Soviet Atomic Energy Vol. 47, No. 3

Body:  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 ISSN 0038-531X Russian Original Vol.-47, No. 3, September, 1979 March, 1980 SATEAZ 47(3) 691-790 (1979)' SOVIET ATOMIC EN ERGY ATOMH'AH 3HEPfNH (ATOMNAYA ENERGIYA) TRANSLATED FROM RUSSIAN C CONSULTANTS BUREAU, NEW YORK Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 SOVIET Soviet Atomic Energy is a Cover-to-cover translation-of Atomnaya fnergjya,`a-publication of the Academy of Sciences of the USSR. An agreement with the Copyright Agency of the" USSR (VAAP) ATOIVIIC makes available both advance copies of the Russian journal and original glossy photographs and artwork. This serves to decrease the necessary time lag between publication of the original and publication of the translation and helps to improve the quality ENE.RY Hof the latter. The translation began-with the first issue of the Russian' journal., Editorial Board of Atomnaya Energiya` Editor: 0. D. Kazaphkovskii Associate Editors:,. N. A. Vlasov and N_. N.' Ponornarev-Stepnoi, Secretary: A. I. Artemov I.. N. Golov.in V. I. l I_'ichev V. E. lv.anbv V. F. Kalinin P. L. Kirillov\ Yu. I. Koryakin A. K. Krasin' E. V. Kulov B. N.-Laskorin V. V. Matveev I. D. Morokhov A. A. Naumov A. S. Nikiforov A. S..' Shtan' B. A. Sidorenko M. F. Troyanov E. 1. Vorob'ev Copyright (D_ 1980, Plenum Publishing Corporation. Soviet Atomic Energy partici-; pates in the program of Copyright Clearance Center, Inc. The appearance of a code line at the bottom of the first page of an article in this journal indicates the copyright owner's consent that copies of the article' may be made for personal or internal us~. 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Second-class postage paid at?Jamaica, New Yor'k 11431.  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya March, 1980 Volume 47, Number 3 September, 1979 CONTENTS THIRTIETH ANNIVERSARY OF THE GERMAN DEMOCRATIC REPUBLIC Development of the Nuclear Power Industry in the German Democratic Republic - W. Mitzinger .................. ............ Nuclear Research of the Academy of Sciences of the German Democratic Republic in the Light of the Decisions of the Ninth Congress of the German Socialist Unity Party -K. Fuks ............................. .................. . ARTICLES Effect of Nonuniformity of Fuel Depletion with Height on the Physical Characteristics of a Reactor - A. M. Afanas'ev and B. Z. Torlin ............ Power Effect of Reactivity in Fast Power Reactor with Allowance for Behavior.of Fuel under Irradiation - G. M. Pshakin and A. A. Proshkin ..................................... . Theoretical and Experimental Investigation of Sodium Void Effect of Reactivity - S. P. Belov, P. V. Gerasimov, Yu. A. Kazanskii, V. I. Matveev, G. M. Pshakin, and P. L. Tyutyunonikov ................... Minimization of Loss of Energy Output by System of Reactors Operating with a Variable Load Schedule - V. I. Naumov and A. M. Zagrebaev ................... ..................... . Effect of Entrance Conditions on the Development of Turbulent Flow in Circular Pipes - B. N. Gabrianovich, Yu. D. Levchenko, Yu. P. Trubakov, and P. A. Ushakov................................. A Graphicoanalytical Method for Determining the Length of Elements along the Height of a Multielement Thermoemissive Assembly - V. V. Sinyavskii ....................... ................. . Fission Neutron Detectors -Z. A. Aleksandrova, V. I. Boll shov, I. E. Bocharova, K. E. Volodin, V. G. Nesterov, L. I. Prokhorova, G. N. Smirenkin, and Yu. M. Turchin .................. .... ...... . Analysis of the Reliability of Radiochemical Plants. with Electron Accelerators - V. M. Kshnyaskin and Yu. D. Kozlov ..................... NEW BOOKS E. P. Anan'ev. Atomic Plants in Power Engineering - Reviewed by Yu. I. Koryakin ................ ................ . Engl./Russ. 691 147 693 149 697 152 703 157 708 161 713 165 715 167 718 169 721 172 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 CONTENTS (continued) Engl./Russ. LETTERS TO THE EDITOR Evaluation of the Selectivity of Electrochemical Reactor-Fuel Recovery on the Basis of Thermodynamic Data - V. A. Lebedev...... .......:.... 731 180 Determination of Neutron and Radiation. Components of Energy Release in Boron-Containing Rods,U.sing Gray Chambers - V.I P. Polionov, Yu.G.Pashkin,andYu.A.Prokhorov.................................... 733 182 Photoproduction of Neutrons in a Thick -Lead Target - V. I. Noga, Yu. N. Ranyuk, and Yu. N. Telegin ..................................... .735 183 Mathematical Model for Calculating Fission Products Concentration and Energy Release in Circulating Nuclear Fuel - L. I. Medvedovskii, E. S. Star-iznyi, VV. A.'Cherkashin, V. A. Rudoi, and K. I. Stepanova ....................................... 737 ?184 LETTERS TO THE EDITOR A Possibility of Reducing the Doubling Time. for Thermal Liquid=Salt Breeder Reactors V. L. Blinkin ...... 740 186 A Possibility for the Use of Highly Active Fuel Regeneration Wastes of Fast Power Reactors -.E. M. Vetrov and E. M. Ikhlov ..................... 742 187 Use of a Crystal Synchrotron Target to Obtain a Positron Beam - V. G. Potemkin and S. A. Vorob'ev ... ' ................................ 744 188 Gas-Chromatographic Apparatus for,, Determining Carbon in Uranium and Uranium Dioxide Pellets with Serial Loading of Specimens - V. A. Nikol'skii, V. K. Markov, A. S. Panov, and B. S. Valyunin ................. 746 190 a Particle Recording with RF-3 Film Track Detectors -I. V. Zhuk, A. P.Malykhin, L. P. Roginets, and'O. I. Yaroshevich .................... 748 191 Identification and Estimate of Tritium Content in VVR-M Reactor Water - A. M. Drokin, V. K. Kapustin, V. -P. Korotkov, V. V. Leonov, V. K. Mironov, and Yu. P. Saikov ........................... `750 192 Calculation of X-Ray and v-Ray: Photoelectric Attenuation -Factors for Statistical ,Modeling of TransportProcesses =S. Marenkov and T. V. Singarieva. 752 194 Purification of .Iron from U and Ra Microimpurities by Zone Melting - I. R. Barabanov, L. P. Volkova, V. N. Gavrin, V. L Glotov, D. S. Kamenetskaya, L. L. Koshkarov, I. B. Piletskaya, and V. I..Shiryaev ..................... ........ 754 195 Reliability of Detection of Sodium Boiling by Correlation of Acoustic and Neutron Noise - B. V. Kebadze and K. A. Aleksandrov .................. 756 197 Calculation of Sanitary-Protective Zones around Accelerators - Yu. A. Volchek .................. ....... ............ ........ 759 1-98 Systematics of (n, p) and (n, a) Cross Sections - V. N. Levkovskii , .. 762 '200. Reactimeter with 'a Pulsed Measurement Channel - V. A. Lititskit, A. G. Kostromin, V. V. Bondarenko, and F. B. Bryndin........................ 764 202 Estimate of the Risk from the Combined Action of Radiation and Chemical Agents - V. N. Lystsov and V. A. Kinzhinkov ..................... 7167 203 Estimate of Doppler Broadening of Resonances - V. V. Kolesov and A. A. Luk'yanov ................................... .......... 770 2,05. Neutron Resonances of 241Cm In the Energy Range 0.5-20 MeV - T. S. Belanova, A. G. Kolesov, A. V. Klimov, S. N. Nikol'skit, V. A. Poruchikov, V. N. Nefedov, V. S. Artamonov, R. N. Ivanov, and S. M. Kalebin ....................... .................. 772 -2.06 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 CONTENTS (continued) Engl./Russ. CONFERENCES, MEETINGS, SEMINARS Soviet-British Seminar on Fast Reactors - R. P. Baklushin ...................... 774 208 Conference on Hydrogen Power Generation - Yu. I. Koryakin ..................... 776 209 Second Conference of the Consultative Group on Nuclear Data for the Isotopes of the Actinide Elements - V. M. Kulakov ..................... 777 210 Soviet-Swedish Seminar on the Burial of Radioactive Waste - L. P. Zavyal'skii .............................................. 779 211 National Conference in the USA on Charged-P article Accelerators - Yu. M. Ado and I. N. Semenyushkin .................................. 780 212 BRIEF COMMUNICATIONS Tenth Spring Symposium on High Energy Physics - A. B. Kaidalov .................. 783 213 Fifth Meeting of the Combined Soviet-Canadian Working Group on Collaboration in the Field of Power Generation - M. B. Agranovich ........... 784 213 First Meeting of the Joint Soviet-French Working Group on Collaboration in the Field of Electric Power Generation - M. B. Agranovich ................... 784 214 First Moscow Kurchatov Lecture - I. A. Reformatskii .......................... 785 214 NEW BOOKS Kh. Wong. Basic Formulas and Data on Heat Exchange for Engineers - Reviewed by P. L; Kirillov ....................................... 786 215 I. I. Malashinina and I. I. Sidorova. Training Equipment for Nuclear Power Station Operators - Reviewed by S. G. Muradyan ..................... 787 215 G. M. Fradkin (Editor). Radioisotope Sources of Electric Power - Reviewed by A. A. Efremov .......... ......................... . 788 215 The Russian press date (podpisano k pechati) of this issue was 8/23/1979. Publication therefore did not occur prior to this date, but must be assumed to have taken place reasonably soon thereafter. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 THIRTIETH ANNIVERSARY OF THE GERMAN DEMOCRATIC REPUBLIC DEVELOPMENT OF THE NUCLEAR POWER INDUSTRY IN THE GERMAN DEMOCRATIC REPUBLIC* W. Mitzingert In the 30 years of the German Democratic Republic (GDR) the power industry of the country has to an every increasing extent become an important factor in the growth of the national economy. In accordance with the further development of the socialist society the energy policy of the country is aimed at ensuring the well- being of the nation, in the service of the working class and all working people. The General Secretary of the Central Committee of the Socialist Unity Party of Germany (CC SUPG) and Chairman of the State Council of the GDR, E. Honnecker, stated at the Ninth Congress of the Party that the creation of a powerful modern energy and raw materials base is a fundamental condition for the development of the productive forces for the gradual transition to communism and its material and technical foundations. The energy policy of the country, deter- mined at the Eighth Congress of the SUPG and reaffirmed at the Ninth Congress is based on three main prem- ises: security of fuel and energy supplies through maximum use of domestic energy and raw-material resour- ces; rationalization of the processes of conversion, transportation, and application of energy for a further decrease in the specific fuel and energy consumption; implementation of a comprehensive program of socialist economic integration for the extensive utili- zation of the scientific and technological advances, the creation of specialized production of highly effi- cient plant, particularly in electrical machine construction, and stable, long-term supplies of raw mater- ials and fuel for the country. In accordance with these premises, low-calorie brown coal will, as before, remain the main source of primary energy in the country, at least up to the year 2000. In the future as well a large proportion of the energy demand will be covered by domestic brown coal. The directives of the Ninth Congress of the SUPG set the goal of "ensuring the production of domestic solid fuel with the minimum possible costs by increasing the power and efficiency of existing strip mines and concentrating plants in combination with the discovery of new open pits." However, because of the limited possibilities of increasing extraction and the increasing deteriora- tion of the geological and hydrological conditions, further expansion of brown coal production, especially after 1990, will be restrained by natural causes. In the light of present concepts, atomic energy presents the only possible alternative for meeting the requirements of the GDR for energy, especially after 1990. Thus, the early development of an industrial base for nuclear power in the country is envisaged. With due regard for the sci- entific and technical potential and the structure of industry, this problem can be solved only in close cooper- ation with the Soviet Union and other COMECON member-nations. As long ago as 2 years after the start-up of the Obninsk Atomic Power Plant (APP) an intergovernmental agreement was signed on the joint construc- tion of the Rheinsburg APP with a VVER (water-moderated-water-cooled power reactor) with an electrical power of 70 MW. During the design, construction, andoperationof this APP scientists, designers, builders, erectors, and operators of our country had the opportunity to become acquainted with the new technology in close cooperation with Soviet specialists. While the technical design and the principal equipment, of the APP were supplied by the Soviet Union, the detail design and auxiliary equipment, including the stream generator, were elaborated and built in the GDR. The construction and the assembly of the equipment of the APP were carried out by GDR specialists. The successful start-up of the Rheinsburg APP on May 8, 1966, on the anniversary of liberation from fascism, demonstrated that in the GDR the prerequisites had been created for mastering atomic energy in fra- ternal cooperation with the Soviet Union. 'Thus, the road was laid for mastering a new source of energy, atomic energy. Already in 1965 a new agreement was concluded with the Soviet Union on joint work on the con- *Translated originally from German; ?1979 by Akademie-Verlag, Berlin. tMinister of Coal Mining and Power. Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 147-149, September, 1979. 0038-531X/79/4703- 0691$07.50 ? 