Soviet Atomic Energy Vol. 32, No. 3

Body: Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Russian Original Vol. 32, No. 3, March, 1972 \ arokiMhyil Translation published-November ' . - SOVIET ? ATOMIC ENERGY ATOMHAII 3HEP114F1 (ATOMNAYA ENERGIYA) TRANSLATED FROM RUSSIAN CONSULTANTS BUREAU, NEW YORK Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 , SOVIET ATOMIC ENERGY \ ts ' 1 ' . - Soviet Atomic Et.ergy is a cove'r-to-cover translation of Atomnaya , Energiya, a publication of the Academy of Sciences of the USSR. arrahgement with Mezhdunarodnaya Knigi, the Soviet book export agency, makes available both advance copie's of the Rus- sian 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 im- prove the quality of the latter. The translation began with the first ? issue of the Russian journal. - Editorial Board of Atorrmaya Energiya: ? Editor: M: D. Millionshchikov' Deputy Director I. V. Kurchatov Institute of Atomic Energy Academy of Sciences of the USSR Moscow. USSR Associate Editors: N. A. Kolokol'tsov N. A. Vlasov 4: A. Bochvar V. V. Matveev, N. A. Dollezhar M. G. Meshcheryakov , V. S. Fursov P. N. Palei I. N:Golovin ' V. B. Shevchenko V. F. Kalinin D. L. Simonenko 'A. K. Krasin V. I. Smirnov A. I. Leipunskii A. P. Vinogradov ? A. P. Zefirov Copyright 1972 Consultants Bureau, New York, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. No article contained herein may be reproduced for any purpose whatsoever without permission of the publishers, Consultants Bureau journals appear about six months after the publication of the -original Russian issue. For bibliographic accuracy, the English issue published by Consultants Bureau carries the same number and date as the original Russian from which it was,transiated. For example, a Russian issue published lr Decem- ber will appear in a Consultants Bureau English translation about the following June, but the.transiation issue will-carry the December date. When ordering any volume or particular issue of a Consultants Bureau Journal, please specify the date and, where applicable, the volume and issue numbers of the original Russian. The material you will receive will be a translation of that Russian volume or issue. Subscription $75.00 per volume (6 Issues) 2 volumes per year (Add $5 for orders outside the United States and Canada.) Single.Issue: $30 'Single Article: $15 CONSULTANTS BUREAU, NEW YORK AND LONDON Davis House 227 West 17th Street 8 Scrubs Lane New York, New York 10011 Harlesden, NW10 6SE England Published monthly. Second-class postage paid at Jamaica, New York 11431. Declassified in Part - Sanitized dopy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 50X1 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 R Next 1 Page(s) In Document Denied Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 THE ROLE OF NUCLEAR POWER IN LONG-TERM FUEL ? ENERGY BALANCE IN THE USSR A. A. Makarov, A. S. Makarova, A. G. Vigdorchik, A. N. Zeiliger, G. B. Levental', and A. M. Belostotskii UDC 621.039.003 The forecasted long-term program of extensive growth of nuclear power calls for comprehensive technological and economical research on the optimum scale and ways of this growth. These studies cannot be based (as they are now) on isolated comparison of the economical performance of nuclear and conven- tional power plants. In fact, in the coming decades nuclear energy will undoubtedly affect the output of traditional energy resources (petroleum, natural gas, and coal) displacing the more expensive fuel sources and thus changing its own economical importance. Moreover, the growth of nuclear power will inevitably have a great im- pact on the structure of the electrical power systems. Thus, rational scales and ways of the growth of nuclear power cannot be determined without a study of the optimal structure of the long-term fuel?energy balance and of a unified electrical power system of the country. One of the fundamental problems which must be solved in the study of the role of nuclear power is the length of the design period On the one hand, this period should not be limited to the initial stage during which nuclear power techniques are mastered and put into practice and thus should be longer than five to 10 years; on the other hand the design period must not be excessively long in order to prevent the error in the initial information from growing into proportions which make the results of the study unreliable and unrealistic (this is particularly true of the effect of some cost indicators on the optimum structure of fuel ?energy balance). Asa result, at the first stage of this study the design period was limited to the time pre- ceding the widespread use of fast reactors The choice of a sufficiently long design period turns the study of the prospects of nuclear power into the problem of predicting the gain it affords inthe fuel?energy balance. A special method has been de- veloped which is based on an analysis of the uncertainty zone in the optimal growth of the country's power resources [1]. The basis of the method is a mathematical model which simulates the power economy of the USSR. On the one hand, this model describes in sufficient detail the conditions of simultaneous develop- ment of fuel production (including transportation of coal, gas, and fuel oil from one region to the other), of the Unified Electric Power System, and of the main classes of consumers. The model considers in par- ticular detail the European Electric Power System block which describes the day-by-day and annual opera- ting conditions of the existing and new power plants (of various types) and of the intersystem electrical transmission lines. Nuclear power is represented in the model by nuclear steam and heat electrical plants under various operating conditions, and by constraints on the total installed power of nuclear reactors. The overall dimensionality of the mathematical model includes 250 constraints and 700 variables. On the other hand, the model makes it possible to obtain from the analysis or to generate inde- pendently a representative set of typical combinations of possible conditions of system development and to find in an acceptable time an optimal version of fuel?energy balance corresponding to each such combination. In the present study such versions were obtained for 100 different combinations of economical indicators of the basic objects and for 50 combinations of such decisive factors as the electrical power consumption Translated from Atomnaya Energiya, Vol. 32, No. 3, pp. 187-196, March, 1972. Original article submitted August 12, 1971. 0 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without permission of the publisher. A copy of this article is available from the publisher for $15.00. 215 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 100 98 30 28 /0 Coal outpu NPP 0 20 30 40 520 60 70 '00 90 80 70 2 50 a.) 40 e 30 !!-,?20 tj 'DO 60 Peek il': ibPi (nTPVndyroecc%farpas(ESP) Petroleum- f ired CPP 80 90 100 20 30 40 50 60 70 80 90 100 1?-.2) 20 '6 18 2 16 14 CenterNorth- West Ukraine Ural 12 - 10 -al 8 - ,- 6 04 5 2 Central Asia Western Siberia Eastern Siberia 20 30 40 50 60 70 80 Total capacity of nuclear power plants in percent of its maximum economically sound value 90100 Fig. 1. Effect of the variation of total output of nuclear power plant (NPP) on: a) the fuel -energy balance structure; b) the growth structure of the Unified European Power System (UEPS); c) terminal fuel cost. Level of gas production: ? ? ?) 1000 billion km3; ? ? ? ? ?) 960 billion km3; ) 880 billion km3. of the USSR, the potential volume of coal and petroleum production, the total capacity of nuclear electric plants and the operating characteristics of power plants. An analysis of the resulting combinations of optimal versions of fuel?energy balance makes it pos- sible to determine the energetic and economical factors that most significantly affect the optimal values of the growti parameters of nuclear power. In this study such parameters are: the total capacity of nuclear power plants, the proportion of nuclear steam and heat plants, the territorial distribution of the different types of plants, and the mode of utilization of nuclear steam plants. The optimal values of these parameters are plotted graphically as functions of the basic economical and energetic factors. These functions are then analyzed for the presence of clearly defined "discontinuities" that separate the domains of the acceptable and unacceptable values of the investigated factors. On this basis it is possible to formulate certain basic rules for the future growth of nuclear power which for the inspected energetic and economical factors (the cost of nuclear power plants, their regulation capabilities, distribution, etc.) are established as domains of their limiting values, and for all other factors as simplified but descriptive relations that reflect their effect on the investigated parameters of nuclear power. Application of this method of prediction to the analysis of the future growth of nuclear power in the USSR led to the following results. Effect of Nuclear Power on the Optimal Fuel?Energy Balance of the USSR. Multifactorial analysis of the optimum fuel?energy balance has been applied to a case in which the predicted level of consumption of electric power in the USSR will be four times as high and the output of all kind of energy resources twice as high as now; this includes an increase of natural gas production by a factor of 4-5 and of petroleum by a factor of 2-2.5. However, in spite of the rapid absolute growth of the production of high-quality fuels, the dynamical structure of the fuel?energy balance is expected by that time to undergo radical changes: the effect of the increasing contribution of petroleum and natural gas to fuel?energy balance, observed for the last 15 years and expected to continue in the coming decade, will be in the long-range much reduced or cease altogether. Taking into account the absolute growth of energy consumption this means the necessity of in- creasing the absolute capacity of coal and the use of nuclear energy. The results of optimization of fuel?energy balance indicate that the maximum economically sound fraction of nuclear energy in the total energy output amounts at the considered stage to 10-12% (Fig. la). Such a growth in nuclear power production could afford a saving of 6-8 billion rubles annually in expenses and 3-4 billion rubles in capital investments. The effect of nuclear power on the optimal growth of the main branches of the fuel industry is il- lustrated in Fig. la. The observed reduction in the use of organic fuels associated with the rise in the 216 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 capacity of nuclear power stations is due principally to a drop in coal production. The available petro- leum and natural gas resources are fully utilized at all the considered levels of their production. Moreover, increased use of high-quality fuels even leads to a certain drop in the optimal capacity of nuclear power plants* as a result of competition of natural gas from the Eastern regions of the European USSR and be- cause of region of variable load on the electric power consumption curve. A detailed analysis indicates that future development of the main coal basins is almost entirely governed by the scale of growth of nuclear power ? from total cessation of growth in case of high-capacity nuclear power stations to maximum growth in case of minimal use of nuclear power. The principal com- petitor of nuclear power in this case is coal from the Kansk-Achinsk and Kuznetsk Basins delivered to the European regions of the USSR by rail and partially in the form of direct current electricity transmitted by power lines (to distances of up to 2000-4000 km). Changes in fuel supply conditions as a result of increased use of nuclear power affect most signif- icantly the terminal consumers of fuel, i.e., the power plants of the European branch of the Unified European Power System. Figure lb shows the changes taking place in the power generation structure of the Unified European Power System (with respect to the kind of fuel employed) as a function of the total capacity of nuclear power plants for minimum and maximum natural gas production. It is seen that, firstly, as their capacity increases nuclear power plants displace more and more conventional coal-burning plants (pre- dominantly those using Donets Basin coal); with still higher capacity nuclear power plants begin to dis- place power plants using Kansk-Achinsk coal (in Ural and along the Volga) as well as the transmission of dc power to the Western regions of the