal insulation between the piles, ns the number of zones, nz the number of series-connected couples, m the number of parallel-connected couples, TH and TC the tem- peratures of the hot and cold junctions of the semiconductors, rs the switching resistance of one thermocouple, and M the ratio of the external-load resistance to the internal resistance of the converter. (Indices 1 and 2 relate to semiconducting material of the n and p types; i or k gives the zone numbers.) Nonuniformity in the distribution of temperatures and thermal flux over the outer surface of the radial reflec- tor is taken into account in this formula in an approximate fashion by dividing the converter vertically into annular zones, the thermal conditions being considered constant over each. The influence of nonuniformity in the Joule and Peltier effects is neglected. A characteristic of the system is the fact that the operating conditions of the converter, and hence also its electrical power output, are determined by the attainable level of temperature in the individual elements of thP reactor and converter elements and the possibility of heat loss by radiation. In view of this, it was necessary, in order to determine the power parameters of the system, to make a thermal calculation of the reactor-converter system as a whole. The problem of determining the temperature in the active zone, the reflector, and the converter, reduces to solving the heat-conduction equations in a multizone system with nonlinear boundary conditions, describing the transfer of heat by radiation. Numerical solution of these equations was effected on an electronic computer. The distribution of temperature in the radial reflector for one operating condition is shown in Fig. 3. The elements of the active zone of the reactor are in stressed conditions both with respect to temperature level and the temperature drops determining the thermal stresses. In this connection we solved the problem of the influ- ence of possible breaks in the continuity of the fuel element on the temperature rise in the active zone. This was done by electrically simulating the temperature fields on conducting paper. In the system considered, the heat which has passed through the converter is carried away by radiation. The maximum heat removal from the surface, for a given mean temperature of the cold layers of the converter, secures the greatest electrical power, other conditions being equal. In order to find the optimum form of the radiating sur- face (number of fins, size, shape), we solved a system of integrodifferential equations describing the distribution of temperature in the fins, allowing for the mutual irradiation of the elements and for thermal conduction. Figure 4 shows the heat removal as a function of the weight and number of fins. Allowing for the results of the thermal calculations of the system, the electrical power was determined as a function of the thermal power passing through the converter on varying the quality factor of the thermoelement. Neutron Physical- Calculations. The neutron-physical characteristics of the reactor were calcu- lated on an electronic computer using the multigroup method of statistical tests (Monte-Carlo method). Use of this method in the present case enabled us reliably to allow for the geometric and physical features of the system asso- ciated with the heterogeneous structure of the active zaie, the presence of channels and gaps of complex configura- tion, the sharply inhomogeneous physical properties of the materials of the active zone and the reflector, the speci- fic system of reactor control, and so forth. In the calculations we used the multigroup (21 groups) system of con- stants which allows for the resonance structure of the U238 cross section, the (n, 2n) reaction for beryllium, and in- elastic transitions in the first nine groups. In the course of the calculation, some50,000neutronhistories were traced. Height Fig. 3. Distribution of temperature over the cross section of the radial reflector. 1074 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 30000 25 000 2200030 50 70 100 Fin weight, kg 130 Fig. 4. Variation of heat removal with fin weight (tem- perature at the base of the fins 600?C): 1) 6 fins; 2) 9 fins; 3) 18 fins; 4) 36 fins. W, sg/cm2.h /0- 10 20 60 100 140 180 220 260 300 340 35'0 420 460 Time, h Fig. 5. Evaporation rate w of uranium dicarbide in an inert medium at t = 2000?C (different points, different samples). Data of L. K. Mizrakhi and Yu. M. Utkin. A, kcal/m ? h? ?C 25 20 15 10 111111.111111MI 11111111"11 apilimmemi , MVO niztImmataussmass matamaammimmomm mommommommin 200 400 600 800 1000 1200 1400 Temperature,?C Fig. 6. Thermal conductivity of uranium carbide as a function of temperature (different points, different sam- ples). Data of A. G. Kharlamov. 1600 1800 Experimental Study of the Charac- teristics of Elements of the System. In order to gain a basis for the projected parameters of the system, we made experimental thermophysical and metallophysical studies of its materials and com- ponents. We studied the contact interaction of the uran- ium dicarbide with graphite and the volatility of uranium dicarbide in an inert medium and in vacuum at temperatures up to 2000?C (Fig. 5). We examined the temperature dependence of the thermal conduc- tivity of uranium dicarbide (Fig. 6), the linear-expan- sion coefficient, and other characteristics, over a wide temperature range. Together with a study of the thermal-strength characteristics of uranium dicarbide, tests on mockup fuel elements, and loop tests of uranium-carbide samples, these demonstrated the ef- ficiency of the fuel elements in working conditions. The use of a beryllium reflector, working at large thermal fluxes in a temperature range close to the melting point, in the reactor demanded an ex- perimental study of the interaction between metallic beryllium and various structural materials, the ther- mal conductivity of beryllium, and its deformability and thermal strength. In order to reduce the heat leakage through the ends of the reactor and between the thermoelements of the converter, high-temperature thermal insulation was used in the system. In connection with this, we studied the thermal conductivity of the thermal in- sulation in various media at working temperature. One of the important aspects of the investiga- tion was the efficiency of the thermoelectric-con- verter elements in neutron and y-radiation fluxes. Repeated many-hour tests of the thermoelements in the loops of the RFT (Physical and Technical Re- search) reactor at total neutron fluxes of 3.1019 ther- malneutrons/cm2 led to the conclusion that the main properties of the thermoelements varied within per- missible limits (Fig. 7). Test-Bed Studies of the Neutron- Physical and Thermal-Power Characteristics of the System Study of Neutron-Physical Charac- teristics. Five different arrangements were setup, differing in the concentration of the fissile material. For each arrangement we made a series of investiga- tions embracing a wide range of questions: the de- pendence of the critical loadings on the composition of the active zone, the efficiency of the reflectors 1075 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 900 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 emf,V 1,8 800 I ;,6 700 600 SOO 400 300 200 100 2 3\ 4 1,4 1,2 1,0 0,8 0,6 0,4 0,2 0 20 40 60 80 100 120 140 160 180 200 220 240 Irradiation time, h Fig. 7. Variation in the main parameters of the semi- conductor thermoelements during reactor irradiation (maximum total neutron flux ",3.1019neutron/cm2). Data of G. M. Pavlov. 1) Temperature of hot junction; 2) thermo-emf of section; 3) temperature drop in ther- moelements; 4) reactor power. 0 Fig. 8. radius and height of the active zone (Rand II= radius of ac- tive zone; rand z are coordinates). Data of A. M. Krutov. Relative distribution of heat liberation over the and control mechanisms, the distribution of heat evolution in the active zone, the effect of structural gaps on the reactivity, and so forth. Great attention was paid to studying the effect of moving the lower end reflector and shaping the active zone on the neutron-physical characteristics of the reactor system, and also to examining the efficiency of the control rods and the heat liberation fields. Some of the results obtained appear in Fig. 8, which shows the distribution of heat liberation over the radius and height of the active zone. The reactivity in all these cases was measured by dif- ferent methods: from the acceleration period, and-pulse and integral methods. Comparison between the results of measuring activity by different methods enabled us to estimate the effec- tiveness of delayed photoneutrons due to the presence of the beryllium reflector. It was established that in reactors of this type photoneutrons are practically absent, and that for analyzing the experimental results the characteristics of six groups of delayed neutrons may be used. Study of the Thermal-Energy Characteristics. The final stage of testing the reactor-con- verter on a full-scale test-bed was preceded by complex tests of a full-scale thermal model of the reactor-converter on an electric-heating bed. The aim of these tests was to check the efficiency of the whole system and its individual components, and also to study the working parameters in the steady and nonsteady states. During the tests, the temperature fields in various elements of the system were constantly measured. For this purpose, 53 tungsten-rhenium and 86 Chromel-Alumel thermocouples were set in the reactor and converter. The electrical characteristics of the converter were measured by means of a special electric panel, which made it possible, first, for each of the four groups of thermocouples, to vary the load smoothly from 0.1 to 10 t2 and make measurements of emf, short-circuit current, working current, and voltage, and secondly, to make electrical measurements not only separately in the groups but also with the groups connected in series or parallel. The elec- trical power of the converter was determined in the maximum-power condition. In the nominal condition the ap- paratus was tested more