1980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 struction of the next APP. The site chosen for it was the Lubmin wasteland on the shore of the Baltic Sea, roughly 20 km to the northeast of Greifswald. The choice of site was due mainly to the presence of cooling water as well as the inadequate supply of electrical energy in the northern regions of the country. The reactors chosen for this APP were of the pressurized water type of reliable Soviet design, similar to those installed at the Novovoronezh APP. The electrical power of each energy unit is 440 MW. Such reactors have come into most widespread use primarily because of their economy, high degree of safety, and use of pressurized water as the working medium, which has been well investigated in ordinary engineering. The first unit of the Greifs- wald APP, which has been named the Bruno Loischner APP, went into operation at the end of 1973. The com- mercial introduction of atomic energy thus began in the GDR. At present, the APP operates four energy units with VVER-440 reactors with a total power of 1760 MW. The APP accounts for more than 9% of the power of all the electricity generating plants of the GDR. The next energy units are to be constructed both on the site of the Bruno Loischner APP and on the Elba, about 25 km from Stendal. Over the next decade the contribution of the APP to the installed capacity of electric power stations in the GDR will rise to 15-20%. At the same time, plans call for a transition to pressurized-water reactors with a higher unit power. The APP built in the country have displayed high operating characteristics. In the first place this per- tains to operating readiness and safety. The Rheinsberg APP has been operating consistently for more than 13 years. In recent years it has been reequipped into a research center and a center for training personnel for the nuclear power industry. With a training unit coming into service here in 1975, conditions approximating those in practice as closely as possible were created for training and raising the qualifications of operating personnel for APP of the GDR and other COMECON member-nations. In the critical months of the 1978-1979 winter the operational readiness of the units of the Bruno Loischner APP was 95-100%. Operation in January and February, 1979, which were particularly severe in the northern regions of the country, demonstrated that the APP is relatively independent of the weather conditions. Operating at full capacity, it ensured electrical energy for the country even when, because of exceptionally heavy snowfalls, the plant was temporarily cut off from the outside world. Under the conditions existing in the GDR a capacity of 440 MW for a power unit is economically justified. As for the capital investment and energy costs, a unit with a VVER-440 reactor is com- parable with a modern 500-MW unit operating on brown coal. The construction and operation of APP in the GDR are under strict government control. Operating ex- perience has confirmed that the characteristics are favorable for the environment. The radiation in the direct vicinity of the APP is less than 1% of the natural radiation. The long-range energy policy of the GDR is aimed at the continuous development of nuclear power with the use of mastered, reliable, and tested APP and invidi- dual APP systems. In order to increase operating safety provision has been made for continuous and annual inspection of the state of the APP, especially the nuclear part, this being done in the form of comprehensive. examinations of the equipment on the basis of instructions strictly laid down by law. The high requirements concerning- the professional training of operating personnel, regular enhancement of the qualifications in com- bination with qualification examinations, and continuous and concrete monitoring of the state of the training of operating personnel by on-the-spot analysis of maladjustments and regular training sessions for handlingemer- gencies are also important measures for ensuring the required high safety of APP. Supplies of nuclear fuel for APP in our country as well as the return of spent fuel are ensured by long- term agreements and treaties with the Soviet Union. Thus, there is no need to reprocess spent fuel on the terri- tory of the GDR in order to extract highly active fission products and recover the uranium and plutonium not used in the reactor. This results in very visible streams of nuclear materials, which makes it possible for the IAEA, on the basis of the treaty of nonproliferation of nuclear weapons, to carry out effective inspection of the pureful peaceful use of nuclear technology in the GDR. Special railway cars and containers have been con- structed for the safe transportation of fresh and spent fuel. Low-level and medium-level radioactive waste formed as the result of the operation of the APP and the application of radioisotopes in many areas of the national economy are appropriately processed and then stored on the territory of the GDR. All of these successes would not have been possible without the close, fraternal cooperation with the Soviet Union and other socialist countries on the basis of bilateral government agreements within the frame- work of COMECON, especially in the domain of science and engineering. Already in the early stage of develop- ment of nuclear energy, in October 1960, COMECON established a Standing Committee on the use of atomic energy for peaceful purposes. It has ensured long and fruitful cooperation among the COMECON member- nations in mastering a new source of energy. The principal task of the Committee consists in coordinating the scientific research and long-term developments in the field of nuclear energy. To this end, working bodies were set up to deal with particular problems, especially the development of pressurized-water reactors, pre- paration for the introduction of fast breeder reactors, elimination of radioactive wastes, reprocessing of spent Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 fuel, as well as problems of radiation protection and protection of the environment. An international research team was created in Budapest to develop reactor physics and study related problems. The COMECON Standing Committee on Electrical Energy also set up a special working group of specialists engaged in the design, con- struction, and operation of APP with water-moderated-water-cooled reactors. Specialists of the GDR are participating actively in the activities of these international working bodies. In this connection mention should be made first of all of the creative teams of the Central Institute of Nuclear Research at Rossendorf, the Academy of Sciences, the Bruno Loischner APP, and the industrial group for the construction of electric power plants. The main efforts of the GDR in the further joint development of water-moderated-water-cooled power reactors and preparation for the introduction of fast breeder reactors are aimed at creating neutron-physics and heat-engineering programs for designing reactors, improving their water conditions and regulating and control systems, developing methods of deactivating an individual piece of equipment and the entire primary circuit of a water-moderated-water-cooled power reactor, methods of monitoring the state of the equipment and metal during inspection and operation of APP, as well as developing methods and special equipment for packaging, transporting, and storing radioactive wastes. This work,.as well as the creation of special methods for monitoring the sodium circuits of fast reactors, are conducive to a further increase in the safety, oper- ational readiness, and economy of APP. The GDR also participates in improvement of the designing, construc- tion, assembly technology, and start-up and loading operations in the construction of APP. On the 30th anniversary of the GDR it is with satisfaction that we note the successes of our workers, engineers, and scientists during the 23 years since the conclusion of the first intergovernmental agreement with the Soviet Union on the establishment of a nuclear-energy base in the country. They have earned our gratitude, as also have our Soviet friends, whose great professional knowledge and participation created the scientific, technical, and economic prerequisites for the introduction of atomic energy into the power industry of the GDR. NUCLEAR RESEARCH OF THE ACADEMY OF SCIENCES OF THE GERMAN DEMOCRATIC REPUBLIC IN THE LIGHT OF THE DECISIONS OF THE NINTH CONGRESS OF THE GERMAN SOCIALIST UNITY PARTY* The first report concerning the birth of the Academy of Sciences of the German Democratic Republic was inscribed with the motto "theoria cum praxi," framing a miniature portrait of Leibnitz on the cover of the journal Kernenergie. Such, too, was the sense of the second report, dated July 1, 1946, and called "enlistment of science in the construction of a democratic Germany." In the first years of nuclear research in the GDR controversies arose over whether "current issues" were a matter for the research and design organizations of the heavy power engineering industry whereas the activities of the Central Insitute of Nuclear Research should be devoted exclusively to future development of reactors. The controversies have long since ceased. Now everyone has taken to heart the words of the General Secretary of the Central Committee of the German Socialist Unity Party (CC GSUP) and Chairman of the Coun- cil of State of the GDR, E. Honnecker, that "socialism is sole appeal to science." The profound sense of this statement can be comprehended when one examines how in fact the ideal of unity of the economic and social policy of the Party is accomplished. Particularly large changes have occurred in the style of management and *Abbreviated translation originally from the German; ?1979 by Akademie-Verlag, Berlin. tChairman of the Scientific Councils of Foundations of Power Engineering and Microelectronics, Academy of Sciences of the German Democratic Republic. Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 149-151, September, 1979. 0038-531X/79/4703-0693$07.50 ?1980 Plenum Publishing Corporation 693 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 psychological climate of enterprises since the resolution adopted by the Ninth Congress of the SUPG stating that acceleration of scientific and technological progress is the basis for the intensification of production. The Academy of Sciences bears a dual responsibility: for the development of science as a source of new knowledge and for the effective use of its results. It is called upon not simply to link theory with practice but to ensure the most efficient performance of the social purposes of fundamental research. In 1975, when the program for physics research was being drawn up, we tried to answer this question. The answer can be for- mulated as follows : work on the development of the fundamental problems of physics and individual areas of physics makes it possible to discover and understand many new physical phenomena with great potentialities for practical use. The discoveries must be directed at solving the basic problems of society, primarily those of the national economy. This approach refutes the mechanistic view that only "appropriate" fundamental research should be engaged in. Also refuted was the demand that each line of research have an unquestionable economic purpose. How did we arrive at this answer and what is the crux of it? It was useful for us. to have the development of nuclear energy in a leading position among the principal areas of research. And this is no accident since the experience gained from the development of nuclear power should be used to solve other major problems. The development of the nuclear power industry in the GDR, therefore, has two tasks in the main. One of them is the result of agreements within the framework of COME- CON. The second stems from the need for such a scientific basis for the technological processes in the nu- clear reactor which will ensure safe and efficient operation of atomic power plants (APP). Accordingly, for this reason and also for the diagnostics, monitoring, and control of the processes reactor physics has been transformed from an exact science into one of the technological areas of research. This has occurred to the extent that the neutron-physical characteristics determine the engineering factors of APP. Thus, neutron characteristics are now studied not as a physical but rather as a technological phenomenon. Finally, the reactor installation, including the primary circuit, attracts attention as an object of research, as evidenced by the in- tensive exchange of specialists as well as instruments and technical information between the central nuclear research institutes and APP of the country. The knowledge that the information participates in the technological process is important for the exten- sion of the experience. The growing size and complexity of industrial plants of another type lead to the same problem. It cannot be resolved without a precise technique for data acquisition and computer processing of the data. At an exhibition devoted to the 275th anniversary of the Academy of Sciences of the GDR, the state of the research program at that time was illustrated schematically. This aspect of information science has beencon- firmed in part. Reactor physics, as a technological discipline, realizes the link between fundamental research and the technology of APP. However, the link between fundamental research and information science continued to be an "uninvestigated area." The gap between fundamental research and the use of the results of research was understood and the first steps were taken to eliminate it. Thus, we were not unprepared when we encoun- tered this problem in microelectronics. Microelectronics is of interest to us as a material carrier of complex informational processes and as a difficult branch of manufacture. It was first used to automate scientific experiments in the domain of nuclear research. In nuclear power engineering the application of microelectronics has taken the route of construction of hierarchical monitoring and control systems which are in accord with the technique of automating the devel- oping socialist society. In turn, nuclear research was conducive to the development of microelectronics. Thus, the many years of experience of the Academy of Sciences with ion implantation was used in the production of some microelectronic devices. The ion-beam technique still has many undiscovered technological capabilities which will also lend themselves to application. The Central Institute of Isotope and Radiation Research and the Central Institute of Nuclear Research have developed nuclear methods of analysis which are used in micro- electronics to solve problems of pure materials. In many cases only such methods are sensitive enough to permit a transition from materials "as pure as possible" to materials "as pure as required." Clearly, there are sufficient examples to show the effect of technology on research in the realm of physics. Technology is that mirror with which a variety of possible areas of development are focused on the most important directions, yielding the greatest effect from the intensification of production today and at the same time significantly im- proving the production technology of tomorrow. The nuclear research of the Academy of Sciences encompasses problems such as those below. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 1. Continued research on the fundamental laws of space, time, and matter. The physics of the nucleus and elementary particles makes a contribution to this work. The goal is to determine the limits of present- day knowledge. The point is to penetrate further into the microcosm and to create the instruments necessary to do this. In this connection one should not neglect the practical application of knowledge, methods, and in- struments which in many ways promote general process. Elementary-particle physics has been stimulated with the appearance of a possibility of experimental verification of the quark hypothesis. New currentresearch has been concentrated in the Institute of High-Energy Physics at Zeuthen and is being conducted in close co- operation with the Joint Institute of Nuclear Research (JINR). The work is planned so as to utilize the unique capabilities of the experimental facilities as Dubna and Serpukhov. Moreover, we can count on the use of the experimental capabilities of CERN. Theoretically validated problems should be subjected to critical analysis at JINR and obtain experimental confirmation with the aid of high-quality modern technology. Major results from such work were obtained at the end of 1978: an international team of specialists at Zeuthen built an ap- paratus for processing and interpreting photographs of particle tracks. In nuclear physics note should be taken of a developed generalized method of describing the mechanism of reactions on the basis of an exact many-particle theory. The high level of nuclear-engineering instrumen-. tation facilitated a center for the development of instruments for scientific research in the Central Institute of Nuclear Research. Interesting results on the dependence of nuclear decay on the chemical bonds were- obtained in the Cen- tral Insitute of Isotope and Radiation Research. Laws which were discovered will probably lend themselves to use in the development of nuclear-medicine preparations. These are only some of the results. Reviews published in Kernenergie in July 1975 to mark the 275th anniversary of the Academy of Sciences of the GDR can be recommended to the interested reader. 2. The next problem is that of studying complex physical structures of a natural origin. Of greater interest in this respect is work done at the boundary with other sciences, which was the case particularly in isotope and radiation research. Thus, research on isotopic effects in geochemical processes makes it possible to obtain interesting data about the history of elements and the origin of deposits, data which could be used in geological prospecting. The similarity theory can be used to model the combined processes and to determine their parameters. Researches carried out in the Central Institute of Isotope and Radiation Research with nitrogen-15 have attained a high level and have produced results applicable in biology, agriculture, and medi- cine. Radiation-chemical research has been enriched with fundamental work on the effect of irradiation on elementary processes which could find application in radiation-chemical chlorination of polyvinyl chloride and development of cable insulations. 3. An important place in the investigations of the Academy of Sciences of the GDR is occupied by the study of artificial physical structures, i.e., the fundamental laws of technology. Research for nuclear power engineering and microelectronics are the most prominent, but not the sole example of work in this area. Along with personnel from the Bruno Loischner APP further research is planned with other APP. Thus, within the framework of the international team in Budapest, established for preparations for the introduction of the VVER- 1000 water-moderated-water-cooled power reactor, specialists of the Academy of Sciences of the GDR are engaged in work on problems pertaining to diagnostics from noise analysis. Other problems for this team are being worked on in collaboration with specialists from the Rheinsburg APP. In conjunction with workers of the Scientific-Research Institute for Atomic Reactors (NIIAR) research is being conducted on fast reactors. In the realm of thermonuclear research two substantial projects were conducted in 1978. One of them was the con- struction by the Institute of Electronic Physics of an instrument for analysis of the interaction of plasma with a wall. Such an instrument was installed in the T-10 tokamak at the I. V. Kurchatov Institute of Atomic Energy. A considerable contribution to the solution of technological problems was made by research with isotopic tracers. The use of the results of the work for analysis of processes in chemistry, metallurgy, and coal dress- ing increased the efficiency of these processes. These are only some examples of work which has been done in recent years in 15 areas of research in the domain of physics. Basically, this work was aimed at ensuring energy, material, and information, as well as the development of instrumentation and health protection. The solution of numerous problems of nuclear research is impossible without international cooperation. Jointworkwith specialists of the Soviet Union means much to us. The successes mentioned above and the pro- blems formulated are at the same time an expression of gratitude for invaluable assistance. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 Faithful to. the behest of Ernst Thaelmann, we know on whose aide we stand in the class struggle. "Human failure" is the usual conclusion when western atomic power plants experience abreakdown. And the question is not asked as to why a human, endowed with talent and ability, proved to be such a weak link in the production sys- tem controlled by automatic devices that it is best to replace that human by a microprocessor. The purpose of our automation of APP is to provide the human with information and instruments for monitoring and control while relieving him of heavy physical and monotonous mental work. We proceed from the premise that atomic energy and microelectronics are indispensable aids in the construction of a socialist society. Atomic energy will never be used for the annihilation of mankind; this is the highest demand of our time. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 EFFECT OF NONUNIFORMITY OF FUEL DEPLETION WITH HEIGHT ON THE PHYSICAL CHARAC TER IS TICS OF A REACTOR A. M. Afanas'ev and B. Z. Torlin UDC 621.039.51 The development of the reactor for an atomic power station usually passes through several stages. The early stages are characterized by numerous variant calculations and economic estimates, in which an impor- tant role is played by the attainable depletion of the fuel (burn-up) S as a function of its initial enrichment, the parameters of the lattice, the construction of the fuel elements, the method used to cool them, the frequency of recharging with fuel, and so on. Comparatively simple relationships for estimating the burn-up are proposed below; they are of acceptible accuracy and they allow for the distorsion of the neutron field with height under steady conditions as a result of burn-up and boiling of the coolant. To refine these estimates and also to cal- culate the neutron field,itself, the isotopic content of the discharged fuel, and a whole string of other param- we describe a simple and effective method of numerical computation. Describing the neutron field by the diffusion approximation, we write the starting' equations in the form: k (z) + f (S, (p, z) Alp (z) = 0; * (0) _* (H) = 0, (1) where, in the one-group approximation, H = (d/dz) Dth(d/dz); A=1; [k (S, (p, z)-1JIM' (S, (p, z); =N (z), and, in the two-group approximation, d. , d 1 1. W d aZ -ti (z)' ti() ? _(0' 1 1 d D'` d_ 1 A-\0; 0) C 2' . th z LZ S z Lo dz d ( , ) k (S, (P, z)- n (z) ti (z) ; (N (z) The notation here is as follows : Dth(d) = Dth(d) (z, cp)/bth(d) (zo, 0) ; Dth and Dd, diffusion coefficients for ther- mal and delayed neutrons; zo, arbitrary point of the core; k, multiplication factor for thermal neutrons; S(N), fuel depletion (burn-up); cp(N), vapor content; M2, L, T, square of the migration length, the diffusion length, and the neutron age, evaluated at Dth(zo, 0) and Dd(zo, 0); N(z) and n(z), flux density distributions of thermal and delayed neutrons; and H is the core height. The dependence of k on the depletion S can be calculated as described. in [1]. The dependence of k on the vapor content qo we expressed in the following manner: k (S, (p, z) = k.(S, yo, z) -I- AkR (H, S) g (z) p (z)/ q) (H); W (z) E (S, z,) N (z,) dz,, hec 0; 0 < z < heC; g(z)=11; heX-z 1. 'These conditions are normally satisfied. The effect on S of end reflectors can readily be estimated using the approximate method. Replacing the, reflectors by an effective increase of size up to heff, we obtain: Sh) = S(0)" H+2heffcos rzheff (14) H H+2heff c1~,_ Radm). Before proceed- ing to describe the algorithm, we shall formulate a geometric-profiling problem somewhat more general than the one in [2] We must determine the optimal number mopt of consecutively joined therrmoemissive electricity-gen- erating elements (EGEs) and the optimal vector for the distribution of their lengths {1f)?Pt i E [1; m?Ptl, along an assembly of height H with n distributed parameters of the form Aj (z), / E [1; nl,which, for k given limiting values R,,, P E [1; kl, determining the resource, reliability, efficiency, technological quality, and other proper- ties of the thermoemissive assembly, will realize the maximum useful electrical power (or total efficiency) of the assembly. When the efficiency is maximized, the total thermal power is equivalent to the maximum useful electrical power for a given value of thermal power. It is possible to have a variant of the problem in which the assembly height H is optimized in addition to mopt and {1i}opt. The problem of geometric profiling, i.e., the actual determination of {il}opt, can be solved by various methods. Thus, the authors of [2] give a` variant of the geometric-profiling problem which is simpler but is most often encountered in practice, in which m is given, n=1, k=1, and, correspondingly, Aj(z) =q(z), Rt= Tee max. A simple but effective graphicoanalytical method for determining {li}opt enables us to solve the problem in the most general formulation, with some distributed parameters and any number of restrictions. It is based' on the use of V-q (voltage-heat) diagrams, which, for a given total current I, show in graphical form how the EGE voltage V varies with the density of the heat flux qF per emitter (or the density of volumetric heat gen- eration in the heat-generating core, qV), the length 1, the emitter temperature To max, and, if necessary, other parameters (the collector temperature Tc, the pressure PCs of the cesium vapors, etc.). Such functional rela- tionships for V (q, 1, Te, max) I= const are easily constructed from the volt-ampere characteristics of the EGEs Translated from Atomr_nya Energiya, Vol. 47, No. 3, pp. 169-172, September, 1979. Original article sub- mitted September 11, 1978. 718 0038-631X/79/4703- 0718 $07.50 ?1980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 424 , ti 4s, W/cm2 Fig. 1. Volt-ampere (V-q) diagram for profiling the length of the elements, taking account of the nonuniformity of heat generation and collector temperature and the restriction on the maximum emitter temperature (I=100 A). at constant thermal power I (V)qF, l ... and constant emitter temperature I (V) Te, max l .... The volt-ampere characteristics of the EGEs may be calculated with any accuracy by means of any algorithm. A typical V-qF diagram for determining {1i}opt for two parameters q(z) and Tc(z) distributed along the height of the assembly and restrictions only on Tee max for I=100 A are shown in Fig. 1. The volt-ampere characteristics of the EGEs on which the diagram is based were calculated on a computer by the algorithm used in [1], taking account of the distributed nature of the heat fluxes entering the emitter jacket, the nonisothermal and nonisopotential nature of the electrodes, the thermal and electrical losses, and the effect of the nonoptimality of the collector temperature Tc. In order to keep the figure clear, the diagram shows the variation of V (q, 1, Tc) I for three values of Tc: 900; 1000; and 1100?K. The working diagrams contain more detailed information on Tc. The coincidence of the characteristics for Tc=900?K and 1100?K is a result of the experimentally observed specific variation of the isothermal thermoemissive transformations (TET) with Tc [10] which was used in calculating the EGEs. From the V-q diagrams for specific values of qi, Tci, and if necessary, other values of Ai, we determine the li which for a given I yields the maximum value of Vi on condition that Tee max < Te, adm and the other Rj Radm? Knowing all the Vi, we determine the total voltage V a = E Vi, ? the total electrical power Wa = Val, and the efficiency of the assembly with the resulting {li}, If the dimensions li can vary continuously, then the profiling is carried out in such a way that Te, max = Te, adm i E [1, m]. This actually means that if we have only q(z), li is selected along the curve Tee max(q) = Te, adm, if we have the nonuniformities in q(z) and Tc(z), it is selected along the surface Tee max(q, Tc) = Te, adm, etc. If for a given {1i} we obtain an assembly height H* = Z (la + lc),where lc is the height of the switching connector, and this assembly height does not coincide with the given H, then we must make a correction to {li} in such away that H H*. This can be obtained by repeated determination of the vector {Ii} from the V-q diagrams for another value of m, another absolute q(z) (the rela- 719 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 tive distribution of q(z) remains as before), or a reduction in the length of some of the EGEs (usually at the edges of the assembly) in such a way that H - H*. It should be noted that in the last case the shortened EGEs will operate at Teo max< Tadm? Naturally, in the absence of other considerations, we should select from the above methods the one which yields the maximum target function. When we have thus determined ill} for some values of I, we can construct the volt-ampere characteristic I(Va) Te, max < Te, adm' from which we can readily .select. the point, and consequently the fli}opt as well, for which we reach the maximum value of electrical power and satisfy restrictions of the form Rp s Radm, At the same time as we optimize fli}opt, we also optimize the number of successively joined EGEs in the assembly, mopt, and its thermal power. Geometric profiling leads to some redistribution of the fuel along the height of the assembly, since for {li}opt the central EGEs are shorter than the peripheral ones, which must be taken into account in calculating the q(z) and Tc(z), and also in estimating the critical parameters of a system with profiled multielement assem- blies. Let us briefly consider some results