USSR. Secondly, it it seen that over the entire range of variation of natural gas resources, most gas- and oil-fired power plants should be designed for peak load operation with possibly only a small fraction of them designed for load switching; most high-capacity base electric plants should be designed for solid fuel. These results apply to average (expected) values of economical indicators. When these inlicators vary within the limits of the assumed error (up to ?25-30%), the conclusions are correct only when; iu- clear energy can compete with natural gas from Northern deposits, i.e., when the economical indicators of its production and transportation are unfavorable. The scale of development of nuclear power has a significant effect on the terminal expenditure on fuel, i.e., on the relative economical indicators that characterize the actual cost of production of additional quantities of fuel in the given region of the USSR.t Figure lc shows the terminal expenditures on fuel in the principal regions of the USSR as a function of the total capacity of nuclear power plants (for minimum re- sources of high-quality fuel). It is seen that the terminal expenditures on fuel remain stable only in the Siberian regions (where they are governed mostly by the cost of mining and transportation of Kansk-Achinsk coal), and decrease by 3-4 rubles/ton of fuel with the growth of nuclear capacity in most regions of the USSR. This once more confirms the high economical efficiency of nuclear power. At the same time, one should consider that all measures of the current decade (1970-1980) for rationalizing the structure of the fuel?energy balanceby increasing the share of petroleum and gas will provide a gain of only 2-3 rubles/ton of fuel. Together with terminal expenditures on organic fuel, the analysis also gave an estimate of the maximum admissible expenditure on nuclear fuel (in it fuel equivalents), i.e., of the maximum cost of its production at which nuclear power plants still can compete with conventional plants (Fig. 1c). Effect of Principal Economical Factors on Optimal Levels of Growth of Nuclear Power. In the preceding section we have discussed the effect of the pace of growth of nuclear power on the long-term structure of fuel-energy balance of the USSR and on the growth of other branches of our power economy. Another task of the analysis is to determine how the power economy, in turn, affects the optimal level of growth of nuclear power. For this purpose we have analyzed its dependence on the following economical factors: 1. In determining the limits of applicability of nuclear power it is of decisive importance to find the maximum possibilities of its territorial coverage. The mathematical model made it possible to con- sider the feasibility of constructing nuclear power plants in practically all regions of the country. The results of optimization obtained on this premise are illustrated in Fig. 2. The horizontal axis represents (in relative *See the straight portion of the sloping line that represents the fuel equivalent of nuclear power in Fig. la. t The economical sense and methods of calculation of the terminal expenditure on fuel have been discussed in detail in [2]. 217 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 units) the required increase in electrical power capacity of the country for the last decade of the design period divided into 11 principal regions. The regions are arranged in the order of decreasing calculated cost of production of electric power by fossil-fuel steam power plants (SPP). These costs were calculated for base steam power plants for two limiting values of terminal expenditure on fuel corresponding to mini- mum and maximum gas production with no restrictions imposed on the growth of nuclear power. In this manner we have found the territorial variation of the cost of production of electricity by steam power plants that directly compete with nuclear power. plants. The results of this competition are illustrated in Fig. 2 which shows the calculated average and minimum costs of electricity produced by nuclear power plants. These costs are assumed to be the same in all regions of the USSR. As seen in the figure, in the European USSR (including Ural) in Central Asia, and in the far East the economical indicators of nuclear power are better than those of steam plants. In West Kazakhstan nuclear power is economically inferior to steam plants for average indicators and prac- tically equal for minimum indicators. Only in the principal regions of Siberia is nuclear power uneconomi- cal even under the most favorable conditions in comparison with electrical plants burning Kansk-Achinsk coal or with large hydroelectric power stations. Thus, from the point of view of territorial coverage, the limits of applicability of nuclear power include all regions of the country except Siberia and East Kazakhstan. These regions should contribute more than two thirds of the total growth of electric power consumption of the USSR. 2. However, not all this growth can be satisfied by nuclear energy. This is due, first of all, to the lower relative economical advantage of nuclear power plants operating under variable loading conditions. A simple economical comparison indicates that nuclear plants are competitive with steam plants operating from 7000 to 4000 h/year in regions of high fuel costs (North-West) and from 5000 to 6000 h/year in regions of low-cost fuel (Ural), i.e., they cannot participate in meeting the demands of the peak and semipeak portions of the electrical load curve. This leads to an additional reduction of the limits of applicability of nuclear power (in addition to the limitations due to territorial considerations) but does not provide a basis of a quantitative evaluation of this restriction. In fact, by partially shifting the existing power plants into the variable-load portion of the load curve and by designing specialized peak- load and regulated plants it is possible to transfer a considerable portion of the anticipated electric load into that portion of the curve were nuclear plants can compete with conventional steam plants. To study the effect of operating conditions on the maximum capacity of nuclear power plants, opti- mal structures of the Unified European Power System (as part of the fuel?energy balance) were computed for two extreme configurations of the possible composition and regulation capabilities of conventional power facilities. The second Unified European Power System structure corresponds to limited (as compared to the first) regulation capabilities. The results were used to plot the maximum (from the point of view of operating conditions) capacity of nuclear steam power plants (NSPP) and the resultant gain infuel? energy balance as a function of the regulation capability of nuclear and conventional plants (Fig. 3). It is evident that the capacity of nuclear plants decisively depends on the type and regulation capabilities of conventional power facilities. Thus, under constant-load operating conditions most favorable from the point of view of nuclear power generation (i.e.. without nightly off-load operation for periods of 7000 h/year) the variation of structure and regula- tion of conventional power plants within the investigated limits leads to a reduction of the maximum capacity of nuclear plants by 15-18% and to an overexpenditure of about 700 million rubles in fuel?energy balance. This proves the great economical importance of improved regulation of electrical power equipment installed in the past period. These measures justify themselves by subsequent savings even if their initial costs amount to 10-15 rubles/kW in additional capital investments or to an increase of relative fuel consumption by 30-50 g/kWh. The second important measure for increasing the capacity of nuclear power plants is to improve the regulation capabilities of nuclear facilities, i.e., to provide technical means for efficient operation under off-load conditions. Although this measure reduces the economical performance of the plants (in spite of high initial capital investment the advantage of low-cost electricity is lost), the measure is on the whole advantageous for the Unified European Power System and the fuel?energy balance as it compensates to a large extent the inadequate regulation capabilities of other facilities and provides great savings by increasing the total capacity of nuclear power plants (see Fig. 3). .218 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 1,1 ? 40 0.9 0,8 i. 0,7 0,6 0,5 O ay 8 ? 43 `:',4) 42 07 ., 4,..., .....f., Itmti.,\ No tiklw [ Center I Caucasus Ukraine ;- 1_ Volga region Central Asia West Siberia East Siberia Total growth of required electric plant capacity, ?Fig. 2 1000 800 .0 o Declassified in 10 9 8 20 10 Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 /2 ?, / ///2, / / , A / 'W _ - v. 4r4r;le A L A . I _ 4 _ 1 il . 1 Ratio of capital investment in NPP and SPP Fig. 4. Optimal capacity of nuclear plants as a function of the ratio of capi- tal investments in nuclear and con- ventional plants at different levels of electricity consumption. plants are at present unknown. Thus, in this study we have to deal with the least favorable conditions (assuming, in par- ticular, reactors designed for steam plants). Under such premises, nuclear heat plants can be economically used in special heat-consuming centers with heat demands of at least 2.5 thousand Gcal/ h. Under such unfavorable conditions the ca- pacity of nuclear heat plants is relatively low and amounts to 8- 10% of the total capacity of nuclear power plants (with respect to reactor power). Taking into account the combined effect of territorial and load factors, the economically sound maximum capacity of nuclear steam and heat power plants at the end of the con- sidered period turns out to be 30-33% of the total installed capacity of all power plants of the country, 42-47% of plant capacity of the European regions, 48-50% of the increase of total installed capacity in the last decade of the considered period, and about 70% of the increase in capacity of the Euro- pean regions. The growth of nuclear power within the limits of its maximum range can change substantially as a result of the following energo-economical factors. 4. The absolute economically sound capacity of nuclear plants is affected to a large extent by the resultant level of electricity consumption in the entire country. If the pro- portion of the European and Siberian regions remain un- changed, a one percent decrease in the demand in electricity results in a linear reduction in nuclear power capacity equal to about 2%. However, if the production of electricity de- creases selectively at the expense of the Siberian or, on the contrary, the European regions, the maximum capacity of nuclear power remains the same in the first case, or drops by 3% for every percent of reduced production in the second case. 5. The optimal scale of incorporation of nuclear power plants into the power network depends also on organic fuel resources that can compete with nuclear power. As noted before, to such resources belong in the first place the Siberian and Kazakhstan coal deposits (provided they are used locally) which limit the territorial extent of nuclear power predominantly to the European regions of the USSR. But even here can nuclear power meet with competition from organic fuels such as natural gas and fuel oil. As shown by several different versions of optimal fuel?energybalance, the variation of high-quality fuel resources within the actual limits does not affect the capacity of nuclear power plants, and only an in- crease in gas production leads to a reduction of maximum capacity by 15-20%. On the whole this factor can be neglected in this study. 6. Obviously, the optimal capacity of nuclear plants significantly depends on the resultant combina- tions of their own economical indicators and the indicators of competing power plants, as well as on the production and transportation of fuel. Consequently, the possible changes in optimal capacity of nuclear plants can be correctly evaluated only after a careful study of the region of uncertainty of the optimal fuel ?energy balance due to the combined effects of the error in the economical indicators of all objects con- sidered. Figure 4 shows the results of such a study in the form of a dependence of the optimal capacity of nuclear plants on the factor K = KNpp/ Kspp, i.e., on the ratio of the capital investments in nuclear and conventional plants. This statistical ratio is of primary importance for understanding the prospects of nu- clear power. 