than 1000 h. Analysis of the results of testing the system led to the conclusion that all the main components of the reactor- converter were very efficient. The characteristics of the system were quite stable over the whole testing period. The thermo-emf of the converter remained practically constant over the whole test. Toward the end of the testing, 1076 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 a certain rise in the internal resistance of the converter was noted, in association with which the electrical power taken from the converter (in the maximum-power condition) fell on average by 100/0. The tests made enabled us to study and demonstrate the efficiency of the active zone, the reflector, and the converter in operating conditions. - LDeclassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 1077 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 DEVELOPMENT OF SUPERHEATING POWER REACTORS OF THE BELOYARSK NUCLEAR POWER STATION TYPE* N. A. Dollezhal', I. Ya. Emel'yanov, P. I. Aleshchenkov, A. D. Zhirnov, G. A. Zvereva, N. G. Morgunov, Yu. I. Mityaev, G. D. Knyazeva, K. A. Kryukov, V. N. Smolin, L. I. Lunina, V. I. Kononov, and V. A. Petrov Translated from Atomnaya gnergiya, Vol. 17, No. 5, pp. 335-344, November, 1964 Introduction Ever increasing needs for electrical power have promoted an increase in the rates of development of energy production. It was found to be most profitable to construct installations with a high unit out. At present in the Soviet Union, turbounits, which operate on various steam parameters, are planned and manu- factured (Table 1). TABLE 1. Steam Parameters and Outputs of As can be seen from Table 1 an increase in Turbogenerators, in Production and Planned, output of the turbounit is associated with an increase within the Soviet Union of temperature and pressure of the steam used. Steam parameters prior to turbine Electrical output, MW pressure, atm. temperature ,?C 90 535 100 130 565/565 105, 200 240 580/565 300, 500, 800,1000 Progress in nuclear power development, as well as in conventional power engineering, is based on construction of nuclear power stations with a high- unit output. It is therefore advantageous to install power reactors which will ensure the generation of steam with high parameters, thus enabling the high- output turbounits, which are being manufactured and developed, to be used and which will ensure a high efficiency for the nuclear power station. Uranium-graphite channel-type uncased reactors meet the stated requirements completely. These reactors are similar to the superheating nuclear steam reactors of the Beloyarsk Nuclear Power Station. A channel-type re- actor has been operated successfully in the First Nuclear Power Station in the USSR [1]. Of the two Beloyarsk reac- tors, the first is at the stage of startup with the planned parameters and the second is at the assembly stage. A description of the design of the first reactor was presented in a report at the Second International Conference on the Peaceful Uses of Atomic Energy [2], and a report is devoted to its startup and adjustment at the present Conference [3]. The steam parameters, generated by the Beloyarsk nuclear power station reactors, are 90 atm and 500?C. Planning studies show that the most advantageous development for the reactors being considered is the transi- tion from high-steam parameters directly to the use of supercritical parameters with a straight-through system of coolant utilization. On the contrary, as will be shown below, the use of turbot:nits at a pressure of 130 atm and with intermediate steam superheating is less advantageous. Technological Flowsheet Development The choice of a proper flowsheet is of importance when constructing high-power plants. The principal cri- teria in choosing nuclear power station designs, just as in conventional thermal power installations, are reliability, simplicity, and efficiency. In addition, a particular demand is imposed by the leak-tightness of the loops in the nuclear power station water cycle. The correct choice of the technological flowsheet, in conjunction with the constructive solution of problems associated with the temperature operating regime of the channels, allows its output to be increased considerably * Paper No. 309 ,presented by the USSR at the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1964. 1078 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 without increasing the over-all size of the reactor, just as was achieved in the second Beloyarsk nuclear power station reactor, whose output was increased by a factor of two compared with the first reactor. Figure 1 shows several possible technological flowsheets for uranium-graphite reactors of the channel type. Figure la represents the flowsheet of the first Beloyarsk NPS reactor with an electrical output of 100 MW. Fig- ures lb and lc show the flowsheet of the second Beloyarsk NPS reactor with and without steam scrubbing. These de- signs, with certain structural changes in the channels, have enabled the electrical output of the reactor to be in- creased to 200 MW. Figures id-lg show possible designs for operating a reactor installation with supercritical steam parameters, thus allowing its power to be increased to 800-1000 MW. These designs are considered in detail below. Naturally, the straight-through design of reactor unit is considerably more simple and inexpensive than the multi- loop design, which requires additional cumbersome and expensive heat exchangers, separators, circulating pumps with the framework for their conduits and fixtures, and also higher coolant parameters. By the use of a design with nuclear steam superheating and especially a straight-through design, radiation safety of the maintenance sections of the station and in particular the turbounits is ensured. From this point of view it is impossible not to notice the ad- vantages of the Beloyarsk NPS reactors. By using reactors of this type, the radiation environment in the technological flowsheet rooms is determined by the salt and oxygen activity of the coolant, since the tubular design of the fuel elements in the event of damage excludes the possibility of fission fragments getting into the coolant. Tests carried out on the First Nuclear Power Station have shown that from the point of view of activity the use of superheated steam in the reactor is entirely possible under conditions of good quality water in the separators and reliable separation of the steam produced [4]. These requirements are guaranteed relatively simply in the first Beloyarsk NPS reactor. In the second reactor, be- cause of the repeated circulation of the water, a buildup of salt activity is possible in the steam separators and also its subsequent introduction by the steam into the steam superheating channels and into the turbine. Consequently, in the second reactor provision has been made for specially careful cleaning of the circuit to remove salt and corrosion products: Ionite filters are installed after the turbine condenser, the purging of water from the steam separators has been increased and increased demands are also imposed on the quality of the makeup water and the compactness of the tubular plates of the condenser. The measures ennumerated are necessary to an even greater degree as a result of using straight-through systems. The increase of output of the reactor above 200-300 MW with conservation of its dimensions and superheated steam pressures of 90 or 100 atm is not sufficiently economical, since in this case either the most efficient operating circuit of one reactor with a single turbine (or as a last resort, with two) must be rejected or the turbine must be used with intermediate steam heating. In addition to the complexity of the installation and maintenance of the power station, in this case the size of the conduits is increased, additional heat exchangers emerge, the separators are con- siderably increased in cost and the resistance of the multicirculation circuit is increased. The increase of the sur- face area of the intermediate superheaters leads not only to their increased cost but also to an increase in the con- tent of corrosion products in the circuit, which causes deterioration of the water cycle of the installation. The re- actor power may be significantly increased by transition to supercritical steam parameters. In this case, the tech- nological layout of the installation is considerably simplified since the necessity for coolant and circulatory pumps in the multicirculation loop is removed and also separators are eliminated. By using a coolant of supercritical pa- rameters the limitations associated with crises in heat transfer and with the appearance of hydrodynamic instabili- ties in the operating channels are removed. Figure id shows one of the possible schematic layouts for a nuclear power station operating on supercritical steam parameters. Feed water from the de-aerator is fed to the channels of the first group, producing superheated steam, part of which is fed to the external intermediate superheater and the other part is led off via a throttling unit to the main steam pipe and is then directed into the turbine. From the intermediate superheater, steam at supercritical pressure is directed into the second group of reactor operating channels and is superheated in them to a specified temperature and also enters the turbine. In order to increase the reliability of operation of the equip- ment in the transition regimes, a temperature regulator has been provided after the intermediate steam superheater. Figure le shows the layout of an energy generating installation similar to the one being considered but with- out high-pressure preheating, which can be excluded to reduce corrosion products in the circuit. As a result of giving up high-pressure regeneration the efficiency of the unit is reduced by approximately 1.5% in comparison with the efficiency of the unit operating on the scheme depicted in Fig. id, but in this case the water cycle of the sta- tion is considerably improved. 1079 Am=m? Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 c) g) 1080 b) Fig. 1. Possible designs of nuclear power reactors with uranium-graphite reactors of the channel type and nuclear steam superheating: 1) reactor; 2) separator; 3) steam generator preheater; 4) steam generator evaporator; 5) circulating pump; 6) feed pump; 7) turbine; 8) inter- mediate superheater; 9) feed pump; 10) throttling unit; 11) water injection; 12) condenser decontamination; 13,14) regenerative preheaters for low and high pressure. Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Figure if shows an arrangement with a turbine which has two intermediate steam superheaters. In this arrange- ment,the requirement for steam superheating in the throttling section after the first group of operating channels is eliminated. All the steam passes from the intermediate superheaters to the second group of channels and then to the turbine. The most promising arrangement for a nuclear power station with supercritical parameters is the once-through arrangement with a single group of operating channels and without intermediate steam superheating (see Fig. 1g). The turbounit used in this arrangement consists of three stages: The first stage uses steam with supercritical param- eters and the two subsequent stages operate on saturated steam. A separator is necessary for removing moisture from the steam at the entrance to the low-pressure cylinder. Of course, the arrangements which have been considered for uranium-graphite reactors of the channel type and with supercritical steam parameters do not exhaust all the possibilities for developing nuclear power stations in this direction. The increase in unit power and the correct choice of flowsheet of the power generation equipment have a considerable effect on the efficiency index of a nuclear power station, firstly on the cost per kilowatt. The reactor output is directly connected with the operating cycles of the evaporative and steam superheating channels which will be considered below. Operating Cycle of the Evaporative Channels A safe operating cycle for the evaporative channels can be accomplished by ensuring reliable and trouble-free cooling of the fuel elements and the elimination of interchannel and intrachannel fluctuations in the coolant supply. In designing the first reactor for the Beloyarsk Nuclear Power Station, experiments were carried out to study water boiling regimes in small diameter tubes [2]. For the second reactor, because of the increase in power, the internal diameter of the fuel element tubes was increased from 8.2 to 10.8 mm, which necessitated additional tests to determine the possibility of crisis-free boiling of water in tubes of small diameter. The experiments were carried out in an electrically heated arrangement of tubes of 10.4 mm diameter and 3.8 ?m length. The relationship between the steam content at the site of origination of a heat transfer crisis and the mass velocity for various thermal fluxes and a pressure of 150 atm are shown in Fig. 2. A heat transfer crisis for the experiments was organized by a stepwise increase of temperature of the tube wall. Investigations, carried out at different pressures and identical thermal loadings, steam content, and coolant mass flow velocities, showed that the temperature increase of the wall as a result of a crisis is higher, the lower the pressure of the coolant. At the same time, it was established that by reducing the pressure of the coolant the critical steam content increases. In addition, a test rig simulating the technological flowsheet of the second Beloyarsk nuclear power station reactor and electrically heated channels were used to investigate the hydrodynamic stability of coolant flow rate along parallel channel tubes in the boiling regime at pressures of 20-150 atm; the coolant flow rate through the channel was 500-5000 kg/h and the channel power was 50-800 kW. Pulsations of the flow rate were observed in the experiments, in the region of low (up to 5 wt./o) and high (above 40 wt./ o) steam content. In the latter case, the pulsations ceased with increase of flow rate of pressure. Thus, with a coolant feed of about 1500 kg/h, a pressure of 40 atm and a power of 400 kW the pulsations in the coolant feed began at a steam content of about 40 wt./o, and for a pressure of 60 atm and the same conditions the pulsations began at a steam content of 80 wt./o. It followed from the experiments that the pulsations in the coolant supply in the region of high-steam con- tent do not present a danger for reactors of the Beloyarsk nuclear power station type, since the nominal pressure in the evaporative loop is not less than 90 atm and the steam content at the channel output is not greater than 35 wt./o. Figure 3 shows the experimental curves obtained on the test rig and which delimit the region of hydrodynamic stability of the interchannel coolant flowrates and the region of interchannel pulsations originating in the evapora- tive channels with low steam contents of the coolant. The curves are drawn as functions of the flow rate and pres- sure of the coolant for a different power of the evaporative channels. The curves shown in Fig. 4 also define the regions of stable and unstable interchannel coolant flow rates, but as a function of various flow rates and steam con- tents of the coolant at the channel output as a result of varying its power from 50 to 800 kW. In plotting the curves, 1081 Ami. Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 0,6 ? ? T? ? , ? ? Al 102 ; 0 ? ? 0 ? ? ? ? . 3 7, . . )C...."......X......... - + + 0 . . t X 1000 2000 3000 Coolant mass flow velocity, kg/m2. sec Fig. 2. Relationship between critical steam content by weight and the coolant mass flow rate at constant specific thermal loading for the following thermal flux values, kcal/m2? h: 1)0.31 ? 106; 2) 0.45 ? 106; 3) 0.64- 106; 4) 0.94.106; 5) 1.28 ? 106; 6) 1.62 ? 106. 4000 5000 150 ? ? ? ? 0 ? 0 1000 20005'000 5000 Coolant supply through channel, kg/11 Fig. 3. Curves, delimiting regions of stable and unstable interchannel coolant flow rates as a function of the pressure and flow rate of coolant through the channel for different power outputs, kW: 1) 800; 2) 400; 3) 300; 4) 200;5) 100; 6) 50. The black points correspond to regimes with- out pulsation. it was observed that their position is almost independent of the coolant pressure. The regions of stable channel flow rates are found to be higher, in Fig. 3 as well as in Fig, 4, than the corresponding curves for the evaporating channel. It follows from Figs. 3 and 4 that with increase in the channel power the stable flow-rate region decreases. 1082 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 TABLE 2. Evaporative Channel Characteristics Characteristic Reactor of the First Nuclear Power Station First reactor of the Beloyarsk Nuclear Power Station Second reactor of the Beloyarsk Nuclear Power Station I zone II zone III zone IV zone V zone Channel power, kW 300 405 771 634 617 545 517 Coolant supply through channel, kg/h . . . 2500 2400 5500 4700 4150 3550 3250 Steam content at exit from reactor, % . . . - 33.6 27.6 29.3 30.5 32.1 34.2 Pressure at inlet to channel, atm 100 155 155 155 155 155 155 Pressure at exit from channel, atm 98 150 145 145 145 145 145 Coolant temperature at channel inlet,?C. 200 300 303 303 303 303 303 Coolant temperature at exit from channel, ?C 290 335 338 338 338 338 338 Maximum thermal loading, kcal/m2 -la ? 10-6 1.8 0.5 0.8 0.7 0.6 0.5 0.5 Circulation velocity, m/sec 4 3.5 4.6 4.0 3.5 3.0 2.7 Maximum temperature, ?C: of tube inner wall 324 355 365 of nuclear fuel 382 400 415 Minimum reserve prior to critical thermal loading - 2 1.85 1.9 i 1.9 2.0 1.95 TABLE 3. Characteristics of the Steam Superheating Channels Characteristic First reactor of Beloyarsk NPS Second reactor of Beloyarsk NPS descending fuel elements ascending fuel elements Maximum channel power, kW 368 767 767 Minimum channel power, kW 202 548 548 Steam supply through channel with maximum power, kg/h 1900 3600 3600 Steam supply through channel with minimum power, kg/h 1040 2570 2570 Pressure at channel inlet, atm 110 132 124 Pressure at channel outlet, atm 100 125 110 Channel inlet temperature of steam, ?C 316 328 397 Channel outlet temperature of steam, ?C 510 399 508 Maximum thermal loading, kcal/m2. h? 10-6 0.48 0.82 0.68 Maximum steam velocity, m/sec 57 76 112 Maximum temperature, ?C: of tube inner wall 530 426 531 of nuclear fuel 550 482 565 of graphite 725 - 735 As a result of the origination of interchannel coolant flow rate pulsations, fluctuations of tube temperatures were observed, which coincided in frequency with the flow fluctuations. With increase of pressure, the amplitude of the temperature and coolant flow rate fluctuations was reduced. This same effect was observed by reducing the thermal loading or increasing the supply of coolant through the channels. Thus, at a coolant pressure of 50 atm and a channel power of 200 kW, the amplitude of the tube temperature fluctuations with a coolant feed of 1000 kg/h amounts to 65?C, and with a coolant feed of 1500 kg/h it is only 30?C. The experiments carried out to determine the conditions of origination of heat exchange crises and coolant feed pulsations have enabled the operating regimes of the evaporative channels to be chosen correctly as far as their operating conditions and their possibilities are concerned. The distribution of coolant supplies in this reactor is not accomplished proportionately with the channel power, but such that identical reserves are ensured prior to a thermal 1083 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 20 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 1000 2000 3000 4000 5000 Coolant supply through channel, kg/h Fig. 4. Averaged curves, delimiting regions of stable and unstable interchannel flow rates as a function of steam con- tent by weight and the flow of coolant through the channel at various power outputs, kW: 1) 800; 2) 400; 3) 300; 4)200; 5)100; 6) 50. (The curves are plotted according to the re- sult? of steam content determination corresponding to re- gimes without pulsation, for different pressures and flow rates of coolant.) ? 