obtained by means of the above method. It has already been noted that Te, max is restricted; even a slight nonuniformity in q(z) leads to a sub- stantial reduction of the electrical power W(Kz) in comparison with the power of an assembly with the constant heat generation, WO - W(Kz = 1) [1-3]. Thus, the authors of [1] obtained an empirical formula for determining the relative power of an assembly with elements of identical length for a cosinusoidal law of distribution of q(z) : wq = W (K,)/W, :s 2,52 - 1,52K;, which, e.g., when Kz = q xC =1,25, yields wq - 0.6. The individual EGEs operate at a Tee max almost 400?K higher than the maximum admissible value. Geometric profiling of an assembly with nonuniform q(z) and the restriction Te, max '< Te, adm+ for con- tinuous variation ofli, enables us to obtain electrical power values which are only a few percent lower than the power of an assembly with constant heat generation. Analogous results were obtained in [2, 7]. For a restricted number of typical dimensions of the EGEs, we also observe an increase in the output power of the assembly in comparison with an unprofiled one. However, since in this case some of the EGEs have a lower value of Te, max, the power of the assembly is found to be lower than for continuous variation of 1. If we assume dis- crete variation of 1 by a value which is a multiple of 0.5 cm, for the same conditions we have wq Pt 0.85 when the Te, max values of individual EGEs differ from Te, adm by only 40?K. The nonuniformity of Tc(z) for constant q(z) but with the restriction Te, max'` Te, adm also leads to con- sidpetrable losses in electrical power W(Tc) in comparison with the power of an assembly with constant Tc = TA Thus, for Tc, max-Tc, min 150?C and Tc, min= Tcp'c = W (Tc)/Wo ~ 0.8 [1] with a difference of more than 100?C in the Te, max of individual EGEs. Geometric profiling with continuous variation of 1i increases we to 0.95 out of the optimal power value when all the EGEs have constant Tee max values. For same condi- tions when we have only two typical dimensions of the EGEs, we -- 0.9. The simultaneous effect of q(z) and Tc(z) on an unprofiled assembly yields an even lower relative power value, wq, c= W(Kz, Tc)/Wo, where WO is the power of the assembly when Kz = 1 and Tc(z) = TgPt, i.e., even lower. For the conditions considered above, wq,c 0.55. Geometric profiling which takes account of the effect of q(z) and Tc(z) enables us to increase wq,c to - 0.9. When it is possible to have only discrete variation of 1 (by 0.5 cm), wq,c se 0.72. Thus, the algorithm worked out above enables us to determine {li}opt in a relatively simple manner for some parameters arbitrarily distributed along the height of the assembly and taking account of the restricted number of typical EGE dimensions that are actually possible. An important advantage of the method is that it can take account of any factors restricting the resource and operating capacity of the assembly. To do this, other boundary curves are drawn on the V -q diagrams in addition to the Te, max isotherms. The determination of the fli}opt is carried out in an analogous manner, but with all the Radm restrictions taken into account. 1. Yu. A. Broval'skii et al., Teplofiz. Vys. Temp., 13, No. 1, 171 (1975). 2. V. M. Dmitriev and V. A. Ruzhnikov, Preprint FEI-704, Obninsk (1976). 3. Yu. Ya. Kravehcnko and G. A. Stolyarov, Preprint IAE-1579, Moscow (1968). 4. E. S. Bekmukhambetov et al., At. Energ., 35, No. 6,-387 (1973). 5. B. A. Ushakov, V. D. Nikitin, and V. Yu. Korbut, At. Energ., 31, No. 5, 467 (1971). Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 6. E. S. Glushkov and N. N. Ponomarev-Stepnoi, At. Energ., 20, No. 6, 478 (1966). 7. E. Wolf and W. Haug, Atomkernenergie, 16, 213 (1970). 8. A. Schock, in: Proc 3rd Int. Conf. of Thermionic Electrical Power Generation. Julich, FRG (1972). 9. V. A. Kuznetsov et al., At. Energ., 36, No. 6, 450 (1974). 10. B. P. Baraksin et al., in: Reports of Soviet Scientists at the Second International Conference on the Thermoemissive Transformation of Energy [in Russian], VNIIT, Moscow (1969), p. 231. FISSION NEUTRON DETECTORS Z. A. Aleksandr.ova, V. I. Bol'shov, I. E. Bocharova, K. E. Volodin, V. G. Nesterov, L. I. Prokhorova, G. N. Smirenkin, and Yu. M. Turchin The average yield of neutrons per fission event u and their energy distribution N(E) are fission neutron characteristics that belong to the category of fundamental constants of breeder materials in reactors. Most practical problems can be solved using the well-known approximation of the fission neutron spectrum by the Maxwellian distribution X (E, 0) _ (2/ a03) I /T exp (- E/0), (1) i.e., N(E)=-vX(E, B). In such a case, knowing the average energy 3 OA, or the so-called neutron temperature ItT, it is possible to find the entire fission neutron spectrum. Thus, most experimenters focus their attention on the determination of the two first moments of the distribution N(E) : the zero moment v and the first moment uc (precisely speaking, their ratio e). These fission neutron characteristics, and especially j, are most frequently measured with the aid of detectors consisting of a hydrogeneous moderator and slow-neutron counters. The moderator is usually poly- ethylene and the slow-neutron detectors are BFI or 3He counters. In such detectors, which in contrast to detectors that detect single events of microscopic interaction of fast neutrons with nuclei are called macro- scopic detectors, neutrons live for tens of microseconds while being slowed down and scattered. Since the moderation length \essentially depends on the energy of fast neutrons entering the moderator and since the spatial distribution of slow neutrons is a strong function of t/V(t being the distance to the neutron entry plane), the energy sensitivity of a slow-neutron counter can be varied within wide limits by changing its position in the moderator, its orientation with respect to the beam of incident neutrons, or the moderator con- figuration. Such an approach is frequently employed to fit the characteristics of the detecting system to the needs of the particular problem. Well-known examples of such detectors are the all-wave (long) counter [1], the isodose neutron detector [2], the Bramblett multispherical spectrometer [3], etc. Here we report on certain new applications of the macroscopic method to the measurement of fission neu- tron characteristics. Three versions of the method are discussed: a macroscopic fast-neutron spectrometer (E detector), simultaneous measurement of the average yield and average fission neutron energy (vE detector), and a detector which measures v without being sensitive to the average fission neutron energy (v detector). Energy Dependence of Slow-Neutron Counters in Polyethylene Moderator The results reported in this paper are based on an investigation of the dependence of the sensitivity c(E, tn) of slow-neutron counters in a polyethylene block on the energy of fast incident neutrons E and on the distance from the neutron incidence surface tn. Although the measurement of these characteristics. over an energyrange from 0 to 15 MeV is rather difficult, no such, difficulties exist when the characteristics are calculated theoreti- cally. We have thus studied the sensitivity function e(E, tn) by calculating its relative behavior by.the Monte Carlo method and by normalizing it experimentally at several points taking into account the individual charac- teristics of the counters. Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 172-176, September, 1979. Original article sub- mitted May 17, 1978; revision submitted November 17, 1978. 0038-531X/79/4703- 0721$ 07.50 ?1980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 -90& 10 ? 5th 10u, . b 10-1 100 ?? ? 6th .10-1 y'4, ? . ~-e~ t tad 2nd ' i A 1r? 4 .8 1091 I .4th S* 10-31 1 1 1 I I I 1 1 4 1 1 I I I l i i 1 1 :1 I I I 1 9 , 1 1 1 4 4 .. I 1 I I III 1 ( ) l0 41 1 10.. E,MeV Mev Fig. 1. Compar1s:on of .results obtained by the Monte .Carlo method {4) ..and experimentally (0) for the function ,,z (E) [cm?] for one ,type- SN10 NA counter. Two basic "counter+moderator" arrangements are discussed: for measurements in the so-called good geometry when the detector registers source neutrons within a small solid angle, and for experiments in 47r geometry in which the detector monitors a considerable portion of space around the source. Good geometry is usually employed for studies of fission neutron spectra :[3-6] and 47r geometry, for v measurements [7]. The first arrangement is most suitable for measurement of the sensitivity function e(E, tn) with mono energetic neutrons [6]. 47r detectors usually have a central channel reserved for a fission-fragment detector in coincidence with whose output pulses neutrons are registered by the coaxially positioned neutron counters. The measured and calculated values of the function en(E) =,e(E, tn) for a rectangular detector are shown in Fig. 1. The Monte Carlo procedure used in the calculations is described in [8]. Measurements were carried out for monoenergetic neutrons and for neutrons of radioactive (a, n) sources having a continuous energy spec- trum. The source of monoenergetic neutrons were T (p, n), D (d, n), and T (d, n) reactions taking place in an electrostatic generator with solid targets. Neutron yield was monitored with a thick-walled fission . chamber with a 235U layer placed near the neutron target within the same solid angle as the detector .being, calibrated. The measured F-n(E) functions were normalized with the aid of radioactive neutron sources consisting of .a homogeneous mixture of 231pu with Li, F, B, and Be using the expression (4nR2/Q) MI,,.= e? (E.)'[c0l. where R is the distance between the source and the detector front surface (in the calculations. and in-a:ll:experi- ments R = 100 cm). The source yield Q was determined to within -' 10%. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 The measured and calculated results (Fig. 1) are seen to be in good agreement with each other indicating that the sensitivity of slow-neutron detectors depends strongly on their position within the moderator. The irregular structure of en(E), associated with resonances of the cross section of neutron scattering by carbon nuclei, manifests itself clearly at great depths tn,'viz., when n >3. The behavior of sensitivity en(E) weakly depends on the counter dimensions and on the distance R to the moderator surface as can be seen by comparing the functions F-n(E) calculated for a prism and for a 41r detector in which these parameters differ'considerably. This is very fortunate from a methodological point of view since it allows one set of en(E) curves to be used without taking into account in the first approximation thegseo- metric features of the detection system. It should also be noted that the resonance structure of &n( ) prominent in 47r detectors because of the large spread of neutron "ranges" in the resonator. Similar tables of the function en(E, tn) are given in [5] for a prism and in [7] for a 47r detector. Stacked Macroscopic Spectrometer. (E Detector) The investigated detector has several advantages over the multispherical detector [3]. First of all, with a stacked structure there is no need to change the moderating envelopes and an entire set of readings for dif- ferent moderator depths can be obtained simultaneously. Secondly, a stacked structure imposes no such strict constraints upon the dimensions of slow-neutron counters as a multispherical spectrometer making it possible to increase the sensitivity by one order of magnitude or more. Finally, in a stacked-counter spectrometer it is much easier to obtain an extensive set of en(E) functions. Thus, the family of en(E) curves of the multispherical .spectrometer is covered by the characteristics of the first four rows of a stacked spectrometer. In other words, the later covers a wider dynamic range which is shifted towards higher energies, and this is one of the main factors that determine the accuracy of measurement of the parameter 0. Let us consider the last problem in some more detail. Considering that the difference I DO I = I O - Oo I 103 neutronsjsec. In many cases this condition can be satisfied only when a considerable amount of fissionable material is available, which appreciably restricts the experimental possibilities of the method. By increasing the,.method sensitivity by -several orders of magnitude, the 4'7r detector eliminates these difficulties. Secondly, because of,the energy dependence of the neutron detector efficiency, the results of measurement of the fundamental nuclear physics constant T must be corrected for the difference between the fission-neutron spectra of .the me'asured?isotope and the standard. The measured experimental ratio of the number, of neutrons :registered per one fission event.,can be written as V p=- V0 .