220 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Total economical loss, millions of rubles 1400 1200 7000 800 600 400 200 I ; 1 I L ? 10 20 30 40 50 50 70 Constraints on total nuclear plant capacity, million kW 80 100 90 80 221,) 70 20 -E `,1 6 18 - 14.2 12 g 4 z 10 8 8 a) 2 r 4 2 90 100 1 -1- .. 2 I ' 0- . 1 _ i.? 4 r . . , ? II Design levels Fig. 5 Fig. 6 Fig. 5. Economical consequences of constraints on the total capacity of nuclear power plants. Fig. 6. Possible trends in the growth of nuclear power: 1) theoretical; 2) calculated; 3) rational; 4) probable. In fact, the figure shows that as long as the relative capital investment in nuclear plants does not exceed the corresponding investment in steam plants by more than 40-50% (K s 1.4-1.5), the growth of nuclear power does not depend very much on the economical indicators of other energy resources and can be aimed for practically the maximum level (with only a 10-15% deviation from maximum capacity). The optimal capacity drops sharply when the factor K exceeds 1.4-1.5 ("discontinuity" point). An increase of the capital investment ratio from 1.5 to 1.7 results in a 30-35% drop in optimal capacity. If the ratio increases still more (1.7-1.8), the rate of the decrease slows down as nuclear power is becoming less and less economical in regions of high-cost fuels. Nevertheless, for the maximum capital investment considered (K = 1.8), the economically sound capacity of nuclear power plants is only 40-50% of its maxi- mum value. Effect of Nonenergetic Factors on the Growth of Nuclear Power. Although the nuclear power growth levels discussed in the preceding section are optimal from the point of view of fuel-energy balance, they may prove unrealistic or unrealizable in practice for the national economy as a whole. This circumstance is now difficult to evaluate economically but its effect can be taken into account in the form of nonenergetic (national-economic) constraints on the growth of nuclear power, for example, on the total capacity of nu- clear power plants. We have thus investigated not only the optimal scale of growth of nuclear power but also the energo-economical consequences of the reduction of nuclear plant capacity below optimum. For this purpose, using average values of economical indicators, we have computed a large number of optimal fuel-energy balance versions in which the constraints on the total capacity of nuclear plants were varied for different levels of electricity consumption and different high-quality fuel resources. These computations made it possible to determine the economical loss to the power economy caused by different constraints on the total capacity of nuclear power plants. The magnitude of the loss was computed as the difference in the optimal total calculated power cost, i.e., of a functional of the fuel-energy balance model, when passing from unlimited (100%) nuclear plant capacity to increasingly "stringent" constraints on its total value. The curves shown in Fig. 5 clearly illustrate the fact that as the total nuclear power capacity is re- duced more and more the magnitude of the total and relative economical loss increases first very slowly and then, after reaching some critical value, very rapidly. This behavior has a quite obvious explanation. Forced reduction of the capacity of nuclear plants eliminates nuclear power first from the regions where 221 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 fuel is inexpensive and at the same time from the variable portion of the electrical load curve in regions of expensive fuel where nuclear power is attractive because of the excessive growth of relatively costly hydroelectric storage plants, the shift of existing plants to less favorable portions of the load curve, and other not very effective measures. In other words, in the presence of restrictions nuclear power plants are first forced out from regions where their economical performance is nearly equal to that of conven- tional plants so that the economical loss is relatively low. After all such possibilities are exhausted, fur- ther restrictions on capacity of nuclear plants affect their use as base load plants in regions of high-cost fuel resulting in a rapid increase of the relative and total economical loss. The dashed lines in Fig. 5 show the approximate location of "discontinuities" in the dependence of the economic loss on the magnitude of constraints on nuclear capacity. An analysis of the position of these points in relation to optimal capacity proved that it remains nearly the same under different conditions of electricity consumption and gas resources, and is governed by the nuclear plant capacity equal to approxi- mately 80% of the corresponding optimal capacity. Thus, a drop in nuclear plant capacity by up to nearly 20% of optimal capacity is not associated with any significant losses to the power economy. In fact, the overall loss in the fuel?energy balance amounts in this case to only 20-100 million rubles (depending on the level of consumption and gas resources), i.e., about 7% of the maximum possible loss. The relative loss does not exceed 2-3 rubles/kW. At the same time, such a 20% drop in nuclear plant capacity is, apparently, justified from the point of view of the national economy as a whole in view of the difficulties in the transfer of certain nonenergetic branches to nuclear power operation and as an insurance against the possible rise in cost of nuclear power plants. The resulting values of total capacity of nuclear plants were additionally corrected in order to allow for dynamical factors. Tha meaning of this correction is illustrated in Fig. 6 which shows the different trends in the growth of nuclear pla -it capacity with time. The first curve in this figure corresponds to the maximum (from the point of view of economical considerations) rates of growth of nuclear power that can be obtained theoretically in the case of total ab- sence (starting with the present five-year period and on) of nonenergetic constraints on the capacity of nu- clear power plants. The portion of this curve corresponding to the first and second time levels has been plotted from the results of optimization of fuel?energy balance assuming that nuclear plants are placed in the entire unoccupied portion of the base load growth in regions were such plants can be effectively used. The second portion of the curve was plotted from the results of the above-mentioned analysis. As seen in the figure, the average annual rate of growth of nuclear plant capacity is under these premises nearly 15%, i.e., leads by a factor of nearly two the rise in electricity consumption. The second and third curves in Fig. 6 characterize the calculated optimal and acceptable (allowing for a 20% reduction) capacities of nuclear power plants. Both are based on the now accepted scale of develop- ment of nuclear power up to the second design level and proceed from the admittedly incorrect assumption that the conditions of development of nuclear power between the second and third levels are practically the same. In reality however the effect of several factors that tend to restrain the growth of nuclear power will be felt during the first years of this period. The fourth curve of growth of nuclear power corresponds to the case when the effect of these factors would be felt only for two-three years. However, if the restraining factors act for a longer time, the growth of nuclear power is represented by the lowermost curve in Fig. 6. Thus, the no longer optimal but rational (from the point of view of the national economy) probable capacity of nuclear plants in the last years of the design period amounts to 65-70% of the economically sound maximum capacity. The above discussion leads to the following conclusions. 1. Together with Tyumen natural gas and Kansk-Achinsk and Kuznetsk coal, nuclear energy will be- come in the considered period one of the basic sources for satisfying the growing demand for elec- tric energy in?the USSR. The analysis indicates that it would be careless and economically ex- pensive to strive for maximum development of any one specific resource at the expense of others; all four resources should be intelligently combined. 222 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 2. Such a moderate strategy of the development of the power economy corresponds to the incorpora- tion of nuclear power plants with a capacity equal to 65-75% of the economically sound maximum capacity. The fraction of nuclear power in the overall output of all energy resources would then amount to 7-8% at the end of the considered period; the proportion of nuclear power plants in the structure of all generating plants will reach in the same period 20-23%, including 32-36% in the European regions (and about one half of the total output). 3. The indicated growth levels of nuclear power are economically sound even if their relative capital investment is 60-65% higher than the corresponding figures for conventional steam power plants. However, with still higher costs, the rational capacity of nuclear power plants should be decreased by 15% for every 10% increase of the investment ratio. 4. These growth levels require that nuclear power plants have an adequate regulation range that would allow off-load at least 20% below the average total installed capacity. If this condition is met, nuclear power plants will be efficiently employed not only for base load operation but also in the 17-18 h region of the 24 h load curve for annual loads of 5800-6000 h/year. If such a regu- lation range cannot be ensured the loss to the national economy will amount to 10-15 rubles/ kW. 5. The following simple rule of the growth of nuclear energy corresponds to the set of the above "boundary" conditions: irrespective of the level of electricity consumption and other energo- economical factors, nuclear power plants should be used to provide for the growth of base (at the expense of heat plants using organic fuel) and some of the fluctuating (in accordance with the preceding paragraph) electric load in the North-West, Center, Ukraine, and Caucasus, without any special effort to extend their use to other regions of the USSR (with the exception of remote regions with specific climatic conditions) and without applying them to still more fluctuating por- tions of the load curve. 6. This work is only the first in a general study. Its aim is to give a deep understanding of the problems associated with the growth of nuclear power and the analysis of long-term development of the power economy including the role of fast reactors The study is a joint work of the staff of four organizations under the Scientific guidance of Academician L. A. Melent'ev. The work was done with the participation of G. V. Agafonov, I. M. Vol'kenau, E. A. Volkova, M. V. Sapozhnikov, G. E. Tkachenko, and L. D. Khabachev. LITERATURE CITED 1. A. A. Makarov, A. S. Makarova, and A. N. Zeilinger, Ekonomika i Matem. Metody, No. 6, 424 (1970). 2. A. P. Andreev et al., Teploenergetika, No. 8, 68 (1967). 3. V. V. Batov and Yu. I. Koryakin, Economics of Nuclear Power [in Russian], Atomizdat, Moscow (1969), p. 41. 223 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 PHYSICAL AND TECHNOLOGICAL PRINCIPLES OF THE CONSTRUCTION OF ATOMIC POWER STATIONS WITH GAS-COOLED FAST REACTORS AND USING A DISSOCIATING COOLANT. NITROGEN TETROXIDE* A. K. Krasin, V. B. Nesterenko. N. M. Sinev. V. P. Slizov, V. I. Khorev. B. I. Lomashev, V. P. Bubnov. B. E. Tverkovkin, and V. A. Naumov UDC 621.311.2:621.039 One of the ways to obtain effective characteristics for gas-cooled fast reactors may be to use the dissociating compound nitrogen tetroxide (N204) as the coolant and working fluid in single-loop schemes of atomic power stations [1-7]. There exists a large class of inorganic polyatomic gases in which there are thermally reversible chemical reactions of dissociation and association taking place with thermal effects and changes in the num- ber of moles. The most interesting of these is a dissociative system that has been studied in considerable detail: N201 1 2NO2 ? 149 kcal/kg 2NO 02 ? 