0 0 0 0 3 -?o -e 5 6 7 - 0 8-n Fig. 6. Diagram of channel arrangement: 1) operating evaporative channels (732); 2) operating steam super- heating channels (266); 3) compensating rod channels (80); 4) channels for shut-off rods (16); 5) automatic control rod channels (4); 6) channels for counting chambers (2); 7) channels for triggering chambers (4); 8) ionization chamber channels (30). 1084 Outlet' fillet Fig. 5. U-shaped steam superheating channel: 1) upper cap; 2) seal- ing rings; 3) tempera- ture expansion com- pensator; 4) absorbing. rod; 5) lower cap; 6) ascending fuel ele- ment; 7) descending fuel element 0,4 0 50 100 150 200 250 300 350 400 450 'Reactor radius, cm Fig. 7. Thermal neutron distribution along the reactor radius. excursion in the most dangerous section of the channel, taking account of the change of the steam content and the thermal loading throughout the channel height. This coolant supply distribution permits the average steam content at the exit from the reactor to be somewhat in- Creased. Thus, for a steam content by weight at the exit from the most heavily loaded central evaporative channels of 27%, the steam content at the exit from the Declassified and Approved For Release 2013/03/14 : CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 1,0 2 1 100 200 300 400 600 500 700 800 Reactor height, cm Fig. 8. Thermal neutron distribution through- out the height of the reactor. 1) At the start of the running period; 2) at the end of the running period. ? peripheral channels can be increased to 35%, which permits an average steam content of 31% to be produced at the exit from the reactor. In order to reduce the thermal loadings and to decrease the hydraulic resistance of the loop, fuel elements are used in the evaporative channels of the second Beloyarsk reactor which have internal tubes of larger diameter than the tubes of the evaporative elements of the first reactor. For an identical ex- ternal diameter of 20 mm, the dimensions of the inner tubes of the evaporative fuel elements of the first reactor are 9.4x0.6 mm, and for the second reactor 12 x 0.6 min. In the channels of the second reactor, the diameter of the central tube, through which the coolant supply is effected, has also been increased somewhat. In the remainder of the evaporative channel con- struction, the first and second Beloyarsk reactors are identical. Uranium-molybdenum alloy is used as nuclear fuel in the evap- orative channels of both reactors, with a magnesium filler for obtaining thermal contact. The evaporative channel charac- teristics are shown in Table 2. Operating Regimes of the Steam Superheating Channels The steam superheating channels are the most highly stressed elements of reactors with nuclear steam super- heating, since they are operated at temperature conditions which are very high in comparison with the evaporative channels. The high operating temperatures of the steam superheating channels limit the choice of nuclear fuel and structural materials. The design of the steam superheating channels in the Beloyarsk reactors differs somewhat from the design of the evaporative channels, although fuel elements of almost the same dimensions as in the evaporative channels are used here. The initial design of the steam superheating channels specified to the supply of steam through the central tube of the channel and it was heated by motion through six fuel elements. Subsequently, the so-called U-shaped design of the steam superheating channels was developed (Fig. 5). Its distinctive special feature consists in that it ensures sequential heating of the steam by its motion initially through three fuel elements downwards and then through the next three fuel elements upwards. The U-shaped design of the channels delimits the operating temperature conditions of the descending and ascending fuel elements and consequently permits the use of simpler down-leading elements and elements which are cheaper to manufacture than the evaporative channels. The use of channels of U-shaped design with sequential steam superheating also makes it possible to reduce the temperature of the graphite masonry of the reactor. Thus, as a result of transition from the initial design of the steam superheating channels to the U-shaped design, the tem- perature of the graphite block is reduced approximately by 100?C (with a channel power of 360 kW). Reduction of the temperature of the graphite is ensured by the fact that heat in the U-shaped channels, which is extracted from the graphite block is led off to the down-leading fuel elements which have a relatively low temperature. Reduction of the graphite temperature has a favorable effect on the operating conditions of the graphite as well as on the phys- ical characteristics of the reactor which are improved somewhat as a result of reducing the temperature of the neu- tron gas. Finally, as a result of transition to the U-shaped steam superheating channels the central tube is eliminated which reduces the quantity in the active zone. In place of the central tube an absorbing rod with fine control, al- lowing the power of the operating channels to be trimmed within known limits is installed. The design of the as- cending steam superheating elements and the fuel composition, based on uranium dioxide, guarantees heating of the steam to 500?C, which was confirmed by loop experiments in the reactor of the first nuclear power station [4]. Table 3 shows the characteristics of the steam superheating channels in the Beloyarsk Nuclear Power Station reactors. The steam superheating channels of U-shaped design are a species of multipass channels, in which the coolant passes successively through a number of fuel elements. The sequential motion of the coolant ensures different oper- ating temperature conditions for the fuel elements and permits these conditions to be varied within specified limits by the appropriate choice of the power of sequentially engaged elements. Thus, in the multipass channels the num- ber of the most important and complex fuel elements, located only at the outlet section of the coolant flow is con- 1085 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 siderably reduced. In the remaining sections of the coolant flow, simpler and more cheaply produced elements can be used. The most complete advantages of multipass channels are manifested as a result of transition to the once- through arrangement with supercritical steam parameters. Arrangement of the Operating Channels in the Reactor Because operating channels of two types (evaporative and steam superheating) are used in the reactors of the Beloyarsk Nuclear Power Station, it is appropriate to discuss the question of their optimum disposition in the active zone. For the startup and cooling regimes of the reactor, the most favorable arrangement is the composite arrange- ment of channels of different type, since this ensures the return flow of heat from the steam superheating channels to the evaporative channels, which facilitates the operating conditions of these channels in these regimes. The arrangement of the steam superheating channels on the periphery of the active zone increases their number, but in return it reduces the loss of pressure in the steam circuit. The operating temperature conditions of the fuel elements and the graphite block depend weakly on the place of insertion of the channel. Thus, by installing the steam super- heating channels of the first reactor of the Beloyarsk Nuclear Power Station in the center or at the periphery of the active zone, the maximum fuel element temperature should differ by only 20?C and the maximum temperature of the graphite block by 60?C. In the second reactor of the Beloyarsk Nuclear Power Station a patterned central arrangement of the steam superheating channels has been adopted (Fig. 6), which with the use of fine control rods makes possible a balanced distribution of the energy release along the reactor radius. Nonuniformity of distribution of the energy release along the radius of the second reactor has been reduced to 1.3, while in the first reactor, it is 1.4. Figure 7 shows the ther- mal neutron distribution along the radius of the second reactor, which determines the energy release and which quite closely corresponds to the distribution established at the end of the running period because of nonuniformity of fuel burnup. Figure 8 shows the thermal neutron distribution throughout the height of the second Beloyarsk NPS reactor. The kink in the distribution curve at the end of the running period is due to nonuniformity of fuel burnup. In reactors with supercritical coolant parameters and the once-through technological circuit, in which all the channels operate in identical regimes, the required distribution of energy release can be ensured by the appropriate spacing of the control rods and of the burnup absorber. Conclusions Some possibilities have been discussed for developing uranium-graphite power reactors of the Beloyarsk Nuclear Power Station type with nuclear steam superheat. Thus, the transition from the double loop technological system of the first Beloyarsk reactor to a single loop system and some change in the design of the operating channels has en- abled the power of the second reactor to be increased to 200 MW [5]. Even greater possibilities for reactors of this type are associated with the use of a coolant with supercritical parameters. In this case, the heat transfer and the hydrodynamics of the coolant flow are significantly improved which, combined with a straight-through technologi- cal system, will enable the net output of the reactor to be increased to 800-1000 MW with almost the same active zone dimensions as for the Beloyarsk NPS reactors. By using a straight-through technological system, the signifi- cantly greater advantages of multipass operating channels are revealed most completely from the point of view of the operating temperature conditions of the graphite masonry and the fuel elements as well as from the point of view of introducing a relatively small quantity of structural materials into the core. Calculations indicate that by using uranium of 5% enrichment, a fuel burnup can be guaranteed which is equivalent to an output of 40,000 to 45,000 MWD from a ton of uranium. The increase in the unit power of the reactor, in the efficiency and uranium burnup will sharply improve the efficiency index of the nuclear power station with respect to cost per kW installed and also with respect to the cost of nuclear power sent out. In uranium-graphite reactors of the channel type, a change in constitution of the active zone is easily ac- complished, which is due to the interchangeability of the operating channels. This permits the use in subsequent reactor charges of more advanced operating channels and fuel elements, in particular it permits the use of low- absorption structural materials and fuel compositions with low nonproductive neutron absorption. Since the active zone of the reactor is located in an airtight housing, it may affect the fragmentproof clad- ding of the fuel elements which, on the one hand makes it possible to remove gaseous fission products from the ac- tive zone and on the other hand leads to a significant reduction in the quantity of steel in the active zone (by 30%). 