~ X (E, 0) s,, (E) dE ? ="k I 'R (E. 8u) P. (E-)-d9 0 where k is the correction mentioned above, and (E) is the detector efficiency. Using expression (2),?the factor k can be written as AO- g k=i + o ,n+ 902 yr[+..., (5) Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 indicating that its accuracy depends in a large measure on the ratio 0/0o. Unfortunately, this ratio is not very reliably known for thermal-neutron fission of even the most common reactor materials. In some cases the error of the correction factor is comparable with the correction proper I k -1 1. Simultaneous measurement of v and 0 eliminates this difficulty. The method of simultaneous measurement of ti and o is implemented by placing in the moderator several concentric rows of counters and recording coincidences of their output pulses with the fission-fragments detec- tor pulses. The ratio of the number of coincidences is used for finding o and their sum, for measuring v. Detector for Measuring v Irrespective of the_ Average Fission Neutron Energy E (p Detector) Above we have discussed a method of relative measurement of u in which the correction for the differ- ence between the fission-neutron spectra of analyzed object and the standard is determined experimentally. The dependence of the parameter xn on the location of counters in the moderator [7] makes it possible to design a detector for which no correction is needed. However, the point is not a strict observation of the condition k= 1, which is true only for all-wave detectors [e(E) =const], but the approximation (6) x=0 Or (E)=E, to which corresponds the energy-dependent efficiency of the detector. The solution of (6) can be easily found from the function x(t) for a 41r detector represented in Fig. 2. It is seen from the figure that condition (6) will be satisfied if the counters are placed at a distance Copt=10.5 cm from the inside surface of the moderator. Figure 2 also shows the parameter y, which defines the second-order term contribution in expression (5), and the difference between the factor k and unity for x=0, as functions of the distance t. It is seen that in the neigh- borhood of t=toptt 0.5 cm, the correction factor lk-11 is less than the error in vo for the standard 0.3%O) A0/0o typical of the range of B for heavy nuclei. The realization of the condition x=0 by placing the counters precisely on the surface t=topt is the sim- plest but not the only and best one since it limits the number of counters. One can make use of the fact that x has different signs to the right and left of t=to t, place the counters on both sides of the optimal surface, and complete the row of counters without violatingthe condition x=0. Detectors for measuring v described in literature satisfy rather the demand of maximum efficiency. Figure 2, in which the top curve represents the integral efficiency to fission neutrons Tn (t) = f x (E, 0) X F (E, t) dE, indicates that this demand does not coin- 0 CONCLUSIONS The methods described in this article have been developed for studies of fission-neutron spectra. The range of problems that can be solved with their help can be greatly expanded. For example, a 41r detector sim- ilar to the one described here has been used to separate prompt and delayed fission neutrons and (y, f) and (y, n) neutrons [101. Although macroscopic spectrometers are most effective in situations when the energy distribution can be represented in an easily parameterized form, their application is not at all limited to such cases. In the gen- eral case, the problem is solved using a group description of the distributions and reduces to a solution of a system of linear equations with experimentally determined left sides M. LITERATURE CITED 1. A. Hanson and J. McKibben, Phys. Rev., 72, 673 (1974). 2. Kh. D. Androsenko and G. N. Smirenkin, Prib. Tekh. Eksp., 5 64 (1962). 3. R. Bramblett, R. Ewing, and T. Bonner, Nucl. Instr. Methods, 9 1 (1960). 4. V. I. Bol'shov et al., Preprint FEI-578, Obninsk (1975). 5. Z. A. Aleksandrova et al., Preprint FL-866, Obninsk (1978). 6. V. I. Bol'shov et al., in: Proc. of the Conf. on Neutron Physics [in Russian], TsNITatominform, part 4 (1977), p. 290. T. V. L Bol'shov et al., Preprint FEI-865, Obninsk (1978). 8. I. E. Bocharova, L. I. Prokhorova, and G. N. Smirenkin, in: Nuclear Constants [in Russian], TsNllatom- inform, Moscow (1974), p. 7. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 9. V. L Bol'shov et al., in: Proc. of the Conf. on Neutron Physics [in Russian], TsNllatominform, part 3, Moscow (1977), p 284. 10. J. Caldwell and E. Dowdy,. NueL Sci. Eng., 53' 767 (19.75)';. B. Berman and S. Fultz, Rev. Mod.. Phys., 47, 713 (1975). ANALYSIS OF THE RELIABILITY OF. RADIOCHEMICAL PLANTS WITH ELECTRON ACCELERATORS V. M. Eshnyaskin and Yu., D. 1{ozl:ov UDC 21.384.6:541.15 The reliability of radiochemical plants must be calculated at an early design stage [1=3]. The reliability of such plants significantly affects their economy during operation. Besides radiation physics parameters, the design of radiochemical plants requires the knowledge of the reliability of their units and elements. Calculations of structural reliability are now an integral part of design in various fields of engineering, e.g., in reactor design [4].' The aim of this paper is the calculation of the reliability of component units of planned radiochemical plants with high-current electron accelerators and the evaluation of plant reliability as a whole. Since such plants are complex systems designed for long operating times, their reliability should be calculated in several stages [3]. The reliability of electron accelerators, radiochemical apparatus, and other technological and auxiliary equipment is calculated first. After the reliability indicators of these units have been analyzed and evaluated, the reliability of the radiochemical plant as a whole is calculated. The reliability of units composed of many components and linked by complex functional relations is calculated by the Monte Carlo method [1]. The com- binatorial method [3, 4] is used to calculate the reliability of less complex units. To analyze the reliability of radiochemical plants the latter are arbitrarily divided into individual units or elements in accordance with the following principles : the failure of an element causes breakdown of the entire plant; an element is a relatively independent functional or structural unit; The number of elements in a plant should be minimal (if, e.g., the reliability indicators of both a plant unit and of its individual components are known, the calculations should be based on the entire unit) ; . if no reliability data are available, elements and units are combined if possible into one or two larger functional units whose reliability indicators are specified as a set of values for the given range of possible mag- nitudes. The reliability of each unit is calculated for all assumed values. The reliability of radiochemical plants is analyzed with the aid of functional block diagrams of radio- chemical apparatus and accelerators that specify the effect of failure of an element (unit) on the reliability of apparatus or accelerator. To simplify the problem it is desirable to consider systems with instantaneous recovery in which the times of failure and recovery coincide [5]. An algorithm and program are designed for models with instan- taneous recovery which compute the probability of no-failure operation P(t),, the failure flow parameter w(t),, the mathematical expectation of the number of failures in time t H(t), and mean time between failures T [5]. Simulation of the operational process of a radiochemical plant with an electron accelerator consists in the following. The plant operation is considered either during its entire operating time or only during the principal operation period. The chosen operating time is split into intervals At and the flow of recovery of failed ele- ments is implemented with the aid of random numbers. The duration of time intervals (At) should not be too long, on the one hand, so that typical variations of the flow are not smoothed, and not too short, on the other hand, so that insignificant properties of the flow are not manifested. Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 176-179, September, 1979. Original article sub- mitted June 26, 1978; revision submitted February 5, 1979. 0038-631X/79/4703- 0726$07.50 ?1'980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 Fig. 1.. Schematic drawing of radiochemical plant with electron accelerator : 1) high- voltage generator; 2) high-voltage cable; 3) vacuum system; 4) irradiator; 5) radiochemical apparatus, (conveyer) ; 6) gas supply system. It is known [61 that the input data on which an algorithm is based are the failure rate and the failure flow parameter (for nonrepairable and repairable units respectively) when the failure distribution is assumed to be exponential, and the mean time between failures, the time to failure, and dispersion when the assumed failure distribution is normal or lognormal. It has been shown in [71 that failures of the cathode subassembly, the vacuum system, and the exit window have an exponential and normal distribution. The probability of no-failure operation.of these units has the form P (t) = e" F (T - t/v), (1) where t is the operating time, h; 1, failure rate due to transient defects, 1/h; T, time between failures due to wear, h; and v, dispersion of the distribution of times between failures due to wear. In simulation, the failures of elements and units caused by transient defects and wear are assumed to be a superposition of two parts: one with normally distributed and the other with exponentially distributed failures. Since the failures are independent, the exponential and normal parts are simulated separately. The model con- siders flows of failures of all units that enter into the failure flow of the radiochemical plant as a whole. Thus, mathematical simulation consists of the following steps : repeated application of an algorithm describing the probabilistic model of the investigated process in individual units of the radiochemical plant; statistical processing of the obtained results and their analysis; calculation of the reliability of radiochemical apparatus, the electron accelerator, and of the plant as a whole (if necessary) ; setting up tables and graphs with recommendations as to the calculation of reliability of other similar units. As an example of the application of the above technique consider the analysis of the reliability of a radio- chemical plant with a high-current electron accelerator used for processing lumped (unmixed) systems. The radiation section consists of the electron accelerator, the radiochemical apparatus, and auxiliary equipment* (Fig. 1). The kinetic energy of accelerated electrons is 0.08 pJ and the total irradiator current behind the exit win- dow is about 50 mA. The high-voltage transformed [81 connected by a cable to the irradiator operates in a gas *The reliability of only the radiation section of the plant is discussed. The technological equipment is the topic of another article. TABLE 1. Reliability Parameters of Radio- chemical Plant with an Electron Accelerator Reliability parameter Unit w (t), 1/h T, h High-voltage generator 1,7.10-3 600 4,2"10-4 2400 Gas supply system 18,8.10-5 5300. 5,3.105 18000 Irradiator 4,6.19-5 21650 3,9.10-5 25500 Conveyer 43.10-4 3000 3:3.10 4 f Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 P(t) 1,0 0 500 1000 1500 1000 250t1 t I Fig. 2 H(t) 50 50! 40 30 20 10 I . i I I I- 1 1000 2000 3000 4000 t,h Fig. 3 Fig. 2. Probability of no4ailure operation of certain units of the accelerator and con= veyer: 1) irradiator; 2) gas system; 3) conveyer; 4) generator. Fig. 3. Mathematical expectation of the number of failures H(t) in time t of the radio chemical plant with electron accelerator. medium (electric gas). To take into account the reliability of operation of individual (smaller) components, the irradiator has been arbitrarily divided into four sections: the cathode subassembly, the vacuum system, the exit window, and the scanning system. The irradiator is located in a special radiation shield which can be dis- mounted in parts for servicing. From the point of view of construction and number of components, the control console is similar to that of a typical radiochemical plant with an accelerator "Electron-311 whose reliability has been analyzed before in [7]. The conveyer is an electromechanical subassembly placed in an individual radiation shield. Its task is to transport in and out products to and out of the irradiation zone and consists of driving and tension drums, electric motor, various clutches, reducing gears, sprockets, chains, etc. All elements of the plant and irradiator within the shield are in some measure affected by the action of electron and bremsstrahlung radiations and ozone. Thus, in analyzing the reliability of these units we have taken into account the possible radiation effects on these elements. For .this we have calculated the y radiation exposure dose rate for various parts of the structure within the biological shield. Using the data on radiation resistance of the materials and instruments [9] of which the radiochemical plant and irradiator are constructed, we have analyzed their usefulness for the given operating conditions and tentatively estimated the useful life of these elements equal to the time during which the allowable dose is absorbed. The maximum and minimum failure rates for the components of the high-voltage generator, the gas sup- ply system, and the radiochemical apparatus (conveyer) have been adopted from data published in [4, 10-191i Table 1 lists the calculated reliability indicators of these units from which the curves of the probability of no- failure operation shown in Fig. 2 have been plotted. To evaluate the reliability of the plant as a whole, the results where supplemented by the reliability indicators of the cathode subassembly, the exit window, and the control console (Table 2) published before for "Electron-3" radiochemical plant [7, 18]. The data in Tables 1 and 2 have been used as a basis for calculating the reliability indicators of the plant; operating time to failure, the mathematical expectation of failure in time t, and the failure rate under steady- state operating conditions. Considering the accuracy of statistical data used for calculating the 'reliability indi- TABLE 2. Reliability Parameters of Certain Accelerator Components Reliability parameter unit' h (t), tih oat). i/h T. Cathode subassembly 5;9.10-9 - 170 i,0.10-; 1000. Vacuum system - 5,9.10-4 2000 Exit window 1,0.10-2 100 7,0.10-9 150 Control console - 8,3.10-4 1200 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 cators, the failure rate parameter of the system as whole was found to be 0.3 to 1.3 ? 10-' per hour,: The cal- culations were carried out by repeating the algorithm 1000 times with a time-to-failure error of about 3 h. The dependence of the mathematical expectation of the number of failures on operating time is shown in Fig. 3. The figure indicates that 25 failures can be expected to occur in the system in 2000 h and 63 failures, in 5000 h. Thus, we have for the first time applied the method of mathematical simulation to the evaluation of the reliability of electrochemical plants at the design stage. The reliability indicators of component units of the plant and of the plant and electron accelerator as a whole have been calculated a priori. The results can be used for comparing the quantitative reliability indicators of individual units and (if necessary) for taking mea- sures to improve their reliability. The obtained data can also be used for calculating the necessary margin of safety and for scheduling preventive maintenance of individual units or plants as a whole. LITERATURE CITED 1. V. M. Kshnyaskii et al., in: Proc. of All-Union Conf. on the Application of Particle Accelerators in the National Economy [in Russian], Vol. 1, Izd. NIIEFA, Leningrad (1976), p. 232. 2. V. M. Kshnyaskii, Yu. D. Kozlov, and L. V. Popova, in: Abstracts of Papers of the All-Union Scientific Engineering Seminar on the Application of High Power Sources of Ionizing Radiation in Radiation Engi- neering {in Russian], VNIIRT (1976), p. 126. 3. Yu. D. Kozlov, K. L Nikulin, and Yu. S. Titkov, Calculation of Parameters and Design of Radiochemical Plants with Electron Accelerators (Handbook) [in Russian], Atomizdat, Moscow (1976). 