294 kcal/kg The temperature range of the first stage of the reaction is 21-170?C, and the range of the second stage is 140-850?C. The most important physicochemical properties of N204 are given in [1]. Both the first and the second stages of the reaction are accompanied by large thermal effects, and this has an important influence on the thermophysical properties of the dissociating gases. At the Nuclear Power Institute of the Academy of Sciences of the Belorussian SSR (IYaE AN BSSR) we conducted a comprehensive study of the thermophysical properties of N204 over a wide range of tem- peratures and pressures. including a study of: viscosity at 30-500?C and 1-150 abs. atm; p?v?t properties at 50-525?C and 8-125 abs. atm; enthalpy at 20-150?C and 100-170 abs. atm; the composition of the gas at 200-4,00"C and 1-7.5-abs. atm [1]. We conducted extensive investigations of the heat exchange taking place during boiling and condensation, as well as the convective heat transfer taking place during cooling. The investigations showed that the effective thermal capacity and thermal conductivity of N204 are 3-9 times as high as the usual thermal capacity and molecular thermal conductivity of nondissociating gases. This means that there is a sharp increase in the heat-exchange coefficients (for example, the maximum values of the effective thermal capacity of N204 at 1 abs. atm and 80-100?C are as high as 2.4 kcal/ kg ? deg. In the nonisothermal flow of a gas taking part in a chemical reaction, there is, in addition to the heat exchange through thermal conductivity, a substantial amount of heat exchanged in the form of chemical enthalpy through concentration diffusion. Chemical enthalpy may constitute a very large part of the total heat-transfer balance. In the experimental investigations conducted at the IYaEAN BSSR, with turbulent flow of N204 in a heated pipe at supercritical pressures of 115-160 abs. atm and temperatures of 250-530?C, *Fourth International Conference. on the Peaceful Uses of Atomic Energy (Geneva, September, 1971). Report No. 431. 224 Translated from Atomnaya Energiya, Vol. 32, No. 3, pp. 197-203. March, 1972. ? 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without permission of the publisher. A copy of this article is available from the publisher for $15.00. Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 heat-flux densities of (1-7) ? 105 kcal/m2 ? h, and Reynolds numbers of (1-2.5) ? 105, we found that the heat- exchange coefficients were 18,000-42,000 kcal/m2 ? h ? deg, or 3-9 times the values obtained for inert gases under comparable conditions. Because of the high thermophysical properties of N204, its use as a coolant results in effective heat removal from the active zone of a fast reactor with a thermal power of 2500-3000 MW at an average power density of 500-600 kW/liter in the active zone, a maximum temperature of 700-720?C in the fuel-element jackets (taking account of local overheating factors), a gas temperature of 520-540?C at the reactor outlet, a pressure of 140-170 abs. atm, and gas heating values of 230-270?C. The most distinctive and advantageous features of an N204 dissociative system is the possibility of using the molecular weight, which varies with the temperature (and therefore with the gas constant R), for bringing about a considerable increase in the effective efficiency and specific power of a gas-turbine cycle. Gas cycles using helium and CO2 have some well-known drawbacks; in such cycles, the coefficient of use- ful work of a cycle at 500-650?C is very low (co = (Lt ? Lc)/ Lt = 0.2-0.35). In the case of N204, since the reactions are reversible and the chemical composition changes from N204 2NO2 at 25-30?C with a gas "constant" of 9.2 to NO + 02 at 650-850?C with a gas "constant" of 27.6, it is possible to have the ther- modynamic gas cycle take place with different gas "constant" values in the turbine and in the compressor. The gas "constant" in the compressor is lower than the value in the turbine, and therefore it is possible to reduce the fraction of the power used up in the compression and circulation of the gas, to 30-45%, while in the case of helium the fraction is 70-80% at temperatures of 500-650?C. As a result, an N204 heat cycle has a much higher coefficient of useful work and a greater effective efficiency than inert-gas cycles [4-6] (Fig. 1). A comparison of the gaseous coolants (He, CO2, and N204) used in fast reactors with respect to the pump-through parameter (the specific power used up in circulating the gas, per unit of transmitted thermal power) at 100-120 abs. atm and 300-600?C showed that the specific power consumption for N204 is less (by a factor of 7-8) than for CO2 and for He. In the system N204 2NO2 2N0 + 02 the parameters of the line of saturation are such that it can be used for gas?liquid cycles with the low parameters of 1.3-1.7 abs. atm and 27-32?C condensation tem- perature [1]. Nitrogen tetroxide has a low heat of vaporization (5.5 times the value for water). This makes it pos- sible to simplify the heat regeneration scheme in the gas?liquid cycle, since the heat of the gases leaving the turbine is quite sufficient not only for heating the liquid to vaporization but also for superheating the gas in the regenerator by 100-200?C. In the regenerator, the chemical reactions of dissociation take place on the high-pressure side with a lower heat of chemical reaction (149 kcal/kg) than on the low-pressure side in the process of recombination (294 kcal/kg), and as a result, in gas?liquid cycles using N204 it is possible to achieve a higher degree of heat regeneration than in cycles using water or CO2, and conse- quently to obtain better values for the thermodynamic indicators. Thus, if N204 is used, it is possible to construct an atomic power station with a gas-cooled reactor, as well as to use a thermal scheme involving a gas?liquid cycle with regeneration. Our study of the mechanism of the chemical reactions and kinetic constants of N204 showed that in the gas-dynamics calculations of turbines and heat exchangers, as well as in calculations of heat-exchange coefficients, we had to take account of the time characteristics of the dissociation and recombination. Our estimates of the time of chemical relaxation indicated that the first stage of the reaction (N20 2NO2) takes place under equilibrium conditions in 10-6-10-8 sec, and in the second stage of the reaction (2NO2 2N0 + 02) the relaxation times may vary between 10-3 and 10-1 sec, depending on the thermodynamic param- eters. The investigations showed that among the many possible schemes for atomic power stations, the scheme with high efficiency is one with intermediate heat regeneration between the high- and low-pressure turbines. At intermediate regeneration pressures of 15-25 abs. atm, we obtain full regeneration effec- tiveness and practically eliminate the influence of the kinetics of the chemical reactions. In gas cycles using N204, in order to make sure of achieving high thermodynamic efficiency, the lower pressure of the cycle should be chosen in the 8-10 abs. atm range, which provides .a sufficient approximation to equilibrium processes. Thus, gas turbines using N204 do not require a vacuum. 225 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 0, 4 2 500 550 600 t,?C Fig. 1. Comparison of the coefficients of useful work of the thermodynamic cycles of atomic power stations using gas- cooled fast reactors. Gas?liquid cycle: 1) N204, p = 240 abs. atm; 1') N204, p = 150 abs. atm; 2) CO2, p = 240 abs. atm; 2') CO2, p = 170 abs. atm. Gas cycle: 3) N204, p = 150 abs. atm; 3') He. For this reason, the use of N204 as the working fluid of turbines has a number of advantages over the use of other working fluids (for example, steam). At a pressure of 1.3-1.7 abs. atm downstream from the turbine, the specific volume of N204 is between 1/34 and 1/40 of the specific volume for steam at a condenser pressure of 0.035 abs. atm. This makes possible a considerable increase in the power of the gas turbine for one exhaust, bringing the value up to 1000 MW [1] (for last-stage dimensions analogous to those of steam turbine). In addition, the use of N204 makes it possible to construct the flow-through part with a small number of stages, since the isentropic drops for N204 are 2.5 times the values for steam. Design studies on a 1000 MW gas turbine using N204 yielded the following basic characteristics: gas- flow rate 2930 kg/sec; gas pressure and temperature before the turbine: 130 abs. atm and 565?C; pressure and temperature before the low-pressure turbine: 15 abs. atm and 91?C; pressure downstream from the turbine, 1.5 abs. atm; speed 3000 rpm. The high-pressure turbine has six stages, with a constant rotor root diameter of 1000 mm for the case of first-stage guide vanes 430 mm high. The low-pressure turbine has two stages. The average diameter of the last stage is 1660 mm, the height of the guide vanes is 560 mm, the velocity at the outlet of the rotor is 122 m/sec. In order to reduce the load on the thrust bearing and the dimensions of the guide vanes, the turbine is constructed with twin exhaust. The weight of the tur- bine is 390 tons. Our design calculations indicated that it is possible to construct an N204 turbine with a metal volume that is less by a factor of 4-4.5 than the volume for a steam turbine, with better aerodynamic indicators (there is no moisture in the flow-through part of the turbine) and considerably smaller dimen- sions. It is quite realistic to envisage the construction of a single-shaft two-stream turbine with a power of 2000-3000 MW in a single unit [1]. The results of the experimental investigation in the 25-550?C range indicate that N204 has sufficient thermal and radiation stability for practical use in atomic power generation. We made an experimental study of the radiation stability of N204 and calculated the radiolysis of this coolant in the n?y radiation field of a fast reactor. In the dissociative system of the coolant (N204 2NO2 = 2N0 + 02), as a result of the specific properties of fast reactors, radiochemical effects are caused by the action of n?y radiation at a high dose rate (approximately 1019 eV/cm3 ? sec) at contact times of ?10-2 sec in the active zone, gas tem- peratures of 200-550?C. and pressures of 130-170 abs. atm. The products of the radiolysis of NO2 are N20, N2. and 02. The fraction represented by the decomposing coolant in a fast reactor is approximately 10-6. A large number of structural steels and alloys were studied experimentally in an N204 medium at tem- peratures of 25-700?C and pressures of 1-150 atm, and the results showed a high degree of corrosion,re- sistance (0.001-0.005 mm/g) [8]; the materials included Kh18N10T, EI-847, ET-629, EI-654, 3KhV, EI-432, 226 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 600 500 40 3:3 N2, kW/liter .do 1; 100 150 200 Fig. 2 250 p, abs. atm NI, kW/liter Fig. 2. Specific thermal power density of gas-cooled fast reactors using N204, He, and CO2, as a function of the pressure at the reactor outlet, for D/H = 2 and maximum jacket temperature 720?C: 1, 2) N204 (dfuel element -.= 6.4 mm, tgas = 350 and 540?C, reactor pressure drop 12 and 10 abs. atm, respectively); 3, 5, 7, 8) He (tgas = 670?C, 0.9, 0.95, 0.95. 0.9; dfuel element = 6.0, 6.0, 6.4, 6.4mm, respectively); 4,6) CO2 (tgas = 570?C; Ap = 16 abs. atm; dfuel element - 6.0 and 6.4 mm, respectively. Fig. 3. Specific thermal power density of gas-cooled fast reactors as a function of the relative flattening of the active zone for Pgas - 150 abs. atm and maximum jacket temperature 720?C: 1) N204 (tgas = 540?C; Ap = 10 abs. atm; dfuel element =6.4 mm); 2) He (tgas = 670?C; Ap = 16 abs. atm;dfuel element - 6 mm); 3) CO2 (tgas = 570?C; Ap = 16 abs. atm; dfuel element - 6 mm); 4) He (tgas -,670?C; 6p = 8 abs. atm; dfuel element = 6 mm). Kh25, and 3Kh3, as well as aluminum and titanium alloys, high-chrome cast iron, AG-1500 graphite, siliconized graphite. Teflon, and other materials in the operating range of temperatures. For the fuel- element jacket materials Kh18N1OT and El-847, a high corrosion resistance was found at temperatures of up to 800?C in tests lasting 10,000-12,500 h. The introduction of 0.7-1% NO into the coolant considerably reduces the corrosion of metals (to 0.02-0.05 mm/g) in the phase-transformation zone in the coolant. Experimental studies in coolant technology are now being conducted at the Institute. A number of thermophysical test stands with N204 pressures of up to 150 abs. atm and gas temperatures of up to 550?C have been set up. These test stands were used for studying the thermophysical, gas-dynamics and cor- rosion characteristics of N204,, for practical tests of techniques for the cleaning, repairing, filling, and emptying of loops in experimental heat-power units with ratings of up to 1000 kW. Some of the test stands have been in operation since 1965. An experimental power unit with a two-stage gas turbine having a useful power of 100 kW at gas temperatures of 500-520?C and pressures up to 6 abs. atm before the turbine, a gas flow rate of 1 kg/ sec, and a test-stand thermal power of 1000 kW was tested in 1968. The first tests on this stand were conducted in order to study the stability of the parameters of the gas-liquid cycle with a maximum temperature of 504?C and a pressure of 5 abs. atm, with a circulation of 360 kg of N204 in the loop and a test-stand power of 1070 kW. The tests were continued for 372 h. During 170 h of continuous testing there were more than 1000 cycles of reversal of the dissociating coolant. The experiments showed a high stability for the parameters, and there were no indications of irreversibility in the coolant. Complete reversibility of the gas-liquid cycle was also achieved in 1966-1967 in experiments con- ducted on a closed-circulation stand for 600 h using 80 kg of N204 at temperatures of 30-550?C and pres- sures of 10-60 abs. atm. Altogether, the stand was used under operating conditions for more than 1500 h, including 1000 h with no replacement of the N204. Chemical analysis of samples failed to indicate any ap- preciable change in the quality of the coolant (the impurities amounted to less than 0.8%). 