1086 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 It is obvious that this improves considerably the neutron balance and the other reactor characteristics which are associated with this. Uranium-graphite reactors of the channel type are free from limitations associated with increase of power. Thus, as a result of increasing the diameter of the active zone by 35-40/o, the output of a single reactor can be increased to 1500 MW. LITERATURE CITED 1. N. A. Dollezhal' et al., Proceedings of the Second International Conference on the Peaceful Uses of Atomic Energy; Report of the Soviet Scientists, 2, [in Russian], Moscow, Atomizdat (1959), p. 15. 2. N. A. Dollezhal' et al., ditto, p. 36. 3. A. N. Grigor'yants et al., Report No. 308, presented by the USSR at the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva (1964). 4. G. N. Ushakov et al., Report No. 314, ditto. 5. N. A. Dollezhal' et al., Power Reactor Experiment, Vienna (1962), p. 41. 1087 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 NOVO-VORONEZH NUCLEAR POWER STATION ? IN OPERATION Dr. Mech. Sci. N. M. Sinev Deputy Chairman State Committee for the Utilization of Atomic Energy in the USSR Translated from Atomnaya Energiya, Vol. 17, No. 5, November, 1964 2 page insert following page 336 By the efforts of highly qualified teams of scientific, design and constructional organizations of many leading mechanical engineering concerns in the country, skilled builders and specialists in the construction of power gener- ating complexes, the complicated many-year cycle of planning, acquisition, construction, and installation of the Novo-Noronezh Nuclear Power Station has been completed. Difficulties of construction and startup of the single- channel equipment, of the laborious scientific and engineering calculations, investigations and experiments have 1088 - Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 been overcome. Many days, weeks, and months of worry and anxiety of the startup and commissioning period, to- gether with the intensive assembly and structural-finishing tasks, have been left behind. At 1545 hours on September 30, 1964, the first turbogenerator of the Novo-Voronezh Nuclear Power Station took the load and fed electric current into the Voronezh Energy System and, by the morning of October 2, the first million kilowatt-hours were produced. Only about 200 g of U235 were required for this instead of the 500-600 tons of coal for a conventional thermal power station. On the sixth day after startup, the turbogenerator load was raised to 65 MW; the thermal power of the reactor was 35% of the calculated power. During this period, two of the three in- stalled turbines were tested in operation (alternately), various effects were checked which are important for evalu- ating the system that ensures the safety of operation of the reactor (for example, as a result of the sudden de-ener- gizing of the station, actuation of the scram system, de-energizing of the load-breaker at the turbogenerator in the event of closure of the shutdown steam valve prior to the turbine), cut-out and switching-in of the steam generator loops was carried out, the reaction of the control mechanism was finalized, etc. Investigations of the physical parameters of the active zone were undertaken in various ways and answers were provided to questions of interest, having verified the good agreement with the calculated data in the magnitude of the reactivity and distribution of the neutron fields as well as in the efficiency estimate and the action of the reac- tivity compensation controls, taking into account the effect of the temperature coefficient. Construction and assembly of the first unit was achieved by organizations of the State Industrial Committee for Energy Generation and Electrification of the USSR, the ideological and scientific supervision in planning was carried out by the I. V. Kurchatov Institute of Atomic Energy, the general designer was the Moscow Division of the All-Union Design Institute "Teploenergoproekt" (MOTgP). Development and supply of plant was undertaken by many well-known structural engineers, scientific research and machine construction enterprises. The first unit of the Novo-Voronezh Nuclear Power Station, with an output of 210 MW is only the first "salvo" of this large-scale nuclear power station. At present, on the site of the Novo-Voronezh NPS and "flush" with the first unit, construction is taking place of a second unit with an electrical output of 365 MW. In the installation, structural and engineering decisions of the reactor portion of the second unit, accumulated experience as well as new advances in atomic science and technology have been taken into consideration. The cost per kilowatt of installed power in the second unit will be only 60% of the cost in the first unit, i.e., 192 rubles/kW compared with 327 rubles/kW. We note that the estimated cost of a coal-fired station operating on Donets coal with a unit of 300 MW would be about 100 rubles/kW here without capital investments in the extraction and transportation of the coal, which for this locality ? situated in the vicinity of Donbassa ? would amount to about 70 rubles/kW. Thus, taking into account future development, large-scale units for nuclear power stations with water- cooled?water-moderated reactors enable the feasibility of earning capacity to be assessed, as a result of installing them in the European part of the country. With the introduction into the system of the first units of the Beloyarsk and Novo-Voronezh Nuclear Power Stations, a new page in our country's energy generation has been opened. The time has come when the role and share of nuclear power stations in the fuel-energy generation balance of the European part of the USSR will steadily increase according to the extent of further accumulation of experience in the construction and operation of large- scale nuclear power stations, organization of mass production of special reactor plant and the development of pilot units of large power which are planned for construction in the immediate future. In this process of development, a specially important role is played by the Beloyarsk and Novo-Voronezh Nu- clear Power Stations, which have been transferred to the State Committee for the Utilization of Nuclear Energy in the USSR as experimental stations and as outposts of large-scale power generation in the USSR. It is planned to install at these stations several different units with high power reactors, which will enable all-round development of basic prototypes of economical nuclear power stations, to concentrate engineering experience and to create the basis of practical training of operating personnel for the future extensive program of nuclear power station construction. "Let the atom be a worker and not a soldier" is inscribed on the pediment of the main building of the Novo- Voronezh Nuclear Power Station. And all involved in the construction of this station ?scientists, engineers, and workers ?have endowed this great goal with their labor, with their enormous many-year efforts, expended on the construction of the first-born of Soviet nuclear power generation, making its first spectacular contribution in the building of a material-technological basis for communism. 1089 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 SODIUM-COOLED FAST REACTORS* A. I. Leipunskii, 0. D. Kazachkovskii, I. I. Afrikantov, M. S. Pinkhasik, N. V. Krasnoyarov, and M. S. Poido Translated from Atomnaya gnergiya, Vol. 17, No. 5, pp. 345-348, November, 1964 The fuel breeding capabilities of fast reactors open up new potentialities to the development of large-scale nuclear power. Fast neutron chain reactions yield large numbers of secondary neutrons per single fission event, re- duce nonproductive neutron capture with no fission in the &lel, or nonproductive capture in structural materials, and make it possible to increase the contribution of fissions in the raw material. As a result, the fuel breeding ratio ex- ceeds unity and fast reactors may be developed which use all the reserves of uranium and thorium, rather than just U235, as the raw material. Attainment of high parameters in the steam turbine cycle and high efficiency is important in the case of power facil- ities, The sodium used as coolant in fast reactors makes it possible to achieve such parameters in a radioactive reactor loop at relatively low pressures. Sodium retains the basic radioactive fission products in the event of rupture of fuel elements. Fast reactors using liquid metal coolants are currently in operation in the USSR, in the USA, and in Britain, and experience is being accumulated in the operation of these facilities. The experimental sodium-cooled fast re- actor BR-5 [1] has been operating successfully for the past half decade in the Soviet Union. Parameters such as up to 500 kW/liter power density, up to 6% fuel burnup, sodium temperatures to 500?C at the reactor exit, come close to the characteristics of large power reactors being developed which will provide appreciable savings. The first stage in the development was reached in the designing of a full-scale industrial power facility, the BN-50, with 200 MW(th), 50 MW(e) output