4. A. L Kiemin, Engineering Probability Calculations in Nuclear Reactor Design [in Russian], Atomizdat (1973). 5. L. G. Gorskii, Statistical Algorithms for Reliability Calculations [in Russian], Nauka, Moscow (1970). 6. B. P. Kredentser et al., Solving Reliability and Maintenance Problems with General-Purpose Digital Com- puters [in Russian], Sov. Radio, Moscow (1967). 7. Yu. D. Kozlov, At. Energ., 39, No. 4, 280 (1975). 8. E. A. Abramyan and V. A. Gaponov, At. Energ., 20, No, 5, 385 (1966). 9. N. A. Sidorov and V. K. Knyazev (editors), Radiation Resistance of Construction Materials in Radiation Engineering (Handbook) [in Russian], Sov. Radio, Moscow (1976). 10. A. M. Polovko, Principles of the Theory of Reliability [in Russian], Nauka, Moscow (1964). 11. B. S. Sot-skov, Methodical Instructions and Reference Data for Calculating the Reliability of Components and Systems [in Russian], Moscow Aviation Institute {1964). 12. B. S. Sot-skov, Principles of the Theory and Calculation of the Reliability of Components and Systems in Automation and Computers [in Russian], Vysshaya Shkola, Moscow (1970). 13. V. V. Akulov et al., see Ref. [1], p. 115. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 NEW BOOKS E. P. Anan'ev ATOMIC PLANTS IN POWER ENGINEERING* Reviewed by Yu. I. Koryakin Monographs published by Atomizdat, which cover the various aspects of nuclear power engineering, go out of print quite rapidly indicating the growing interest to nuclear power industry. The reviewed book will help in satisfying this interest. The book treats its subject on various levels and cannot be simply characterized. Nevertheless, one can distinguish two special features : 1) the material is based on experience gained in the Soviet Union and 2) the main stress is on advanced nuclear power technology. Both these qualities are attractive and important espec- ially in conditions of intensive nuclear power plant construction. Another important topic of the book is radiation safety which can be provided by a set of measures which are consistently .and with deep insight presented by .the author. The author focuses his attention on channel and vessel reactors. Their evolution,. present state, and .tech.- nological problems concerning such reactors are the main subjects of seven (out of eight) chapters of the :book. 'The author notes that large-scale solutions of problems make it possible to achieve new technological and :eco- nomical levels of operation with these.reactor types (especially with channel reactors), while on the other hand generating new technological problems. The author makes an attempt to estimate the scale and significance of certain socioeconomical factors associated with the development of nuclear power engineering. Unfortunately., this subject is barely touched upon. In view of its importance, the subject attracts widespread attention and heated discussions. Certain inaccuracies in the treatment ;of reactor physics must be mentioned. Possibly, they result from the .attempt of the author to simplify :the discussion and make it intelligible to less trained -readers. The book leaves ;a favorable "impression. Notwithstanding certain min or weak .points, .the ;book is a valu- :able addition to .the shelf of nuclear :power :.engineering literature. *Atomizdat, Moscow (1978), 190 pp., 1 ruble 70 kopecks. Translated from Atomnaya Energiya, Vol. 47, No. 3, p. 179,. September, 1979. 0038-631X/79/4703-0730$07.50 ?1980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 LETTERS TO THE EDITOR EVALUATION OF THE SELECTIVITY OF ELECTROCHEMICAL REACTOR-FUEL RECOVERY ON THE BASIS V.~ A. Lebedev UDC 669.536.7 The separation factors of uranium and rare earths, uranium and zirconium, uranium and plutonium, and uranium and thorium have been determined from thermodynamic data. The elements listed above have similar oxidation-reduction potentials in chloride melts and thus determine the efficiency of electrochemical recovery of reactor fuel. The separation factor Q was calculated from the expression [1] 1g Q = (n-m) FED- mFE2 -nFEi +lg yt ? (1) 4.575 T 72 where E*,.Ez are relative standard potentials Met/Mel , Me2/Mem+ [2, 31; y1, y2 are the activity factors of Mel and e2 in a liquid metal electrode [4]. If n=m, the separation factor is independent of the melt potential E. For n ;e m the calculations were made for two-phase (L+compound) melts of electropositive metal (Me2) with a 1 mole % concentration of its ions in the electrolyte (co. The comparative efficiency of various solvents salts used in separation was evaluated from Ig Q/Q'= nF (E, -Er-Er?Er), (2) 4.575 T The error in log Q estimated from Eq. (1) is t (0.3 to 0.6) and from Eq. (2), t (0.2 to 0.4). The initial data and the obtained results are listed in Tables 1 and 2. The separation factor of uranium and lanthanum increases regularly when the solvents are light metals located higher and to the right in the periodic system. The factor is close to one for thallium electrodes, 10_102 TABLE 1. Separation Factors of Mei and Mee in Liquid Metal Me-KCI melt-LiCl System Containing Me1+ and Mem+ Ions 1gQ=A+BT-1+nm Igc2 A I -B Zn 5,05 12475 2,44 3685 2,14 586 750 535 Cd 4,16 9738 3,03 885 0,66 523 20 15 Al 01 3 9892 3,84 7505 -1,30 6994 5.105 Ga , 4,72 13333 1,38 4984 2,87 1023 1,4.10+ 8.103 In 73 2 9508 2,36 2033 -0,10 1901 189 63 TI , 2 36 8153 0,49 -2678 1,40 -1455 0,4 0,9 Sn , 4 26 13377 1,97 4546 1,82 545 318 232 Pb , 07 -0 6618 2,97 2360 -3,51 5118 770 41 Bi , 94 0 11158 1,05 4328 -0,53 2546 400 92 Zn , 5 38 9770 2,44 3635 -t,03 2489 107 25 Cd , 4,44 6354 3,03 885 -2,61 3105 19 .3 Al 26 3 9050 3,84 7505 -4,60 7029 270 Bi , 40 1 8284 1,05 4328 -3,67 4618 121 9 U3+ Zr4+ Zn * , 0,56 0,374 7,66 10164 -2,26 8150 2.108 2.108 Th4+ U3+ Zn* 4,78 9511 0,79 0,559 -0,03 -213 0,10 0,11 A1* 83 3 9470 0,65 0,307 -1,20 3002 14 Ga* , 29 2 8710 0,53 0,195 0,33 822 5 3 In* , 77 0 3850 0,43 0,256 -2,52 2028 0,24 0,10 Sn * , 0,46 6459 0,67 0,288 -2,60 3123 4,4 0,8 Pb* 1,14 ? 3620 0,41 0,253 -2 , 74 2457 0,5 0,1 Sb * -0,02 9830 0,82 0,226 -1,24 2773 8 Bi* 1,33 7430 0,475 0,157 -0,11 901, 2,8 1,4 *The column log yU shows for these systems the coefficients of the equations E2=A+ B 10-8 T [4], where E2 is the emf between the metal and its two-phase (L+compound) melt. Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 180-181, September, 1979. Original article sub- mitted February 13, 1978. 0038-631X/79/4703-0731$07.50 ?1980 Plenum Publishing Corporation 731 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 TABLE 2. Ratio of Separation Factors of the Elements Mel and Me2 Obtained with KCl-NaCI Melt "(Q) and Other Solvent Salts (Q') el + M Mem+ =A+ B?T M e0l E = A +,B4 ig Q/Q' = A + B?T-1 Q/Q' -A B?104 -A 1 B-104 A B 1000 1100 [ Ua+ Zr4+ LiCI NaCI 2,83 6,0 2,40 6,0 --------------- 0,30 -1227 (1 12 0 5 NaC1-KCI 2,99 3;01 6,7 6,6 2,58 2 66 6,7 -6,8 0,30 0 00 -907 0 0 , 0,25 , 0,30 KCl SCI 10 3,12 7,1 , 2,82 -7,7 , -0,61 , 0 756 1,0 1 4 1,0 1 2 pu3+ U3+ C I 3,18 346 , 7,4 9,4 2,;88 283 7,7 6 0 -0,14 1 06 756 907 , 4,2 , 3,5 NaCI 'LiC1-KC1 3 ,60 . 3 57 9;8 9 7 2,99 2 84 , 6,7 5 4 , 0,61 605 1,4 1,0 1,7 1,6 NaCI-KC KCI , -3,58 , 9,3 , 3,01 , '6,6 -2,42 0,00 -2420 00 0 1,0 1 0 0,6 1 0 CsCI 3,76 3;87 10,8 11 6 3,12 . 3 18 7,1 7 4 1,51 27 2 , 1059 , 2,8 , 3,6 Th44 U3+ Ti CI N 2,90 , 6,0 , 2,83 , 6,0 , 0,40 1815 202 2,8 4 0 4,2 3 9 aCI LiCI-KCI 2,98 3;00 5,9 6 0 2,99 2 84 6,7 5 4 -1,21 1 6 1815 , 3,4 , 2,8 NaCI-KCI K 3,09 , 6,4 , 3,01 , 66 , 1 0,00 -1613 0 00 1,2 1 0 1,4 1 0 C1 CsCI 3,17 3,25 6,7 6 9 3,42 3 18 7,1 7 4 -0,40 0 6 , 605 , 1,5 , 1,4 , , , - , 1 202 0,4 11,4 for cadmium, lead, and indium electrodes, 102 103 for bismuth and zinc electrodes, and 104 105 for gallium and alumlinum electrodes. Considering the results of separation of rare-earth metals with liquid metal electrodes [1, 51, one should expect that the separation factors of uranium and cerium `(neodymium, praseodymium) will be -similar to, of uranium and dysprosium (erbium, yttrium) by 1 order of magnitude greater than, and of uranium and samarium (ytterbium) by 2-3 orders of magnitude greater than, the separation factors of uranium and lan- thanum. The separation factor of uranium and plutonium varies within 2 orders of magnitude depending on the nature of the liquid metal solvents employed, increasing in the order Cd, Zn, Bi, and Al and reaching 102 103 for aluminum electrodes. Zinc electrodes are efficient separators for uranium and zirconium. With zinc, indium, and lead elec- trodes one should expect enrichment of the metal phase with thorium, and of the KCI-NaCI melt with uranium. No significant separation of these elements should be expected with tin and bismuth electrodes. Preferential accumulation of uranium should be observed on gallium, antimony, and aluminum electrodes. Selection -of the liquid metal solvent makes it possible to change the separation factor of uranium and thorium within more than 2 -orders of magnitude. Less marked is the effect of the solvent salt on the selectivity of electrochemical processes in the liquid metal-salt system (see Table 2). The effect becomes appreciable only when the ions of elements being .separated have markedly different z/r parameters. For example, in the solvent series from lithium chloride to cesium chloride, the separation factors of uranium and zirconium decrease, and of thorium and uranium in- crease by more than 1 order of magnitude as a result of more intense complexing in the salt melt and stronger complexing bonds of Zr4+ (z/r = 4.88) and Th4+ (4.21) in comparison with U3+ (2.88) ions. Higher temperatures, as a rule, reduce the selectivity, particularly with bismuth (Pu-U) and zinc (U-Zn) electrodes. It is interesting to compare the electrochemical separation factors of elements in metal-salt systems with those obtainable in well-known separation processes (e.g., by extraction). The separation factors of ions of the above-considered elements by extraction of 0.3 M by tertiary amine solutions out of 2 M H.NOare as fol- lows [6) : 4 for U4+/Th4+, 2 for Th4+/U(V1), 103 for Pu4+/U(VIJ,, 102 for Pu4+/U4+, 4.104 for U4+/Sni3 ? 57. 103 for U(Vi)/Sm3+, 4.104 for U4+/Zr4+, 3 ? 103 for U(Vl)/`Zr4+, It IS seen that the selectivity of electrochemical pro- cesses in -liquid metal-salt systems is similar to that of extraction by organic solvents. The above results indicate the possibility in principle of separation of uranium from fission products,, plutonium, and thorium by electrochemical methods, and the considerable effect of the nature of metal solvent salts and temperature on the process selectivity. LITERATURE CITED 1. V. A. Lebedev et al., Zh. Fiz. Khim?, 46, No. 9, 2356 (1972). 2. M. V. Smirnov, Electrode Potentials in Molten Chlorides [in Russian], Nauka, Moscow (1973). 3. V. I. Silin and 0. V. Skiba, Preprint NIIAR P-118, Dimitrovgrad (1971). 4. V. A. Lebedev, At. Energ., 41, No. 1, 33 (1976). Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 5. A. V. Kovalevskii, V. A. Lebedev, and L F. Nichkov, Tsvetnye Met., No. 11, 45 (1973). 6. Oke.W. Hultgren, in: Reprocessing of Power Reactor Fuels [in Russian], Atomizdat, Moscow (1972), p. 103. DETERMINATION OF NEUTRON AND RADIATION COMPONENTS OF ENERGY RELEASE IN BORON-CONTAINING RODS USING GRAY CHAMBERS V. P. Polionov, Yu. G. Pashkin, and Yu. A. Prokhorov UDC 621.039.562.24 Well-known methods based on reaction speed and calorimetric measurements were used to determine the energy release in boron-containing -fuel rods in the PF-4F8 critical assembly [1-3] in which the ionization method using a Gray chamber was also tested. Unlike the methods used in [1, 2], the ionization method makes it possible to find the neutron and radiation components of energy release which is important for design pur- poses. The method is quite simple and. provides reliable results with an acceptable accuracy. Its sensitivity is approximately 10 times the sensitivity of the calorimetric method [2] and amounts to ^-10-7 W/g for natural boron. . The determination of energy release in boron-containing material is based on the well-known Bragg-Gray principle [4] which establishes the relation between the measured ionization of a gas confined in a cavity of a solid body and the release of energy in the walls of the body. The energy release per unit volume of the medium is given by ,z sot AU (C6+ Cini) K1, (1) Q=Wfsagas tVceN where W is the average energy of ion pair production in gas; f, ratio of stopping powers of the chamber mater- ial and the gas per electron; nsol and ngas, number of electrons per unit volume of the solid body and the gas, respectively; AU, change of chamber potential during irradiation time t with a power N; Cc and Cmi,,capacity of the chamber and the measuring instrument, respectively; Vc,. chamber volume.; e, electron charge; 77, ion collection efficiency in the gas gap of the chamber; and K, an extrapolation factor described below. The term DU(Cmi +Cc)/t is henceforth called the chamber current. Measurements were conducted with plane-parallel chambers. The gas gap was formed between the faces of two boron carbide cylinders 20 mm in diameter which are a part of the boron rod. The chamber is fitted with a device for varying and monitoring the gas_ gap. If a chamber with a gas gap S is placed in a reactor radiation field, the current 16 flowing in it is given by I6=Ia+IL1+IB,C+Ie+I1k, (2) where I?, ILi, and IB,C are neutron components due, respectively, to the products of (n, a) reactions in 10B and to recoil nuclei of boron and carbon produced by their interaction with neutrons, the latter component being negligible in comparison with the first one; Is is the radiation current component measured with a special boron carbide chamber in which the gas cavity is shielded by an aluminum foil from the (n, a) reaction products, and Ilk is aleakage current determined in the absence of reactor radiation and found to amount to only 3% in our conditions. For a "zero" gas gap, for which Eq. (1) is true, the chamber current components differ from currents flowing in the chamber with a finite gas gap for the following reasons : the Bragg-Gray condition [4] that the fraction of particles entering the gas gap with a residual range less than the gap length is negligible is not com- pletely satisfied; the neutron flux on the inner surfaces of chamber electrodes increases in the presence of a gas gap; an edge effect takes place in which a fraction of charged particles, mainly from the edges of elec- trodes, emerges at an angle to the chamber axis and leaves the sensitive volume of the chamber without fully Translated from Atomnaya Energiya, Vol. 47, No. 3, p. 182, September, 1979. Original article submitted April 17, 1978. 0038-631X/79/4703-0733$07.50 ?1980 Plenum Publishing Corporation . 