227 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 TABLE 1. Physical Characteristics of Unprofiled Fast Plutonium Reactors with Power Ratings of 1000 MW (Electrical) Parameter Coolant Na N20a Flattening, DM 1 1,69 2,5 1 1,69 2,5 Critical charge of Pt?", tons 1.91 4,93 2,00 1,89 1,91 1,96 Critical enrichment of Pu2", % 9,00 9,10 9.36 8.88 8.99 9,25 Conversion ratio of active zone 1,242 1,226 1,183 1,256 1,237 1,194 Conversion ratio of side blanket 0,299 0,226 0,182 0,341 0,263 0,209 Total conversion ratio 1,657 1,661 1.677 1,727 1.731 1,752 Fuel resiStanCe tittle its active zone, yr 1,120 1,126 1,181 1,160 1,14 1,211 Fuel resistance time in side blanket, yr 3,56 4,43 5,42 3,21 3,93 4,84 Doubling time, yr 7,67 7,60 7,58 7.14 7,00 6,93 At thelYaEAN BSSR, tests were conducted for 500 h during 1969 and 1970 on a closed stand with a gas?liquid cycle in order to study the characteristics of heat exchange at gas parameters of 80-160 abs. atm pressure and 250-540?C temperature. These tests also indicated complete reversibility of the conden- sation cycle. From all of the above-mentioned experiments, it was possible to conclude that the gas?liquid cycle is completely reversible and N204 is usable in practice as a working fluid for power plants. In order to determine the prospects for the use of He. CO2, and N204 as gaseous coolants at atomic power stations with gas-cooled fast reactors, the efficiency and the coefficient of useful work of single-loop gas cycles and condensation cycles were compared under conditions of maximum thermal power density and using identical surface heat exchangers in each case. The thermophysical characteristics were studied by using fuel elements with metal jackets and a fuel composition of UO2 + 30% Cr and Ni (Pu02 + 30% Cr, Ni) [11]. The maximum temperature of the ceramic fuel was 1300-1400?C, and the power density was 450- 500 kW/liter, i.e.. equal to the accepted value for the active zone of the BN-1000 sodium-cooled fast reac- tor [7]. Currentexperience in the development of metal vessels lined on the inside with stainless steel has shown that it is possible to construct vessels that can be transported (by railroad) for fast reactors with thermal power ratings of 2600-3000 MW. At operating pressures of up to 170 abs. atm, such vessels should have a maximum diameter of 4.2-4.3 m and a height of 10-12 m. They can house an active zone with a diameter of 2.2-2.4 m. with a thermal power density of 500-600 kW/liter. Figure 2 shows the variation of the thermal power density of gas-cooled fast reactors of single- loop atomic power stations using He, CO2. and N204 at various pressures. Fnr the He and N204 gas cycles the maximum gas temperature varied between 500 and 670?C, while for the CO2 and N201 gas?liquid cycles the range was 350-650?C. The investigations made it possible to identify the preferable range of gas parameters if a thermal power density of 450-500 kW/liter was to be attained: a temperature of 670?C and pressures of 200-240 abs. atm for He; 550-570?C and 190-220 abs. atm for CO2; 520-540?C and 120-140 abs. atm for N204. With N204 at a temperature of 350?C, the same thermal power density values are also attained at pressures of 80-120 abs. atm. and a thermal power density of 650 kW/liter at 180 abs. atm. In designs for gas-cooled fast reactors using N204. it is desirable to use a pressure of 150-170 abs. atm. a reactor outlet temperature of 520-540?C. and a specific thermal power density of 550-600 kW/liter. A further increase of the specific thermal power density can be achieved by flattening the active zone to a D /11 value of 3.5-4.0. Increasing the active zone flattening D/H from 2.0 to 4.0 at 170 abs. atm makes it possible to increase the specific thermal power density: with CO2 it is increased from 420 to 530 kW/liter, with He from 450 to 540 kW/liter, and with N204 from 550 to 650 kW/liter at 540?C and from 640 to 700- 720 kW/liter at 350?C (Fig. 3). However, for such flattening values it is necessary to construct and test reactor vessels of prestressed reinforced concrete for pressure levels of 160-180 abs. atm. We compared the physical characteristics of a number of plutonium reactors, all with cylindrical geometry. a power rating of 1000 MW (electrical), and identical active-zone composition but different values of flattening. As a result of our calculations, we obtained the fundamental physical characteristics for unprofiled reactors (Table 1). 928 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 TABLE 2. Physical and Thermophysical Characteristics of a Fast Plutonium Reactor with a Power Rating of 1000 MW (Electrical) In the calculation in the active zones and degree of burnup is 10%. 18-4 RZ 15-B) [9]. The tern of constants [6]. Characteristics N204 Na Average thermal power density, kW/liter 525 400 Flattening 1.25 3 Percentage of the volume that is repre- sented by: fuel 33 50 coolant 32 33 steel 35 17 Diameter of fuel element (thickness of jacket), mm 6.2/0.30 7.8/0.30 Diameter and height of active zone, m 2.0/1.6 Thickness of breeding blanket, mm 400 400 Matrix fuel Pu02 + UO2 + 300/,, CrPu02 + UO2 Relative dimension of inner zone 0.52 0.53 Critical charge, tons 1.800 2.4 Enrichment, % 9.6/12.8 10.1/13.5 Total conversion factor 1.51 1.50 Average conversion factor in active zone-- 1.00 0.95 Depth of bumup, 10 10 Doubling time, yr 7.5 8 Coolant pressure, abs. atm 164/154 ?1-7 Coolant temperature, ?C 240/540 400/580 Maximum temperature of fuel element (taking account of overheating factors), ?C 730 700/720 Maximum temperature of fuel core (taking account of overheating factors), ?C 1280 2400 of the doubling time is is assumed that the time required for reprocessing the packs blankets is one year, the average power density is 500 kW/liter, and the maximum The calculations were carried out for a two-dimensional 18 group program (MIFI group constants used in this program were obtained on the basis of the BNAB sys- 35 30 2 2 7/ APS - , PEWEE ill AIWA KR UU UtIV 5 Ix Fig. 4. Comparison of the thermo- dynamic efficiency of single-loop atomic power stations using gas- cooled fast reactors. Gas-liquid cycle: 1) N204, p = 240 abs. atm; 2) p = 170 abs. atm; 3) N204, p = 150 abs. atm; 5) CO2, p = 240 abs. atm; 6) CO2, p = 170 abs. atm. Gas cycle: 4) N204, p = 150 abs. atm; 7) He, p = 240 abs. atm; 8) He, p = 170 abs. atm. An analysis of the calculation results shows that because of the harder spectrum and the higher leakage of neutrons from the active zone of the reactor in the case of the dissociating gas, the conversion factor can be increased by 3% over the value for sodium, the doubling time can be reduced by about 0.6 years, and the characteristics can be made more favorable from the point of view of plant safety (the effect of coolant re- moval is only one-half to one-third as great). Calculations were carried out for gas reactors with carbide and matrix fuel compositions. Using carbide fuel reduces the doubling time by one year, which will make it possible in the future to obtain a doubling time of less than six years. The physical characteristics of a reactor using N204 with a matrix fuel composition are no worse than those for a sodium coolant. Table 2 shows the comparative charac- teristics of reactors using N204 and Na; it can be seen from this that the reactor using N204 has the preferable physical charac- teristics. In determining-the effectiveness of single-loop schemes for power stations using gas-cooled fast reactors, the hy- draulic resistances of the main units were chosen on the as- sumption that for each coolant the maximum thermal power 229 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 density was achieved in the reactor, and the minimum surface areas were assumed for the heat ex- changers. Figure 1 shows a comparison of the coefficient of useful work of gas cycles using He and N204 with gas?liquid cycles using CO2 and N204 In the gas-turbine cycle the coefficient of useful work of the atomic power station using N204 at a maximum temperature of 500-670?C reaches avalue of 0.6-0.7, which is two or three times the value for atomic power stations using helium, and the efficiency of an atomic power station using N204 may be 28-35% (Fig. 4), whereas in a reactor using helium, for a regenerating scheme in which cooling takes place twice, the efficiency under comparable conditions is 20-31%. The coefficient of useful work of the cycle using N204 at a temperature of 50-550?C is 0.94-0.96, which is 1.9-2.3 times the value found for CO2 under analogous conditions (see Fig. 1). The thermodynamic efficiency of the con- densation cycles at maximum temperatures is 33-37% for the case of N204, whereas it lies between 26 and 35% in the case of CO2. In single-loop schemes of atomic power stations using a dissociating gas, it is possible to accept large values of pressure drop, attain relatively high values of specific thermal power density and have smaller dimensions for the heat-exchange devices. In single-loop schemes for atomic power stations using CO2, even when the cooling-water temperature at the inlet is 15?C, the heat-exchangers weigh 2.2 times as much as do analogous heat exchangers in an N204 reactor with a temperature of 20-22?C. In 1968-1970 the staff of theIYatANBSSR carried out design studies for atomic power stations using N204 gas-cooled fast reactors with a single-loop scheme for the station and a gas?liquid cycle with inter- mediate regeneration. The design studies showed that it was possible to construct a gas-cooled fast reactor with a thermal power density of 500-600 kW/liter, a doubling time of less than seven years, and a single- shaft gas turbine with a power of 1000 MW whose metal volume was less by a factor of 4.5 than in the case of steam. The cycle used was a gas?liquid cycle with intermediate regeneration at 20-23 abs. atm, with a maximum cycle pressure of 150 abs. atm and a gas temperature of 520-540?C. On the basis of the design studies made at the Institute, engineering and economic indicators were calculated for an atomic power station with an electrical rated power of 2000 MW. These calculations show that it is possible to achieve an electrical rated power of 2000 MW in single-loop atomic power stations, with two gas-cooled fast reactors rated at 1000 MW each, using N204 and having fairly high engineering and economic indicators. LITERATURE CITED 1. A. K. Krasin, Dissociating Gases as Coolants and Working Fluids in Power Plants, Proceedings of .the All-Union Conference [in Russian], Nauka i Tekhnika, Minsk (1970), p. 6; V. B. Nesterenko, ibid., p. 11 2. A. K. Krasin, V. B. Nesterenko, and N. M. Sinev, Proceedings of the Symposium "Mathematical simulation of thermal processes in power engineering" [in Russian], Nauka i Tekhnika, Minsk (1970), p. 95. 3. A. K. Krasin and V. B. Nesterenko, Thermodynamic and Transport Properties of Chemically Re- acting Gaseous Systems [in Russian], Nauka i Tekhnika, Minsk (1967). 4. P. M. Kovtun, A. N. Naumov, and S. A. Kosmatov, Byul. Izobr. i Toy. Zn., No. 21, 62 (1964). 5. V. B. Nesterenko, V. P. Bubnov, and A. M. Matyunin, Izv. Akad. Nauk BSSR. Ser. Fiz.-Tekh. Nauk, No. 1, 57 (1966). 6. M. A. Bazhin, V. P. Bubnov, V. B. Nesterenko, and N. M. Shiryaeva, Optimization of the Param- eters of Power Plants Using Dissociating Working Fluids [in Russian], Nauka i Tekhnika, Minsk (1970). 7. A. K. Krasin, V. B. Nesterenko, et al., Proceedings of the SEV Symposium on Atomic Power Sta- tions Using Fast Reactors [in Russian], Vol. 1, Obninsk (1967). 8. A. M. Sukhotin, N. Ya. Lantratova, and V. A. Gerasimova, Izv. Akad. Nauk BSSR, Ser. Fiz.