rating, sodium temperature of 315?C at the pile entrance and 450?C at the pile exit. The core was a cylinder 67 cm in diameter and standing 67 cm high. Suggested fuel materials were uranium-molybdenum alloy, uranium oxide in a nickel matrix and nickel cladding, or sintered uranium oxide in a stainless steel jacket. The breeding ratio of the reactor using uranium-molybdenum alloy fuel was 1.35. Engineering designs for the BN-50 facility were frozen in 1960. The work on this project revealed not only the feasi- bility of building a reactor of this type, but also the possible design of power reactors of still greater output and improved performance. A decision was made in this connection to shelve the plans for the BN -50 project as an intermediary stage. Work on the design of a new reactor project is currently entering the finishing stages; this is to be a prototype of high- output nuclear fueled electric power stations of 1000 MW thermal rating and 350 MW electrical rating (BN-350). The basic reactor components are housed in a variable-diameter sodium-filled tank (sodium volume of 165 m3). The sodium coolant runs down six tubes at 300?C through the tank bottom to collect in the discharge header. The sodium becomes heated to 500?C, on the average, as it courses through the reactor, and it is then pumped through the heat exchangers from the reactor tank. The discharge header plenum has grids for mounting the fuel element assemblies. The central 211 assemblies contain the core fuel elements in the middle and the fuel elements belonging to the upper and lower blankets on the face of the reactor in their top and bottom, respectively. The peripheral 500 assemblies form the lateral blanket. Cells beyond the side blanket are used as storage area for the cool-down of fuel element assemblies before they are removed entirely from the pile. Steel slugs positioned be- yond the storage area constitute neutron shielding for the reactor tank. A general view of this reactor appears in an accompanying diagram. The core volume (about 2000 liters) and the irradiation level (500 kW/liter) minimize the fuel quantity in a Cycle. As the irradiation level is stepped up, the amount of fuel used increases because of the increased fuel used in chem- *Report No. 311 presented by the USSR delegation to the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1964. 1090 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Fig. 1. General view of reactor. 1) Core; 2) blan- ket; 3) vessel; 4) central column with control rod drives; 5) rotating plugs; 6) discharge elevator; 7) discharge control box. ical reprocessing or fabrication of fuel elements, and as the irradiation level is lowered the amount of fuel increases because of the increased critical load. The core shape determines the fraction of coolant present in the breeding zone, the flowspeed of the sodium and the hydraulic resistance to the pumped flow, as well as the extent to which the sodium becomes heated. These factors improve measurably as the core diameter/height ratio (D/H) is increased. However, the ratio of the vol- umes of the face shield and side shield increases con- comitantly, and at D/H > 1 the critical mass increases. A D/H ratio of 1.4 was decided upon in view of the heat transfer, hydraulic, and nuclear physics characteristics of the reactor. The core diameter was 1.5 m, core height 1.06 m, maximum sodium flowspeed 10 m/sec, average heating level of sodium in the core 230?C, volume frac- tion of sodium 39%. In order to effectively minimize neutron leakage from the reactor, the thickness of the side and face blankets was set at 60 cm. The reactor is so de- signed that the core dimensions can be either expanded or contracted by varying the power density, and to facilitate conversion to fuel element assemblies incorporating dif- ferent fuel materials. Ceramic fuel: a mixture of plutonium dioxide (19% plutonium) and U238, is used as fuel material for the BN- 350 reactor. A loading of enriched (23%) uranium di- oxide is possible in the first stage. Criticality is attained at 780 kg plutonium (950 kg 11235. As in the case of the BR-5 fast reactor, BN-350 fuel elements constitute a stainless steel tube 5 mm in diameter and 0.4 mm thick filled with sintered dioxide pellets. The pressure exerted by the gaseous fission prod- ucts sets up stresses in the cladding even when they have been completely released from the oxide, but these stresses combined with thermal stresses fail to exceed tolerances. Higher burnup levels can be attained when the average dioxide density is brought below that in BR-5 fuel elements. The operating life of a fuel element must be 300 days to achieve 5% burnup. Fuel elements are assembled in hexa- hedral assemblies ( of 217 fuel elements each) measuring 96 mm "across." Fins on the cladding aid in achieving an even distribution of the fuel elements. The fuel elements are spaced above and below with the aid of a grid into which the tailpieces of the fuel elements are fitted, The face reflectors are made up of 74 elements in the same assembly (37 above and 37 below, tube diameter 12 mm, tube thickness 0.4 mm) filled with depleted uranium dioxide pellets. The assemblies in the side reflector contain 37 elements each (14.2 mm in diameter, 0.5 mm thick), with depleted uranium dioxide. The fuel elements in the side and face shields are spaced to a density of 9.5 g/cm3on the average. The described composition of the core and blankets provides for a fuel breeding ratio of ?1.5; the internal breeding ratio is 0.62. The reactivity change with reactor on-power was 0.6% over a month's time and was chiefly due to the lowered concentration of plutonium in the core. Reactivity losses were compensated by displacing six fuel element assemblies with elements positioned in the central part of the core and featuring 1.4% reactivity, there- by assuring operation free of shutdown for two months. The reactivity change while the reactor is heating up and being brought up to full power is 2.4. 10-5/?C and (0.5-0.7). 105/MW (a larger value is attained when burning fresh fuel, in which case the oxide pellets are not bonded to the cladding and the elongation of the oxide column takes place independently of the elongation of the cladding). The principal contribution to the power effect is that of the U238 Doppler effect. The sodium ratio integrated over the core comes out negative. 1091 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 The temperature and power effects are compensated by the withdrawal of fuel assemblies incorporating NC- loaded fuel elements from the core and shifting these to the top face reflector; B10 enrichment here is as high as 60%. The total compensator efficiency is luio when 0.7 kg B1? is present. During transients, one of the two automatic control rods will be operating in unisom with the temperature com- pensator. The control rods are made of enriched B4C compacted slugs enclosed in stainless steel tubes. The efficiency of each control rod is 0.2%. Since the control rods are constantly present in the core while the power level is being maintained, a special volume is provided in the rod design to collect the helium released in the rods. Three assemblies of enriched B4C rods are employed for scramming the reactor. The total excess reactivity in these rods is 4.5%. Depending on the signals operating on the scramming system, boron rods may be dropped at vari- ous speeds. The shielding of fast reactors incorporates some special features due to the high fast-flux intensity com- ing out of the reactor. The neutron flux at the edge of the breeding blanket attains the quite considerable figure of 5.1013 neutrons/cm2. sec. The side shield of the reactor is therefore designed as a primary shield to lower the flux impinging on the wall of the reactor tank, backed up by secondary shielding outside the tank to lower the radiation flux impinging on the concrete, and then the concrete shielding. Lowering the neutron flux striking the tank and a reduction by a factor of ten in the amount of heat released in the tank walls is achieved by means of steel slugs (120 mm thick) placed on the far side of the storage volume, by a layer of sodium (500 mm) and by the steel ther- mal shield of the tank (60 mm). The heat released in the tank wall is 0.1 W/cms. The secondary shield is made of a steel layer (150 mm) and an iron oxide layer (1000 mm) designed to lower the radiation flux to 5.109 MeV/cm2. sec. The ordinary concrete shielding is 2000 mm thick. The top shield incorporates a layer of sodium, a steel plate sunk in the sodium, and sandwiched iron and graphite layers. A system of rotating shielded plugs carrying a reloading mechanism, and the refueling control box with trans- mission gearing, is provided for refueling the reactor with fresh fuel assemblies and for unloading spent fuel assemblies. The first loop includes the reactor, a sodium-sodium heat exchanger, the pump, cleanup system, and a system for monitoring oxides in the coolant. Stop valves are installed on the exit and entrance piping (630 and 529 mm in diameter, 15 and 13 mm thick, respectively). Ahead of these stop valves, the piping is enclosed in a protective car- bon steel jacketing 5 mm thick. Immersion-type vertical-shaft centrifugal pumps working in parallel build up a sodium flowrate of 14,100 tons/h through the reactor at a head of 12 kg/cm2. The U-leg heat exchanger presents a heat transfer surface area of 850 m2. The secondary loop includes a sodium-sodium heat exchanger, a sodium- water-steam steam generator, a pump, a cleanup system, and a system for monitoring sodium oxide in the flow. The steam generator was designed for natural circulation with a single baffle. Wall rupture experiments and experiments involving interaction between sodium and water or steam were set up in steam generator models to test out the single-baffle design of the steam generator unit. The experiments demonstrated the feasibility of this de- sign. Steam generated at 430?C temperature and 50 atm pressure is supplied to the turbogenerators. The primary loop equipment is so fabricated and so arranged as to permit operation and replacement of the equipment