733 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 TABLE 1. Energy Release in Boron Rod Measured by Different Methods Energy release, 104 W/g neutron component total energy release Ratio of reaction speeds [1] Calorimetric [2] Ionization 1.43.+0.16 1.43:1: 0.1 spending their energy. These effects diminish with decreasing gas gap and vanish completely in a "zero" gap. In this case, the chamber current satisfies all conditions of the Bragg-Gray equation [4]. To find the chamber current at "zero" gap, we have subtracted from the currents obtained with 0.07 and 0.5 mm gas, gaps the radiation current component and the leakage current at these gaps. Then, we have deter- mined the parameter x in expression (3) which describes the current-in a plane-parallel chamber due to a par- ticles and lithium nuclei taking into account the gas gap dimensions expressed in a particle track lengths: I~{ 1L1 2 a./J+11LiJ 2 11+1n IT (3) where k is the ratio of track lengths of a particles and lithium nuclei. This expression has been obtained for a small gap assuming an isotropic angular distribution of reaction products : helium and lithium nuclei. Next we have determined the neutron component of the chamber current for a "zero" gap which is 1.13 times the c{irrent for a 0.07-mm gap. The obtained data are listed in Table 1. The results confirm that the measurements are correct and indicate the possibility of using the ionization method for measuring the energy release components in boron rods in low-power critical assemblies with an accuracy acceptable for practical purposes. The method is quite simple and easy to implement. The reactor power was measured by the frequency method by S. A. Morozov to whom the authors express their gratitude. LITERATURE CITED 1. V. A. Kuznetsov et al., At. Energ., No. 5, 926 (1972). 2. A.S. Zhilkin et al., At. Energ., 42, No. 6, 502 (1977). 3. A. L Mogil'ner et al., At. Energ., 24, No. 1, 42 (1968). 4. J. Hein and H. Brownell (editors), Radiation Dosimetry [Russian translation], IL, Moscow (1958). 5. A. L Mogil'ner et al., Preprint FL-98, Obninsk (1967). Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 PHOTOPRODUCTION OF NEUTRONS IN A THICK LEAD TARGET V. I. Noga, Yu. N. Ranyuk, UDC 621.384.64.038.624:539.125.5.03 and Yu. N. Telegin The purpose of the paper is a study of the yield of neutrons from a thick lead target bombarded by 230- and 1200-MeV electrons in linear accelerators of the Physicotechnical Instutute of the Academy of Sciences of the Ukrainian SSR. The experiment was staged as follows. A beam of electrons hit a lead target in the form of a 0.2- to 8-cm-thick cylinder with a diameter. of 2.5 cm. Neutrons were counted by the method of radioactive indicators [1]. Neutron detectors (aluminum samples) in the form of 0.5-cm-thick disks 3 cm in diameter were placed at a distance of 15 cm around the lead target at definite angles with respect to the electron beam direc- tion. The activity induced by the 27A1(n, p)27Mg reaction was measured with a y spectrometer consisting of a Ge(Li) detector and a "Langur" spectrometer. The spectrometer was connected to an M-6000 digital computer which recorded and processed the spectra. The various reactions of neutron-nuclei interaction used for neutron detection by the method of induced activity result in the production of radioactive nuclides. Most suitable for this purpose under conditions of y background are (n, p) reactions for which the background process is the photoproduction (y, 7r+). Since the photoproduction process is about 140 MeV, y quanta do not contribute to the measured activity below this energy, while for Ey>140 MeV their contribution is negligible because of the small photoproduction cross sections of it mesons as compared with the (n, p) reaction cross section. However, the use of (n, p) reactions leads to practical difficulties associated with the need of a monoisotopic detectors. The 27A1(n, p)2 7A% reaction used by us is one of the most suitable reactions for practical applications. The angular distributions of neutron yield obtained for different target thickness and electron energies are nearly isotropic indicating that evaporation is the main neutron production mechanism. The integral neu- tron flux was calculated from w- 3 ? 6 00 000 0 0 1 2 3 4 5 6 7 L,y, CM Fig. 1. Neutron yield fn as a func- tion of target thickness lpb at 230 (0) and 1200 MeV (O). Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 183-184, September, 1979. Original article sub mitted June 26, 1978. 0038-631X/79/4703- 0735$07.50 ?1980 Plenum Publishing Corporation 735 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 0 500 1000 E0, MeV Fig. 2. Neutron yield fn as a func- tion of electron energy Eo under saturation. conditions : results of [3, 4] (0), our results (0); calculated in [5], calculated in [6]. where k is a factor accounting for neutrons with an energy below the reaction threshold, and R is an activation integral (cm2/sr ? electron) [1]. The neutron production cross section was described by the single-step function j0 for En < .Qef ; an(En) = `aef for En ~> Qef , where Qef and. aef are, respectively, the effective threshold and the effective threshold cross section. In our case Qef=4 MeV and aef=30 mb. To find k the neutron spectrum was represented by the expression s(p (En) -En exp En/T), where En is the neutron kinetic energy and T is a constant. The constant T was calculated from experimental data on the spectrum of neutrons from a tantalum target obtained for a maximum bremstrahlung beam energy of 140 MeV [3]. The estimated error of fn is f 30%. Absorption of neutrons in the lead target was neglected. Data shown in Fig. 1 as well as similar results of other works [4] indicate that the behavior of fn(lpb)is the same for different electron energies E0 and is characterized by the fact that neutron yield saturation begins at a certain target thickness. The minimum target thickness 1min corresponding to saturation depends on E0. The thickness lmin shifts toward greater thickness with increasing E0. This must be taken into-account in selecting optimum target dimensions for a given initial energy. Of special interest is the dependence of neutron yield on electron energy since there is a practical possi- bility to vary this yield in particular when the accelerators employed have a high upper energy limit. Figure. 2 shows the available calculated and experimental data concerning the dependence of the yield of neutrons from a lead target on E0. To within an acceptable error the results obtained in [5] agree with experimental results. The results calculated -in [6] are considerably lower than other data. It can be assumed that the discrepancy is a result of the fact that the processes leading to neutron production were not fully accounted for. This is parti- cularly important at energies exceeding the pion photoproduction threshold. LITERATURE CITED 1. E. A. Kramer-Ageev, V. S. Troshin, and E. G. Tikhonov, Activation Methods of Neutron Spectrometry [in Russian], Atomizdat, Moscow (1976). 2. C. Burgart et al., Nucl. Sci. Eng., 42, 421 (1970). 3. R. Alsmiller and M. Moran, Nucl. Instrum. Methods, 48, 109 (1967), 4. W. Barber and W. George, Phys. Rev., 116, 1551 (1959). 5. W. Swanson, SLAC-PUB-2042 (1977). 6. J. Levinger, Nucleonics, 6 No. 5 (1950). Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 MATHEMATICAL MODEL FOR CALCULATING FISSION P~ROD,-UCTS CONCENTRATION. AND. ENERGY RELEASE IN CIRCULATING NUCLEAR FUEL L. I. Medvedovskii, E. S. Stariz.nyi, V. A. Ch.e.r-:kashin, V. A. Rudoi, and K. I. Ste.panova UDC 621.039.55 The solution of certain problems associated with the design and application of uranium radioactive loops requires calculation of fission products concentration and of the three-dimensional distribution of -y- and /3- radiation energy release and its spectral composition in circulating nuclear fuel. Methods and results for cal- culating these characteristics in nuclear reactors with noncirculating fuel have been published in severalworks [1-3]. Design works, as a rule, do not take into account isomeric transitions, direct generation of nuclides in chains, and burnup of active nuclides. We have designed a mathematical model of the accumulation of fission products in circulating nuclear fuel. and devised a method for calculating the radiation characteristics ofura- nium radioactive loops (distribution of the power of .y and. (3 radiation and their spectral composition in the uranium radioactive channel). This mathematical model can also be employed for calculating the radiation characteristics of fission products in pulsed reactors, in which burnup of fission products is especially high, and in nuclear.reactors with fixed fuel. Consider all possible transmutations of fission product nuclei. A nucleus (including isomers) can be gen- erated directly in fission, as a result of neutron capture by an isotope of lower mass, as a result of decay of its Fig. 1. Initial (a) and linearized (b).chains (O,.isomer; .40, ground state) : a: 1) (1-asi-4j-(31-1j) Al-ijAi 1j; 2) a +iAi-ij+iAi_ +i. 3) .1-.j42~Ai- +2;4) o11j+2 Ai-1j+2; 5) (1-aij+i)Xij+iAij+i; 6) aij+1. Xij+1Aij+1; RijXijAii; 8) y1j-iZf!; b: 1) X3i-1j-2Ai-ij-2; 2) X21-ij-i Ai-j-1;-3) Xii-jAi-j; 4) vi-ij+ltAi-1j+i; 5) Yij-iEf I. Translated from Atomnaya Ener.giya, Vol. 47, No. 3, pp. 184-186, September, 1979. Original article sub- mitted July 17, 1978;. revision submitted January 25, 1979. 0038.-531X/'79/4703-,0737$07.50 ?.1980 Plenum Publishing Corporation 737 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 10-5 101 102 103 104 105 Time after instataneous fission Fig. 2. Comparison of results obtained by the proposed method with experimental data: CQ data- [21, 0) data. [41, V) data [5], ?) data [6], --y our results. isomer (if the nucleus is not an isomer), and as a result of decay of an isobar and of an isomer of isobar with smaller nuclear charges. The generated nucleus can capture a neutron, pass into a ground state (if the nucleus is an isomer), into an isobar next in the chain, and into an isomer of the isobar next in the chain. All these pro- cesses are reflected in the following system of differential equations : dAsj/dt = yjjZJm-~1JAjj-a, /DAti-F at,-i~,J-jAzJ-i+ I +aii-1, , Aii-i+piiXiAti+at-lJAt-1i; dA;jfdt=Uii EJcD-XiiAii-aiiDAii+(1-al-i- ~ii-i)~'ii-iAi5_1 +(1-atJ-1) XjJ-1At,-1+ ai-1J@Ai-Sir (1) where Ai is the concentration of the j-th nucleus in the i-th chain; yij, probability of output in fission; Ef, nuclear fuel fission macro cross section; 4), thermal neutron flux density; 7Lij, decay constant; of , neutron cap- ture micro cross section; Cr and (3, probability coefficients :(Fig. la); quantities marked with strotes corre- spond to isomers. To simplify the solution let us arrange isomers with ground states in one chain numerating them so that a nucleus with a higher number cannot turn into a nucleus with a lower number; transitions take place between nuclei with numbers differing by not more:than three. The concentration Aij of isotope (i, j) in the mixture can then be -described by a single equation (see Fig. 1b) : dAij/dt=YtJ-AtjAtj+Pi-1J+A1-1J+?st1-sA1j-s +'siJ-2AiJ-z+21jj-1At,-1; (2) Au (0) = B1J+ y iJ=y11X1D; Ptj=atjD; AtJ=X1J+PiJ; 3 4iJ-k=XtJ-kati-k J; 7, aiJJ+k=1. Here aijj+k is the probability of transition of nucleus (I, j) into the nucleus (i, j + k). Since nuclear fuel in a uranium radioactive loop enters the neutron field periodically, for each elementary volume we have ( t ) f 0t is averaged for the whole population and takes account of all i possible effects leading to the death of an individual. If, for each effect individually, the sensitivity is defined by the quantity x,i, then the total sensitivity to a given agent can be found from the relation x =Y xt . (12) tIn Soviet literature, there is no generally accepted equivalent for the conversion of the English term "dose commitment." The term "expected dose," which is used sometimes, clearly is ambiguous. Therefore, by analogy with the mathematical expectation, it is proposed to denote "dose commitment" as dose expectation. In this case, the probable nature of this quantity is emphasized, and also its difference from the actual absorbed dose. Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 The relations considered are valid in the case of linearity of the risk-exposure relation. However, even in the case of nonlinear relations, different sections of the risk-exposure curves can be approximated by straight lines and their relations can be used. The proposed methodology also allows the introduction of uniformity into the quantitative description of the phenomena of synergism and the mutual suppression of certain agents simultaneously affecting the popula- tion. For example, if on the average there are first and second agents, the combined effect of which is non- additive, the..following expression can be written for the effective dimensionless exposure E: E_co (Xi, X2) [X1/x1+k2/x2]? (13) The numerical value of the coefficient w is greater than unity in the case of synergism and less than unity in the case of mutual depression. Obviously, the quantity E should serve as the basic normalization and can be used for comparing all types of practical worker, where there exists a risk of death of the individual. In this case, data about vc are es- sential for the most different agents. They can be obtained partially on the basis of analysis of the published data; however, special purposeful investigations are required to a considerable degree for this, in the first place experiments on animals, and also natural hygienic investigations, including a study of the environment and the health of staff and-population. It can be verified that the most complete information necessary for quantitative estimates and forecasts, are in the field of radiation hygiene [7]. The dose-effect relation for chemical carcinogens has been studied very inadequately [6]. Data which might be used for quantitative estimates of the effect in the case of combined action are still few. Numerous gaps in the available data indicate the direction of the immediate investigations which will be necessary for a correct .estimate of the effect of the environment on man in the actual conditions of the combined effect of many agents. LITERATURE CITED 1. Radiation Protection [Russian translation], Atomizdat, Moscow (Publication 26 ICRP) (1978). 2. Standards of Radiation Safety [in Russian], Atomizdat, Moscow (NRB-76) (1978). 3. V. Lyscov, "Comparative evaluation of risks from physical and chemical mutagens and carcinogens in the environment, ".in: Seventh International Biophysics Congress, September 3-9, 1978, Kyoto. 