- knerget. Nauk, No. 2, 56 (1968). 9. V. V. Khromov and A. M. Kuz'min, in: Physics of Fast Reactors [in Russian], Atomizdat, Moscow (1968), p. 92. 10. L. Lais et al., Atomnaya Tekhnika za Rubezhom, No. 1, 7 (1970). 11. Advanced and High-Temperature Gas Cooled Reactors, SM-111/12, IAEA, Vienna (1968). 230 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 MEASUREMENT OF REACTIVITY EFFECTS IN THE REACTORS OF THE BELOYARSK ATOMIC POWER STATION B. G. Dubovskii, A. Ya. Evseev, I. M. V. V. Korolev, V. F. Lyubchenko, Yu. I. Mityaev, and E. I. Snitko UDC 621.039.519 The reactivity of a reactor is usually calculated on the basis of information relating to the manner in which the neutron flux varies with time. In recent years, special rapid and highly-sensitive analog com- puters (reactimeters) have been developed [1-8] to accelerate the determination of the steady-state re- activity and analyze its transient and instantaneous values. Reactimeters may be divided into two types: open-cycle and closed-cycle, the latter incorporating a model of the reactor kinetics in the feedback circuit of the servo system. Reactimeters of the second type are used in the reactors of the Beloyarsk atomic power station. A spatially-independent, single-point, monoenergetic model of the reactor kinetics is realized (with- out allowing for the temperature coefficient of reactivity) by combining a multiplier, an operational am- plifier, and RC circuits designed to simulate the delayed neutrons. The monoenergetic model of the reactor is constructed with due allowance for the following reactor kinetic equations: 6 dtt pn dci s; dt To dt i?a deiatn at To in which we have employed the generally-accepted notation of reactor theory. Data relating to the ai and Ai of U235 subjected to fission by thermal neutrons are incorporated in the model of the reactor kinetics. The power-measuring system of the reactimeter is made in the form of a multiple-decade electro- metric operational amplifier, enabling ionization-chamber currents in the range 10-12-10-4 A to be mea- sured. The reactimeter has 14 scales, and measures positive reactivities in the range 10-5-5 ? 10-1 Peff and negative reactivities in the range 10-5-10 geff. The dynamic range of the reactimeter corresponds to approximately one decade of reactor power; the rapidity of action is -1 sec, and it is mainly determined by the inertia of the-noise- and interference- smoothing system and the inertia of the reading and recording device, which is an EPP-09 electronic po- tentiometer. An example of the recordings of the reactimeter is presented in Fig. 1. Figure 1 clearly demonstrates the high sensitivity of the instrument, even slight reactivity "noise" being readily perceptible. In power measurements it is easy to see the feedback associated with the nega- tive temperature and power coefficients of reactivity (see Fig. lb). The rate of change of reactivity due to these effects is -1.5 ? 10-7 Keff/ sec. Translated from Atomnaya Energiya, Vol. 32, No. 3, pp. 205-209, March, 1972. Original article submitted June 19, 1970. C 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without permission of the publisher. A copy of this article is available from the publisher for $15.00. 231 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Keff ? 105 10 Keff .? 105 a 20 -2 21 -20 90 120 150 MO t, sec 63 - 84 t, sec Fig. 1. Change in the reactivity of the first reactor of the Belo- yarsk atomic power station on displacing the automatic control rods through distances equivalent to 0.05% (a) and 10% (b) of the nominal power; c) effect due to reactivity "noise." The reactimeter was calibrated by comparing the results of measurements made by different methods for the same reactivity effects. The scatter in the resultant readings was no greater than ?5% of the mea- sured quantity on all scales of the reactimeter. Figure 2 illustrates a recording of reactivity taken during the removal of water from the steam- superheater channels, from which we readily see that the reactimeter records changes taking place in the reactivity outside the range of sensitivity of the automatic regulators. Study of Reactivity Effects in the Reactors of the Beloyarsk Atomic Power Station In starting the reactors of the Beloyarsk atomic power station, great attention was paid to an anal- ysis of the reactivity effects. The reactivities of the fuel channels, the control-rod channels, and the water in these were measured at various points of the active zone for full loads in the critical assemblies; the compens,.ting capacity and calibration curves of the control rods were measured, to estimate the reserve of reactivity, and the manner in which this varied with the amount of water in the fuel channels, to obtain relationships for the efficiencies of various absorbers, and also to plot the neutron distributions over the active zone of the reactor (by reference to the effectiveness of identical sections of the absorbing rods) and to correct calculated (design) data. The results of some of the measurements are presented in Tables 1 and 2. The results of the measurements carried out on the fuel-channel reactivity effects were used in connection with the recharging of the reactors in the Beloyarsk atomic power station. Using the reactimeter, we estimated the reactivity effect associated with the filling of the gaps in the graphite stack of the first reactor with water. For this purpose we used special fuel channels enclosed in thin-walled sheaths, which were filled with water. The reactivity effects were measured at various dis- tances from the center of the reactor and then extended to the whole active zone. We found that the total reactivity effect due to filling the gaps in the cells of all the IK-1.5 with water was equal to zero, while for the cells of all the IK-2 it equalled 0.94% Keff. The reserve of reactivity in the reactors of the Beloyarsk atomic power station was determined from the sum of the efficiencies of the compensating rods, measured by means of the reactimeter as each rod was extracted in turn. In these measurements, the compensating rods were arranged uniformly in the ac- tive zone with gaps of 60-70 cm. The neutron distribution corresponding to this arrangement will subse- quently be called the "uniform" distribution. 232 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Keff 105 20 10 120 -10 ?..1%J\s 360 .? ? A 420 980 540 t, sec -20L Fig. 2. Reactivity recording during the removal of water from the steam-superheater channels (A = instant at which the automatic reg- ulator operates). TABLE 1. Reactivity Effects during the Charging of One Evaporating Channel with 2% (IK-2) and 1.5% (IK-1.5) Uranium En- richment in the Reactor of the First Unit (10-5 Keff) Distance from cen- ter of re- actor, cm 10 30 70 190 190 230 270 310 390 IK-2 with water IK-1.5 with water 15,0 ?7,5 10,0 ?7,5 17,0 ?1,2 12,5 ?3,5 12,5 ?2,5 14,0 ?2,0 9,00 ?1,0 3,0 ?3,0 3,0 0 Further experiments showed that, for the uniform neutron distribution, the interference coefficient between a particular compensating rod and all the other rods (Kint) averaged over the reactor was almost equal to unity. The coefficient K1 allowing for the influence of a change in the neutron leakage from the reactor on the total efficiency of the rods was also close to unity. The relative increase taking place in the neutron flux in the cell of the extracted rod exerted the greatest influence. Direct measurements of the neutron distribution over the cell of the rod showed that this effect made the measured reserve of reactivity 6% higher than the true value. Thus the error in determining the reserve of reactivity from the sum of the efficiencies was ?10%. The reactivity effect associated with introducing perturbations into the reactor may be measured by reference to the corresponding change in the reserve of reactivity or to the kinetic characteristics of the changes in the neutron flux passing through the reactor. If the perturbation introduces no serious changes into the diffusion characteristics of the active zone or the leakage of neutrons from the latter, the reac- tivities determined by the two methods should coincide; otherwise substantial differences may occur. Let pi be the reactivity measured by reference to the change in the reserve of reactivity, p2 that measured by reference to the period of the change in reactor power (or by means of the reactivity meter), and p3 that associated with the change in the diffusion characteristics of the zone and the leakage of neutrons. Let us give some examples relating to the measurement of reactivity effects in the reactors of the Beloyarsk atomic power station The Critical Assembly. In the critical assembly with 212 fuel channels of the reactor, in the first unit, but without any water (one IK-2 to two IK-1.5), the efficiency of the IK-2 in one particular cell differed by roughly a factor of three for different positions of the compensating rods, this being mainly due to the great sensitivity of Keff to a change in the leakage of neutrons from the reactor (which is small in the radial direction) and to the interference of the rods. The Designed Charge. For the designed charge and a uniform neutron distribution, the sum of the efficiencies of the compensating rods E pi = 5.3 ? 0.5% Keff, while for a severely-deformed distribution, quite independently of which part of the distribution (central or peripheral) is protruding, ?E pi = 3.5 ? 0.4% Keff. For severe deformations of the radial field of the neutrons there is a considerable change in the neutron leakage, and Kint differs substantially from unity. In this case the reserve of reactivity is not equal to the sum of the efficiencies of the compensating rods, as confirmed by the foregoing data. Expelling Water. On expelling water from the fuel units of the reactor in the first section of the power station, the reactivity effect was p2 = ?0.65% Keff, while the reserve of reactivity changed by only +(0.1- 0.2)% Keff, corresponding to the calculated value. On expelling the water, in fact, the number of compen- sating rods introduced into the reactor fell by 12% (from 54 to 48), while their efficiency rose by 16%; the 233 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 0 60 sec Fig. 3. Effect of the spatial redistribution of the neu- trons on the initial readings of the reactimeter in re- lation to the distance r between the sensor of the re- actimeter and the displaced rod: a) r = 1.1 m; b) r = 2.9 m; c) r = 4.9 m. TABLE 2. Reactivity Effects in Various Fuel Channels of the Second Unit [9] in the Central and Peripheral Cells of the Active Zone (10-5 Keff) Effect measured In the center (cell 18-17) On the periphery (cell 19-34) Arrangement of IK-2 without water 2.6 ?0.9 Pouring water into the IK-2 3.2 1.5 Arrangement of IK-2 with water 5.9 0.7 Arrangement of I1-3 without water 31.8 8.3 Pouring water into the IK-3 13.0 4.7 Arrangement of IK-3 with water 44.8 13.0 Arrangement of steam-superheating 7.5 channel without water 29.0 7.5 Pouring water into the steam-superheating channel 8.8 2.2 Arrangement of steam-superheating channel with water 37.7 9.8 neutron distribution in both cases was almost uniform, and it was considered that Kint remained constant. The negative value of the reactivity effect was due to the increase in the efficiency of the compensating rods arising from the increase in the diffusion length in the water-free system. The Reactor of the Second Unit. In the reactor of the second unit, with absorbing rods in the steam- superheating channels, the pouring of water into the fuel channels caused hardly any difference in the re- serve of reactivity; the reactivity effects were: on pouring water into all the evaporator channels +1.3% Keff; on pouring water into the steam-superheating channels containing absorbing rods +0.45% Keff; on pouring water into the steam-superheating channels without absorbing rods +0.6% Keff. The mean efficiency of the IK-1.5 for the reactor of the first unit equalled-(1.5-2.0) ? 10-3% Keff. However, on placing the IK-1.5 into a cell adjacent to an immersed compensating rod, the effect equalled +12.10-3% Keff, this being due to the reduction in the absorption of neutrons in the rod as a result of screen- ing by the IK-1.5 channel. Spatial and Power Effects in Reactimeter Measurements. The single-point representation of reactor kinetics is satisfied the less closely, the greater the dimensions of the reactor, and the closer to the re- actor surface the reactimeter detector is placed. 