under high-level radioactive contamination conditions. Acylindrical shielding cover of stainless steel, 9.5 m in diameter and 10.4 m high, is placed above the reactor to localize possible leaks of radioactive products. A removable access hole 2.5 m in diameter is situated above the cylindrical cover. Emergency cool-down of the reactor is achieved by natural circulation and by utilizing the energy of inertia of the turbogenerators. For example, in the event of an outage of the electric power supplies, the greatest hazard to be faced of likely emergency situations, flow of sodium through the reactor during the first 60 to 100 sec will be brought about by circulation pumps powered directly by the inertia of the turbogenerators. The heat will be re- moved subsequently by natural circulation of the coolant through the primary and secondary loops. The BN-350 is the USSR's first high-flux fast power reactor. This accounts for the decision to rely on already achieved and experimentally amply justified system parameters ensuring adequate reactor reliability in performance. The installation and operation of the BN-350 reactor, in addition to the development of research and experimental work with the reactor, provides sure grounds for assessing the possibility and feasibility of various new ideas holding promise of further progress in the field of fast power reactors. Some new obvious pathways for improving this type of reactor are: 1. Improving fuel burnup, which entails reduction of the amount of fuel per cycle, reducing the number of fuel elements spent and reprocessed. 1092 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 2. Reduction in the amount of steel in the core, and increasing the content of the fuel. At the present time, the choice of 0.4 mm thickness for cladding in BN-350 fuel elements rests primarily on the fabrication technology conditions and on the need to withstand stresses imposed by gaseous fission products. The use of fuel elements de- signed to release fission products in gaseous form into the coolant will make it possible to reduce the cladding thick- ness, and this will mean a substantial improvement in the breeding ratio. 3. The use of other fuel materials. Carbide and metal fuel makes it possible to achieve a breeding ratio higher than that associated with oxide fuel. To date, however, there is no record of any experiment on high burnup of compositions of that type. Uranium monocarbide fuel elements are now being tried out in the BR-5 reactor, and almost the entire core of BR-5 is scheduled to be converted to monocarbide fuel by the end of 1964. 4. The use of promising short-cut techniques in fuel reprocessing, combined with enhanced power density in the core to bring about a substantial reduction in the amount of fuel per cycle. 5. Increasing plant thermal efficiency. An increase in coolant temperature at the reactor exit makes it possi- ble to raise the steam parameters. Projects are now underway to develop a power facility with a rating of one million kW(e) (the BN-1000 proj- ect). The use of a mixture of uranium monocarbide and plutonium monocarbide will produce a breeding ratio of 1.75 in the BN-1000. In-core breeding will completely compensate for the fuel burnt up, and this will greatly sim- plify requirements for reactivity compensation, while facilitating long-term continuous reactor service life. The reactor core will be shaped flatter than the BN-350 core [2]. Steam parameters of 580?C and 240 atm are the goals of this project. A decision has been made to proceed ahead on the development of the fast experimental reactor BOR, with an eye to providing the groundwork for the basic technological and design decisions relating to this type of high- temperature high-stress reactors, as well as other promising fast reactor types. The BOR sodium-cooled reactor is designed to achieve 40-60 MW(th) with up to 1500 kW/liter power density per unit core volume; the temperature of the sodium coolant leaving the facility will be 630-650?C. Plans call for studying the possibility of achieving higher than 10% bumup and employing thin-clad fuel elements capable of releasing gaseous fission products. The operation of the BOR reactor will provide extensive experience on the stability of fuel elements of different design and of dif- ferent compositions of fissile materials at high-bumup levels, as well as experience in operation of the reactor, an- cillary equipment and process instrumentation, experience in the technology of radioactive sodium at elevated oper- ating temperatures. The experience accumulated will pave the way for the building of low-cost fast reactors pro- viding cheaper electric power than that available from fossil-fuel power stations. LITERATURE CITED 1. A. I. Leipunskii et al., Report No. 312 presented by the USSR delegation to the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva (1964). 2. A. I. Leipunskii et al., Report No. 369, ibid. 1093 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 OPERATING EXPERIENCE WITH THE NUCLEAR PROPULSION PLANT ON THE ICEBREAKER "LENIN"* I. I. Afrikantov, N. M. Mordvinov, P. D. Novikov, B. G. Pologikh, A. K. Sledzyuk, N. S. Khlopkin, and N. M. Tsarev Translated from Atomnaya gnergiya, Vol. 17, No. 5, Pp. 349-359, November, 1964 The nuclear-powered icebreaker *LENIN" undertook its shakedown cruise as part of the Arctic Sea Fleet of the USSR on December 3, 1959. Since that time, it has participated every year in Arctic navigational practice. By the end of 1963, the ship had traversed about 60,000 miles, including about 40,000 miles through the ice. Together with its sister icebreakers, the nuclear-powered icebreaker opened a path for over 300 vessels on the Northern Sea Route. In October, 1961, the ship made a delivery of equipment and servicing personnel to the drifting patrol station "North Pole-10." The use of a nuclear power plant made it possible to design an icebreaker capable of enormous power output and navigational independence, featuring excellent maneuverability and iceworthiness. The combination of these positive qualities on an icebreaker powered by a propulsion plant burning chemical fuels is unattainable in practice. The, by no means negligible, operational advantages to be had in using a nuclear propulsion plant are the possibility of keeping the water displacement of the icebreaker practically constant, and the possibility of operating at full power over a protracted time span combined with more reliable use of the icebreaker under severe ice conditions. The use of the powerful nuclear icebreaker on the Northern Sea Route contributes to a greatly increased speed through the ice of vessels negotiating icy and icebound seas, and to app. eciable lengthening of cruise times. Cases of forced wintering of icebound ships are virtually eliminated, the number of accidents is reduced, and the possibility of loss of ships nipped by ice packs is minimized. This is exemplified by the successful trip by vessels on the Arctic run during 1960, when the icebreaker "LENIN" forestalled a threat of the loss of several ships in the ViPkitskii Strait. The long-term operation of the "LENIN" on the open seas under the most hazardous conditions in navigation showed that the nuclear propulsion plant operates in a stable manner, is easily controlled under any and all sudden changes in load, and fulfills all the power requirements of the vessel. The vibration-and shockproof equipment of the nuclear propulsion plant operates reliably in the face of collisions with ice formations and heaving and rolling of the ship. The reactors on board the "LENIN" are now operating on their second fuel loading. The reactors were refueled in the spring of 1963. Each of the reactors operated for over 11,000 h on the first fuel loading, developing from 430,000 to 490,000 MWh of thermal power. The average burnup over the reactor cores was 11,000-13,000 MW ? days/ton U, and the maximum burnup was about 30,000 MW days/ton U. The fuel elements were immersed forabout 30,000 h in the water of the primary loop. The reactors operated stably at all power levels, including the maximum level of 90 MW. The total power developed by the icebreaker was 44,000 hp, when all three reactors were developing 65 MW each. Performance Characteristics of the Nuclear Power Plant The total weight of the nuclear propulsion plant, shielding included, is about 3100 tons. The plant is designed to handle 360 tons of steam per hour at a pressure of 28 kg/cm2 and a temperature of 300-310?C. Figure 1 shows a general view of the nuclear power plant, and Fig. 2 shows the basic layout of the plant. The primary loop is made' upof three autonomous sections, each of which includes the following equipment: a reactor, *Paper No. 313 presented by the USSR delegation at the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1964. 1094 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Fig. 1. General view of steam power plant. 1) Heat exchanger for third and fourth loops; 2) pump Mr internal cooling loop; 3) primary-loop principal circulation pump; 4) primary-loop emergency circulating pump; 5) tem- perature monitors display room ; 6) cooler for primary-loop filter; 7) steam generator ; 8) steam lines; 9) control and protection system actuators; 10) ionization chamber; 11) carborite; 12) reactor;, 13) core; 14) primary.-loop piping; 15) iron-water shielding block; 16) primary-loop filter; 17) concrete; 18) gate valve for primary-loop main line; 19) heat exchanger serving internal cooling loop; 20) pressurizer. two steam generator units, four Orincipal.purnps,and two emergency circulation pumps, four pressurizers, and 'two. ion exchange filters. Each section has two loops: a foie loop and an aft loop; this is convenient both for regular operation and for maintenance of the fittings and equipment:. .The principal operating mode is operation of both loops. One principal circulation pump is in operation in each loop in this mode,the other pump' being on hot standby. The reactor is capable of operating at up to 50 MW power output in, the single-loop Mode, with the principal - and emergency circulation pumps working. ? - . - The throughput of the pumps on the primary loop turned out to be slightly greater than the rating, so that the water was not heated as high is, originally planned in the reactor in compensation. This becomes particularly clear in Fig. 3, where we have the theoretically predicted (curve 1) and experimentally derived (curve 2) dependences of the temperatures at the reactor exit and entrance on the reactor power. 