4. NKDAR Report United Nations 1977 [in Russian], New York (1978). 5, Science, 187, 503 (1975). 6. V. A. Knizhnikov, Gig. Sanit., No. 3, 96 (1975). 7. E. I. Vorob'ev et al., At. Energ., 43, No. 5, 374 (1977). Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 ESTIMATE O_F DOPPLER BROADENING OF RESONANCES V. V. Kolesov and A..A. Luk'yanov UDC 539.5.173.162.3 . The effect of the thermal motion of the nuclei of a medium on the form of the energy dependence of neutron cross sections in the resonance region must be taken into account in the analysis of neutron spec- troscopy data and in estimates of nuclear temperature effects in reactors. The problem consists in the transformation of the cross section a(E t), determined theoretically in the center of mass system as a function of the energy of the relative motion of the neutron and the nucleus E 1, to the laboratory system where the neutron energy is E ; a(E)= 1 a(E')F(E-E')dE'. (1) The distribution function F(E - E 1) characterizes the statistical spread of the energy E I resulting from the thermal motion of the nuclei of the medium. Usually the gas model approximation is used, where F (E-E') dE' _ (1/ 1/rzA) exp[-(E-E')2/A2] dE (2) Here A= 21M/ (A +1) is the so-called Doppler width and kr is the average energy of thermal motion of the atoms [1, 210 The energy structure of cross sections at resonances is determined by the superposition of the known functions [1]: exp [-(z-y)2t21 dy; 1/n 1-f-y' X(x, ~)= C exp-(x-y)2t21" ydy, 1+y2 where x = (E - E 7,)2/ rX, t r7,/24, and r is the resonance width. These functions have been well studied, detailed tables of them exist, and descriptions of algorithms and numerical calculation programs are available [1-4]. The functions It and X are widely used in the analysis of neutron cross sections in the region of resolved levels, and also in the study of resonance effects in nuclear reactors [2]. However, the necessity of turning to numerical calculations even for qualitative estimates of the Doppler broadening of resonances frequently 'leads to considerable complications. Thus, in existing programs for seeking resonance parameters from ex- perimental data, up to 90% of the machine time is consumed in calculating the functions ' and X. The integral representation of these functions makes the construction of the solutions of the transport equation for resonance neutrons difficult even In the simplest problems. For rough estimates of effects related to the Doppler broadening of resonances it is convenient to have rational approximations of the functions i and X obtained by using a distribution function of the Lorentz form [5] in (1): F (E-E') dE'=(A/2-1) dE'l [(E-E')2+42/41, (4) where 0 is the characteristic width of the distribution at half-height. By averaging (1) we obtain approximate expressions for the Doppler functions: '(x, F)=(1+6)/[x2+(1 H--b)21; j (X, t)=x/1x2+(1+6)21, where 6 =0/ r [5]. Relations between A- and A are established by comparing specific integral combinations of the Doppler functions. Thus, from the equality of the integrals of the squares of these functions it follows that Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 205-206, September, 1979. Original article submitted November 20, 1978. 770 0038-531X/ 79/4703- 0770$07.50 ?1980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 Fig. 1. Exact ( ) and approximate (---) functions for calculating Doppler broadening of.a resonance for various values of t; a) the functions T and'; b) the functions X and X. The results of a numerical solution of this transcendental equation can be represented by the approximate relation 6-1=t [2.5+2t+~(0.t+t) (0.12+b)]. Figure 1 shows that the general qualitative agreement of the exact and approximate functions improves with increasing Various integral characteristics of cross sections in the resonance region are of practical interest. These include transmissions averaged over the resonances as.,a function of sample, thickness (exp[-nv]) , average cross sections measured with filtered beams (aaexp[-na]), and effective resonance integrals. Thus, the temperature dependence of the effective integral of an isolated resonance is characterized by the self-shielding factor x1 r 'dx n J 1?a(Tcos2cp-xsin2p)' where q is the phase of the potential scattering, a = u?/vp is the ratio of the cross section at the resonance maximum to the potential cross section of the medium per nucleus of the resonance absorber [2]. When using approximation (5) this integral is calculated as K;t~1l/(1+1+Scosz(p) (1-1+Ssin?cp ). A comparison of the results of numerical calculations of the integrals (8) given in [4] with our results (9) yields the approximate relation: 6-1=~(2.5-j- 2C+ [(1+acos 2(p) C](0.i+010.12+6)), which when used in E q. (9) reproduces the values of the integrals with an error of less than - 3% over the whole range of the parameters. The result of averaging a resonance cross section with the distribution function (4) is equivalent to the ordinary Breit-Wigner formula, where taking account of Doppler broadening appears only in the redefinition of the total width (F-=r+2~). This enables. one to obtain simple analytic expressions for the estimation and parametrization of the temperature dependences of various integral characteristics of cross sections used in- reactor physics applications. The fundamental criterion of the accuracy of the approximation must be a com- parison with the data of integral experiments, since calculations with the integral Doppler functions are gener- ally approximate per se. LITERATURE CITED 1. H. A. Bethe, Rev. Mod. Phys., 9', 69 (1937). 2. A. A. Luk'yanov, Slowing Down and Absorption of Resonance Neutrons [in Russian], Atomizdat, Moscow (1974). - Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 3. V. N. Faddeeva and N. M. Terent'ev, Tables of Values of the Probability Integral of Complex Argument (in Russian], Gostekhteorizdat, Moscow (1954). 4. L. P. A bagyan et al., Bulletin of the Nuclear Data Information Center. Propagation of Resonance Neu- trons in Homogeneous Media. Theory and Special Functions [in Russian], Atomizdat, Moscow (1968). 5. A..A. Lukfyanov, Structure of Neutron Cross Sections [in Russian], Atomizdat, Moscow (1978). NEUTRON RESONANCES OF 247Cm IN THE ENERGY RANGE T. S. Belanova, A. G. Kolesov, A. V. Klinov, S. N. Nikol'skii, V. A. Poruchikov, V. N. Nefedov, V. S. Artamonov, R. N. Ivanov, and S. M. Kalebin UDC 621.039.556 The neutron resonance parameters of 247Cm were calculated on the SM -2 reactor from the transmission of a sample of curium, which was measured by the time-of-flight method. The neutron pulse was shaped by a mechanical selector with three rotors, suspended in a magnetic field [1]. The best resolution on the flight base of 91.7 m amounted to 120 nsec/m. The sample for investigation was made from powder, calcined at a temperature of 900-1100?C, of the stable oxide of curium (Cm20a with a known oxygen content. Included in the impurities were 243Am and 240Pu; the latter is built-up in the sample as a result of the decay of 244Cm. The maximum 247Cm content at the time of measurement amounted to 0.64.10-4 atom/b. The content of inert impurities, with the exception of oxygen, did not exceed 3%. The transmission was measured in the neutron energy range of 0.5-20 MeV with a statisti- cal error on the resonance limbs of 1-2%. The neutron background did not exceed 2% of the effect. The neutron resonance parameters were calculated by the shape method according to the Bright-Wigner single-level formula [1]. As the neutron resonance parameters of 244Cm, 245Cm, 246Cm, 248Cm, 243Am, and 240Pu are well known [2-6], the 247Cm resonances could be identified in the measured transmission and their parameters were calculated (see Table 1). In [3, 5, 6], the 247Cm resonances with energies of 1.247, 3.19, and 18,1 eV were erroneously ascribed to 2451Cm. The neutron resonance with an energy of 2.919 eV was not previously detected. Only 5 neutron resonances of 7Cm were identified with large values of 2grn, because the 247Cm content in the sample was low (1.7 mg) and the resonances of this isotope were identified on the background of the large number of resonances of 244Cm, 24sCm, 241Cm, 24$Cm, 24Am, and 240Pu located in the energy region being investigated. Eo, eV r. MeV 2a rn, MeV 1,247+0,05 74?4 0,56?0,09 2,919+0,010 70+30 0,10?0,04 3,189T0,010 103?6 1, 0?0,1 9,55?0,03 166+60 0,91?0,33 18,1?0,1 210?170 3,7?1,5 LITERATURE CITED 1. T. S. Belanova et al., Preprint, Scientific-Research Institute of Nuclear Radiation P-6 (272) [inRussian Dimitrovgrad (1976). 2. Neutron Cross Sections, BNL-325, Third Edition, Vol. 1 (1973). 3. T. S. Belanova et al., At. Energ., 42, No. 1, 52 (1977). Translated from Atomnaya Energiya, Vol, 47, No. 3, pp. 206-207, September, 1979. Original article submitted December 12, 1978. 772 0038-531X/79/4703-0772$07,50 ?1980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 4. - R. W. Benjamin et al., Nucl. Sci. Eng., 55, No. 4, 440 (1974). 5. T. S. Belanova et al., Preprint, Scientific-Research Institute of Nuclear Radiation P-13 (307) [in Russian], Dimitrovgrad (1977). 6. T. S. Belanova et al., Proceedings of the Conference on "Neutron Physics," TsNltatominform, Moscow, Pt. 3, 224 (1976). 773 Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 CONFERENCES, MEETINGS, SEMINARS SOVIET-BRITISH SEMINAR ON FAST REACTORS- R. P. Baklushin At the Seminar on nExperience in the Design, Experimental Development, and Operation of the Basic Plant for Fast Sodium Reactors," the British specialists presented 17 reports on the construction of units and plant for the newly designed commercial CFR reactor and on the various material-behavior problems. The Soviet specialists participating in the Seminar, visited nuclear centers at Risley, Harwell, and Dounraey, and also the PFR, operating at a power of 20 MW (thermal) and with a nominal capacity of 600 MW. The nominal thermal capacity was achieved in February 1977, but the electrical output was below the design output as a consequence of one of the intermediate steam superheaters and certain.of the regenerative water preheaters (mixing type) being switched off, and also because the vacuum in the condenser was below the calculated value. The reactor, in addition to the generation of electric power, is used for testing fuel elements, plant, and com- ponents of the CFR reactor. In the last 2 years, the PFR frequently has operated at only 66% of its power be- cause of these investigations. There were no cases of failure of the regular fuel elements and the burnup was 5%. There were 6 experimental cassettes with fuel elements in the core, and after 6 months there were two' cases of fuel element failure among them. In January 1979, after this failure, the power was reduced to 66%. The burnup of the experimental fuel elements then in the reactor, amounted to 9%, but in general it reached 22%. In the week before the arrival of the delegation, a leak appeared in the evaporator of the second steam generator, which was the reason for further reduction of power. Although the leaks in the steam-superheaters of stainless steel attracted the most attention of the specialists, and after which cracking occurred in conse- quence of alkaline corrosion, these leaks amounted in all to two or three out of 15. Others were observed in the evaporators at the site of the tube weld with the tube plate because of corrosion pitting in the zone of the welded seam from the water side. i Great attention was paid to the reprocessing of PFR fuel (plutonium dioxide). The British specialists consider that they have solved this problem. In the summer of 1979 at Dounraey, it is proposed to start up a facility which has been designed for the reprocessing of all the fuel unloaded from the PFR. On this same area, it is planned in the future to manufacture fuel element assemblies from the reprocessed fuel, and thus to close the fuel cycle. The decision to construct a nuclear power station with a C FR has not been taken and the area has not been assigned. The reasons for this were named as the reserves of petroleum discovered in the North Sea and the opposition of the protectors of the environment. It is expected that the construction of the nuclear power station will be started in 1984. The following problems were discussed in more detail. In the core of the CFR, three types of control and safety rods are distributed: 19 control rods (these are the burnup compensators) and 9 scram rods -main and auxiliary. The design of the control and scram rods is conventional. Their actuators are located above the core on rotatable plugs. The main interest is the auxiliary group of scram rods. It is not connected mechanically with the rotatable plugs, but can provide pro- tection of the reactor during fuel recharging. The rods are retained above the core by the action of a stream of sodium, fed into the guiding sleeve from below by special electromagnetic pumps. The rods are divided into three groups, with three in each group, and fed with an individual pump. When the pump is switched off, the rods fall downwards under the action of their own mass with a velocity of 0.4 m/sec. On falling into the core, the three rods inject a reactivity of -1.12% Ak/k. The geometry of the sleeve and the rod is such that in the. case of erroneous switch-on of the pump, the latter remains in the lower position. In the upper portion, in which it is retained by hydraulic forces, it is returned by a special pickup mechanism. After switching on the pumps, the pickups are disengaged from the rods and raised upwards. For the scram system, a special electromagnetic pump was developed and tested; it has a winding of copper strands surrounded by a magnesite insulation and a winding of stainless steel. The pump can be oper- ated when immersed in sodium at 600?C. It was designed so that its central part with the electrical winding could be withdrawn from its channel, in the event of the occurrence of a failure. The diameter of the central part of the pump of different stub-size varies from 32 to 300 mm, and the flow-rate correspondingly from 0.4 to 50 liter/sec, 7 714 Translated from Atomnaya Energiya, Vol. 47, No. 3, pp. 208-213, September, 1979. 0038-531X/79/4703-0774$07.50?1980 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2  Declassified and Approved For Release 2013/02/15: CIA-RDP10-02196R000800020003-2 The fuel recharging system of the CFR has significant differences from the PFR fuel recharging system. Fuel element assemblies are withdrawn from the core by three mechanisms of the "direct" type (without panto- graphs). They are lined-up with the fuel element assemblies by three rotatable plugs. The change of design