234 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 IC(A) Reactimete SU IC(W) I SU ACS L IRM RC Technological-parameter recorder 1ICon TU SU R(h) R(p) R(N) 5,,-5 I fi5 Fig. 4. Block diagram of the set of reactor-monitoring instruments: IC(A)) actuating ionization chamber; IC(W)) working ionization chamber; SU) supply unit; ACS) automatic chamber switch; IRM) instantaneous reactivity meter; RC) RC circuits of the monoenergetic model of the reactor kinetics; D(h)) detector (sensor) for the position of the automatic control rods; K) switch (key); D) detector (sensor) of technological parameters; ICon) information converter (converting the information into a form suitable for recording); R) recorder; TU) timing unit (generator of timing pulses); R(h)) recording unit for the position of the automatic control rods; R(p)) reactivity recorder; R(N)) power re- corder; P) parallel instruments on the control desk. The spatial distribution assumed by the neutrons in the reactor on introducing a perturbation of reac- tivity is described by a superposition of ground and higher harmonics, which add up in such a way that in the zone of perturbation the sign of the deformation in neutron flux follows the sign of the perturbation in reac- tivity, while in the diametrally opposite zone the signs of the deformation and perturbation oppose one another. For a reactimeter incorporating a single-point model of the reactor kinetics, a change in the neu- tron density will be perceived as an instantaneous distortion of the initial conditions. Hence there may be an error in the determination of the instantaneous values of the reactivity: if the sensor of the reactimeter is situated close to the zone of perturbation, then in the initial period the readings of the reactimeter will be too high; if the sensor is situated on the opposite side of the reactor from the perturbation, then the readings will be too low (Fig. 3). In order to eliminate these errors, a time delay is required (after the perturbation), during which the neutrons may be redistributed and the higher harmonics attenuated, while the RC circuits in the reactor- kinetics simulator may be recharged. Hence when making measurements with the reactimeter it is essential to create conditions conducive to a single-point representation of the reactor, for example, by connecting several sensors (distributed around the active zone) in parallel to the reactimeter, or by arranging the sensor a fair distance from the active zone. For measurements in the power mode, the reactimeter will measure the total change in reactivity, i.e., that due to the effect being studied (for example, the displacement of a rod) plus that due to the asso- ciated power, temperature, or other reactivity effects. The desired effect may be separated out from the others by making additional measurements, for example, by subtracting the reactivity effect associated with the motion of the compensating rods, which may be obtained from the calibration curve or from preliminary measurements of the effect under consideration at zero power. 235 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 It is also possible to design a circuit enabling the reactivity introduced by the rods to be evaluated automatically, and separating the fast and slow reactivity components [7]. Set of Reactor-Monitoring Instruments Experience in starting the reactors of the Beloyarsk atomic power station showed that, in order to monitor the operation of these, it was desirable to use a set of monitoring instruments enabling the changes taking place in the reactivity, the power of the reactor, and the position of the automatic control rods to be measured and recorded all at the same time. The recording of these parameters together, with those of the temperature and flow of the coolant, the temperature of the fuel elements, and other technological param- eters (Fig. 4) may be used in order to study the transient modes and dynamic characteristics of the reac- tor, to monitor the absence of nuclear hazard, to verify the operative control of the technological process and the operation of the regulating mechanisms, and to analyze possible emergency situations. The reactimeter constitutes a fundamental part of the set of reactor-monitoring instruments. The reactimeter is used to measure changes in reactivity lying outside the range of sensitivity of the automatic control system. The reactimeter may also be used for measuring the effects of temperature and power level on the reactivity. The set of instruments is particularly useful for those reactors which are not furnished with infor- mation and computing facilities capable of executing continuous measurements or calculations and record- ing the parameters in question. This set of instruments was used in bringing the second unit of the Belo- yarsk power station up to full power, and a similar set has been incorporated in the standard equipment of the reactors in the Bilibinsk and Shevchenko atomic power stations. LITERATURE CITED 1. G. Stubbs, ME Trans. NS-4, No. 1, 14 (1957). 2. C. Sastre, Nucl. Sci. Engng., 8, No. 5, 37 (1960). 3. P. Bonnaure, L'onde Electrique, No. 377-8 (1958). 4. P. Shea, IRE, Intern. Convent., Rec., 9, No. 4, 53 (1962). 5. E. Suzuki and T. Tsunoda, J. Nucl. ScE Techn., 1, No. 6, 210 (1964). 6. R. Shomo, S. Hamilton, and R. Learner, Trans. ANS, 9, No. 1 (1966). 7. I. I. Sid2rova, Analog Simulation in Nuclear Power [in Russian], Atomizdat, Moscow (1969). 8. B. G. Dubovskii et al., "Measurement of reactivity in the reactor of the Beloyarsk atomic power station," Contribution to the Symposium on the Physics and Technology of Thermal-Neutron Reactors [in Russian], London (June 27-29, 1967). 9. I. S. Akimov et al., At. Energ., 28, 321 (1970). 236 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 DETERMINATION OF CONTENT AND SPATIAL DISTRIBUTION OF URANIUM IN FLUORITES FROM TRACKS OF FISSION FRAGMENTS OF URANIUM I. G. Berzina, I. V. Mel,nikov, UDC 549.755.35 and D. P. Popenko Fluorites are common minerals occurring in deposits of various types. They contain impurity elements, one of which is uranium. The uranium content in fluorite crystals has been determined earlier [1-4] by chemical and lumines- cence methods which have a low lower sensitivity limit and permit only a determination of the gross content of uranium in the sample without elucidating its spatial distribution characteristics. In the present article we present the results of determination of concentrations and spatial distri- bution of uranium in fluorite crystals by the method of f-radiography [5]. -This method enables one to de- termine the amount of uranium in the form of isomorphic impurity and the nature of its distribution in fluorite crystals and the enclosing rock. The possibility of using this information as a prospecting cri- terion for discovery of ores, which contain uranium is demonstrated. Crystals and slices of natural fluorite and enclosing rocks of fluorite, tin ore, and uranium?molyb- denum deposits were investigated. The sensitivity of the method used is 10-12 g/g and the accuracy of the determination is 10-15% [6]. The method is based on the possibility of detecting tracks of fission fragments of uranium nuclei in the minerals capable of spontaneous fission or fission under the action of neutrons [5]. A crystal, in which the fission of uranium nuclei has occurred, or a detector that has recorded mul- tiply charged ions from the surface of the crystal, retains a "memory" of the fission of uranium nuclei and shows the location of these nuclei. The tracks of the fission fragments are revealed with thp use of selec- tive chemical etching of the investigated surface. It is well known [7-9] that a large number of dislocations are revealed on the surface of fluorite crystals on chemical etching; the density of these dislocations is 103-106 cm-2. The etch figures on the dislocations and the tracks of the fission tracks on the shear plane of fluorites are difficult to distinguish: in both cases they have the shape of pyramids with bases that vary depending on the type of etching. How- ever, when the density of the dislocations does not exceed 5 ? 104 cm-2, tracks of the fission fragments of uranium can also show up along with the dislocations. To verify this multiple etching of the mineral was done, as a result of which it was possible to distinguish between the etch figures on the dislocations and the tracks of fission fragments [10]. The most significant difference between the tracks and the dislocations is that a track can end at any point of the crystal and its length is determined only by the specific energy losses of the fission fragment in the crystal lattice. A dislocation can not terminate inside the crystal. It must come out either on the free surface or on the surface of an extended defect, or close on itself. In the process of prolonged etching the etch figure, forming on a dislocation, will become deeper and wider, retaining the shape of a pyramid. With the etching of a defective region the etch figure on a track becomes plane-bottomed and takes the shape of a truncated pyramid. Thus a track can be distinguished from a dislocation with double etching in the ap- propriate regime. The scheme of etching and the microphotograph of one of the samples, subjected to etching, are shown in Figs. 1 and 2 by way of example. Translated from Atomnaya Nergiya, Vol. 32, No. 3, pp. 211-215, March, 1972. Original article submitted April 19. 1971. 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without permission of the publisher. A copy of this article is available from the publisher for $15.00. ' 237 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 z II Fig. 1 Fig. 2 Fig. 1. Scheme of formation of etching holes on dislocation (D) and track (T). I and II) Shape of the etch figures inthe body and at the surface of the sample respectively. Fig. 2. Etch figures on tracks and dislocations at the shear surface of a natural fluorite. Etching in 3 N HCI at 20?C for 40 min. The experiments on the identifications of etch figures on tracks and dislocations enabled us to detect tracks of spontaneous fission of uranium in natural fluorite crystals. This inturn facilitated discrimination of uranium which entered the crystal lattice of the fluorite during its formation and uranium penetrating into the crystal along fissures in a subsequent process, if any. For the latter case a nonuniform distribution of uranium from the periphery to the center is typical. Colorless, green, and violet fluorites were investigated. The density of tracks of fragments of spon- taneous fission of uranium did not exceed n ? 104 cm-2 (n = 1, 2, 3) and was uniform in uniformly tinted fluo- rites. The last fact indicates that in these crystals uranium entered into the fluorite lattice during its crystallization and the uranium content in the fluorite is extremely low (C 10-6-10-7 wt. %). It should be mentioned that in fluorites from uranium?molybdenum deposits a nonuniform distribu- tion of uranium is observed along the zones of growth of the crystal, accompanied by a sharp increase in the density of dislocations in the parts adjacent to these zones (Fig. 3). The increase in the density of dislocations in these parts does not permit determination of the distribution of uranium directly in the crystal itself. In order to reveal the characteristics of the spatial distribution of uranium and to determine its concentration in these cases one should use the records of tracks of forced fission of uranium by external detectors. For this purpose a detector is annexed to the investigated surface of the crystal; the detector records fission fragments escaping from the crystal during its bombardment by neutrons in a nuclear reactor. After irradiation the detector is subjected to chemical etching, as a result of which the tracks of the fission fragments can be seen under an optical microscope; they are observed in the form of black points and dashes on a clear background. The concentration of uranium C is computed on the basis of the following reasoning. The density of tracks p, recorded by the detector, is proportional to the concentration of uranium in the crystal and to the integral flux of neutrons. The neutron flux is determined with the use of especially prepared standard tar- gets having mass mi. The thickness of the target is such that fission fragments escaping from it reach the detector placed in contact with the target during its irradiation [5]. The concentration of uranium in the material is computed from the formula [6] C? 2 104 P 7 'VA Pi where X is the average atomic weight of the subject material (for fluorite X = 26.8 amu); pi is the density of tracks in the detector adjoining the target. In order to compute p curves of statistical distribution of the density of tracks over the sample are constructed. The form of these curves serves as the basis for the determination of the forms of occurrence 238 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Fig. 3 Fig. 4 Fig. 3. Dislocation bands along the zones of growth of the crystal, containing increased amount of impurities. Fig. 4. Microphotograph of a thin section of fluorite with uraninite deposited along the faces of growth of fluorite (a), and of the detector (b) recording the tracks of fragments of forced fission of uranium. Detector: Lavsan film with thickness of 12 ?; etching regime: 40% KOH, T = 20 min, t = 60?C. TABLE 1. Uranium Concentration (-10-5 of uranium in the crystal. Revealed in this process are wt. %) in Fluorites from Different Seg- uranium distributed uniformly along the crystal, uranium ments of the Vein* of Fluorite Deposit deposited along the growth faces, and uranium timed with Color of fluorite Number of section 1 2 3 4 5 Porcelain type Colorless Light green Green Violet 1,2 0,4 0,6 0,2 0,3 0,9 0,2 1,2 0,3 0,3 ? 