1095 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 330 320 310 p 3oP - 290 a czt k 280 -270 ? H ? 290 250 1 240 2300 , 10' Fig. 2. Diagrattatic flowsheet of power sheet. ? cool- ant lob/3; SG'"steam-generator; K07-pressnrizet; MCP?. main circulating pump; ECP?emergency cireulating pump; IF?ion, exchange filter; FC?filter cooler unit; X?cooler for internal coolant loop ; FP?coolant Make- _ up pump; 1.) reserve coolant tanks; 2) surge tank; 3, 5, 10)to drainage system; 4) feedWater line; 6) frcim drainage tank; 7) to mixing tank; 8) feedwater line; 9) steam bleed line; 11) discard line in case of embrittlernent; 12) filter. 1096 ? 2 '4 0 ? 0 20 30 40 50 00 70 80 90 100 110 Power output, % ?Fig. 3. POWer dependence of reaCtOr entrance and exit ? terriperatares.... In 1961,411:alteration was made in the temper- ature'conditiOnt of the leadtors in order to bring them .closer to 'Self-regulating conditions where the tem= peranireeffect is?practiOally,eritirely Compensated by the boppler'effect. As we see in Fig.3, the water ternperature.4t the reactor it rernains..prdctically Constant (curve 3) at all_pOwer levels in the self- regulating mode. , The Characteristics corresponding to the self- regulating mode changed slightly during-the react& operating peridd because of Changes in the tempera- ture effect and in the Doppler effect, but-this did not occasion any peed to readjust the control system. The design of the mechanical side Of the 'Plant proVides for the.Possible supOlying-Of all on-board 'itern consuming units from an overall main line serving, all parts of the Ship: the main turbines, the electric power plants, evaporators, and so forth. -Sections including groups of steam generator units from one or two- reactors furnishing steam to the fore or aft echelon's of on-board equipment May be 'delineated in the common Main steam line by means of valves and fittings. Experience has shown that -operating all the steam-generator units into a-single common main steam line offers Significant advantages - not found in the echelon setup , When one of the reactors is scrammed with :steam generators being operated on the echelon pat- tern, the supply of steam to consumer .units from the- stearn'generators of that reactor-may be cut off. . ? ' When the stearin generators are working into a single'common main steam line, it seems possible to Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 TABLE 1. Design and Operational Performance Characteristics of the Icebreaker "LENIN" at Full Power Performance parameters Design values for 65 MW Performance data for reactors 1 2 3 Water flowrate in primary-loop lines, tons/h 415 435/430 458/467 435/453 Temperature at reactor exit (flowing through loops), ?C 317 311/311 312/313 311/313 Temperature at reactor entrance, ?C 261 260 261 260 Steam throughput on loops, tons/h 43.3/43.3 42/42 47/42 43/46 Steam pressure, kg/cm2 29 32/31 31.5/30.5 31/31 Steam temperature, ?C 307 310/309 308/308 308/308 Power output computed for primary-loop parameters, % 72.3 69.6 75.4 73.1 Power output computed for secondary-loop parameters, % 68 67.3 71.1 71.3 Remarks. Figures in numerators of fractions refer to fore loop, those in denominators to aft loop. maintain the steam pressure in that main line by increasing the power output of the other reactors, so that there will be no need to completely shut off the steam consuming units on board. As a result, the number of valve operations in the transient and emergency modes of operation is reduced, and this diminishes the probability of erroneous judg- ments on the part of operators at the most crucial instants in the operation of the plant. The operating experience showed that the measures provided for on the icebreaker are adequate to ensure outage-free electric power supplies for the nuclear propulsion plants, by relying on two electric power plants each generating an output of 3000 kW. In the event that the voltage should disappear at one of these electric power plants, there are two emergency diesel generators of 2100 kW total rating which would be switched on automatically, and a standby diesel generator unit of 1000 kW rating can be switched on manually if required. On the suggestion of the servicing personnel, the originally planned two-way relay contactor circuitry for elec- tric powering of the control and protection system for the reactors was replaced by a semiconductor switching system for enhanced reliability in operating the plant. It is important to emphasize the fact that not a single instance of malfunction in the emergency protection equipment was encountered during the entire period the reactors were on-power. Studies showed that in some cases, when one of the circulation pumps in the primary loop was shut down for instance, there was no need to cut the power to zero; on the contrary it was quite sufficient to rely on an automatic rapid power drop to the 30% level, a proce- dure known as second-order emergency operation. Some of the signals previously used to dump the power to zero level were converted to trigger second-order scramming signals. This meant an increase in the ship's service life and diminished the number of abrupt thermal surges experienced by the equipment. Automatic dumping of the power to zero level is known as first-order scram- ming in this context. The number of scrams during the sailing of the icebreaker over the Northern Sea Route was limited: six to eight such events for each of the reactors. The primary-loop equipment of the power plant on board the icebreaker underwent about 15,000 h of service under regular operating conditions (180 kg/cm2 pressure and tem- peratures of 250-310?C), including service in 1963. The main circulation pumps operated over 8000-9000 h without overhaul. Some of them were withdrawn from service because of a loss of resistance on the part of the insulation on the stator windings. The emergency circula- tion pumps operated reliably. The steam compensation system maintained the high stability of the pressure in the primary loops at constant levels and permitted pressure deviations no greater than ?5 atm during transients. Radio- active corrosion products were precipitated in components of the bottom parts of the pressurizers, and this added dif- ficulties to the dismantling and replacement of the electric heater units. At this writing, that phase of the work had not been completed. 109'7 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 Declassified and Approved For Release 2013/03/14: CIA-RDP10-02196R000600120001-5 The ion exchange filters made it possible to maintain the required high quality of the water used in the pri- mary and secondary loops of the propulsion plant: a resistivity of 1-2 MO. cm; chlorine ion content no greater than 0.02 mg/liter, and pH = 6-8. The ion exchange resins KU-2 and AV-17 have been used successfully in these filters In the recent period. Hydrazine was tried out in the water of the secondary loop. A short-term increase in the con- tent of salts in the feedwater of the secondary loop was observed when the mechanical power of the icebreaker was first set into operation. As a result of the heightened quality of individual units of equipment, and improvements in the inspection system, these phenomena occurred only very rarely subsequently, and were rapidly localized in their effects. The reader should take note of the effective use of double tubesheets for the condensers and coolers treating the on-board water, thereby practically eliminating any taking in of sea water. The steam generators of the nuclear power plants operated stably under both steady-state and transient condi- tions. Cases of leaks in their piping systems did occur during operation. The detection of leaks and corrective ac- tion in cutting out the steam generator involved from the line were carried out with rapidity and in well organized steps. As a result, only a short-term increase in activity was observed in the secondary loop, with the background values exceeded only by several times. Cases of dropwise leaks through the stuffing boxes of the principal slide valves due to the gland packing drying out were observed in the primary-loop lines. The packing was replaced by higher-quality material, but the stuffing boxes still had to be tightened up once every 1500 or 2000 h. Bellows valves in the drainage system of the primary loops proved to be insufficiently reliable components. These valves had to be checked out once a year. There are no outstanding remarks to be made on the control and protection system actuators, all of which oper- ated reliably. The biological shielding around the reactors, around the equipment and the main steam lines of the primary loops, proved quite sufficient, and no damage of any kind was observed in this shielding as a result of impact load- ings produced by the motion of the icebreaker ramming ice formations, or by heaving and rolling of the ship when sailing on stormy weather. Some highly active slurries did get into a few of the pulse tubes leading from the bottom of the pipes. The y -emission levels were higher at the points where these tubes protruded beyond the biological shielding to the monitoring and measuring instruments, so that additional local shielding was provided for at those points. Nuclear power plant overhaul needs dictated an expansion of the health physics room at the entrance to the rigorous exposure control zone and an increase in the volumes occupied by the storage areas for liquid and solid wastes. The health physics rooms and the storage racks were reequipped to handle the situation. Health physics instrumentation was adequate to monitor the radiation environment on the icebreaker and to keep track of activities in the process loops of the nuclear plant. During operation some of the