0,3 0,4 { 0,3 {?'3 0 0 ,2 ,4 0,3 0,9 0,4 *Uranium concentration in enclosing rocks is equal to 13.6 ? 10-5 wt.% microinclusions of the rock or minerals [11]. The concentration and spatial distribution of uranium in fluorites from deposits of different types were deter- mined. Since small crystals (-1 cm2) were used, it was necessary to determine the stability of the distribution along the course and depth of the fluorite vein. For this purpose crystals from five segments of the fluorite vein with overall distance of ?190 m from the first to the last segment were used. The results of the analysis showed that the uranium concentrations in fluorites of the same color depend prac- tically on their location (Table 1). The stability of the uranium content in fluorites of a given color made it possible to reduce the num- ber of analyzed crystals considerably. The uranium content in fluorites taken from different deposits vary within an order of magnitude (Table 2), but the nature of the distribution of uranium is different. Thus, for example, in crystals from fluorite and tin-ore deposits the distribution of uranium is uniform, just as in porcelain type (white) fluorite from a fluorite deposit, which is distinguished from the remaining fluorites of this deposit by large uranium content. As mentioned above, in the case of a nonuniform distribution of the content of uranium deposited along the growth zones of fluorites an increased density of dislocations is observed, which makes it difficult to reveal the tracks of fission fragments directly on the crystal. In such segments of the crystals an increased concentration of uranium (-10-3 wt. %) in zones with thickness of a few tens of microns could be detected with the use of an external detector (Fig. 4). The de- position of uranium along the growth zones of fluorite crystals in the form of extremely thin crusts of in- dependent uranium minerals is apparently accompanied by recrystallization; as a result stresses develop in closed volumes, since the growth of the crystals does not stop after the deposit of the uranium minerals. 239 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Fig. 5. Microphotographs of polished section of fluorite with inclusions of rock (a) and of the corresponding de- tectors: Lavsan detector (b) and photo- graphic plates for nuclear investi- gations (c). TABLE 2. Uranium Concentration (-10-5 wt. %) in Crystals of Fluo- rites and in Enclosing Rocks from Different Deposits Characteristic of sample Type of deposit fluorite tin-ore uranium?molyb- denum Violet fluorite 0.3 0.8 1.4-3.6 Green fluorite 0.9 1.5 1.5 Light green fluorite 0.4 0.2 0.8 -3.3 Colorless fluorite 0.3 3.8 Enclosing rock 13.6 4.3 103.5 As a rule, 20-100 p wide violet bands are observed on both sides of the zone along which the crust for- mation occurs; the intensity of their tint for single-age fluorites depends on the amount of deposited ura- nium. In this case the violet color of fluorite can be explained by the effect of the natural radioactivity of uranium, causing the formation of definite tint centers. This is indicated by the fact that the tint of the fluorite at these points disappears on heating. In fluorites taken from uranium?molybdenum deposits an increased uranium content was noticed in the entire violet zone of the irradiated crystals. Due to the small width of the zone uranium cannot be separated from the mineral in its sampling in the corresponding analysis. Therefore, in gross uranium analysis an increased value of the concentration of uranium can be obtained in polycrystalline fluorite. Apparently an increased uranium content in fluorite is often obtained due to the presence of minute inclusions of the rock (Fig. 5). It should be noted that the blackening of the plate for nuclear investigations occurs only in contact with the fluorite, whereas the parts of the plate, that adjoin the chips of the rock, remain unilluminated, even though the uranium content in the rock is higher by an order of magnitude. Thus the blackening of the photographic plates cannot be an unambiguous indicator of radioactivity of violet Iluorites as assumed earlier [3]. We observed nonuniform distribution of uranium only in crystals taken from veins of uranium?molyb- denum deposits. It was found from mineragraphic investigations that the growth zones of fluorite crystals are enriched by a uraninite (Fig. 6); the uranium content of the fluorite in this case is ?10-3 wt. %. Minute depositions of uranium occur also in the enclosing rock. The maximum concentration of uniformly distri- buted uranium in such fluorite crystals (see Table 2) is 10-5 wt. %. The increased uranium content in the growth zones of fluorite crystals can be explained in the fol- lowing way. Fluorite veins with increased uranium content, coming out to the surface of the earth above 240 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Fig. 6. Microphotographs of a section containing fluo- rite (a) and of Lavsan detector (b) recording uranium pressed back into the peripheral zones of the fluorite crystal (uraninite band is observed at the base of the crystal). hidden ores, could be formed from hydrothermal solutions washing in their path the ore materials which would enrich the solutions with uranium. Since the solubility of uranium in fluorite is finite, the surplus uranium in the solution must be deposited along the growth zones of the crystals and around fully grown crystals in the form of minute bands of uraninite. The nature of distribution of uranium and its concentration in rocks enclosing fluorite veins are also different. Tin-ore and fluorite deposits are characterized by low uranium content. In these cases the uranium is timed to definite accessory minerals, characteristic for the given rock, and no redistribution of uranium occurs. A different pattern is noticed near uranium?molybdenum ore materials, where the uranium content is two-three times higher in the enclosing rocks than in fluorite crystals. This is apparently explained by the deposit of uranium from solutions enriched by this element. Besides, in the rocks uranium is distributed nonuniformly: a large number of fine grains of uranium minerals and formations, which appear on the sur- face of the detector in the form of star-shaped segregations and cannot be diagnosticated, are observed. Thus the proximity of the ore material, through which the hydrothermal solutions pass, affects the forms in which uranium appears in the rocks enclosing the fluorite veins. The following conclusions can be made on the basis of these investigations. As a rule, in the lat- tice of fluorite crystals uranium is distributed uniformly in the form of impurity. The content of this ura- nium does not exceed n ? 10-5 wt. %. During the growth of the crystals the uranium "surplus" is deposited along the growth zones or is pushed back to the peripheral parts of the crystals in the form of independent uranium minerals. A nonuniform distribution is characteristic of rocks enclosing fluorites in the case of uranium surplus. The presence of surplus uranium, deposited along the growth zones of fluorite, and an increased uranium content in the crystal lattice can be used as a prospecting criterion for the detection of ores with high uranium content. LITERATURE CITED 1. N. N. Vasil,kova and S. G. Solomldna, Typomorphic Characteristics of Fluorite and Quartz [in Russian], Nedra, Moscow (1965). 2. F. Daniels, Ch. Boyd, and D. Saunders, Uspekhi Fiz. Nauk, 51, 271 (1953). 3. K. Pshibram, Tint and Luminescence of Minerals [Russian translation], IL, Moscow (1959). 4. H. Haberlandt, in: Rare Elements in Volcanic Mountain Rocks and Minerals [Russian translation], IL, Moscow (1952), pp. 17740. 5. I. G. Berzina et al., At. Energ., 23, 520 (1967). 6. I. G. Berzina et al., Izv. Akad. Nauk SSSR, Ser. Geol., 8, 70 (1969). 7. I. G. Guseva and A. A. Urusovskaya, Kristallografiya, 6-, 432 (1964). 8. W. Kleber, M. Hahnert, and L. Ludke, Kristall und Technik, 1, 585 (1966). 9. A. Patel and C. Desai, Z. Kristallogr., 123, No. 1, 51 (1966).- 10. I. G. Berzina and I. B. Berman, Dokl. Akad. Nauk SSSR, 174, 3, 553 (1967). 11. I. G. Berzina, B. G. Lutts, and A. P. Akimov, Izv. Akad. Nauk SSSR, Ser. Geol., 1, 14 (1967). 241 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 THE PENETRATION OF BEAMS OF HIGH-ENERGY PARTICLES THROUGH THICK LAYERS OF MATERIAL V. S. Barashenkov, N. M. Sobolevskii, UDC 539.12.17 and V. D. Toneev A procedure for calculating a nucleon?meson cascade in a slab of material was described in [1] and the distributions of particle fluxes inside and outside slabs of various compositions and thickness initiated by high-energy primary radiation were investigated also. Using the same procedure we study more de- tailed characteristics of secondary particle fluxes beyond a a thick shield. Calculations simulating the fate of each par- ticle in the material are performed by the Monte Carlo meth- od. Each inelastic interaction of a particle with a nucleus 104. 40 50r, cm Fig. 1. Spatial distribution of a) neutron fluxes and b) proton fluxes beyond an alu- minum shield of thickness z bombarded with 660 MeV protons. Histograms are calculated; experimental points are taken from [7, 8]; (4), z = 75 g/cm2; ? ? ?(0), z = 150 g/cm2; r is the distance from the axis of the beam. is treated by the Monte Carlo method using the cascade-eva- poration model [2-4]. The behavior of neutrons with energies E < 10.5 MeV is simulated on the basis of reactor constants [5]. For comparison we used the results of measurements performed on the OIYaI synchrocyclotron for energies T = 340 and 660 MeV [6-8].* In the calculations the geometry of the experiments was reproduced exactly: aluminum slabs having thicknesses z of 75 and 150 g/cm2 were bombarded with a collimated beam of protons approximately 1 cm in radius. Since the exact shape of the beam was simulated by using a matrix determined experimentally, it was not neces- sary to introduce any geometric corrections. The calculated results, plotted in Fig. 1, show that the neutron and proton fluxes fall off almost exponentially with the distance from the axis of the beam. The calculation gives *Some preliminary results obtained within the framework of a simplified model were reported at a conference on dosimetry and shielding physics at Dubna in 1969 [9]. TABLE 1. Fluxes of Secondary Particles Beyond an Aluminum Shield of Thickness z Bombarded by 660 MeV Protons Particles, energy Neutrons,T >50 MeV Neutrons,,F< 30 MeV All protons z = '15 gicmz z = 150 g/cm2 ? experiment theory experiment theory 0,50+0,15 0,34+-0,01 0,29+0,11 0,33+0,02 1,05+0,32 0,65+0,05 0,76+0,22 1,14+0,13 0,82+0.15 0,69+0,01 0,40+0,07 0,39+0,01 Translated from Atomnaya Energiya, Vol. 32, No. 3, pp. 217-221, March, 1972. Original article submitted June 24, 1971. 242 0 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without permission of the publisher. A copy of this article is available from the publisher for $15.00. Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 TABLE 2. Energy Spectra of Protons (%) beyond a Shield of Thickness 75 g/ cm2 at a Distance r from the Axis of a 660 MeV Primary Beam E, MeV r = 2 cm r = 16 cm r = 29 cm experiment theory experiment theory experiment theory 0-50 -1,0 1,7 6,0 7,2 19,2 15,6 50:100. 1.3 2,7 9,1 13,4 20,4 28,1 100-150 0,7 2,4 12,1 24,1 16,8 28,1 150-200 4,6 2,1 22,0 17,0 19,7 15,6 200-250 2,3 1,6 6,1 12,5 7,1 9.5 250-300 5,1 1,2 6.1 15,1 7,3 3.1 300-350 5,2 1,5 6,8 6,2 2,9 350-400 4.9 1,1 5,3 3,6 1,5 >400 74,9 85,7 26,5 0,9 5,1 Total number of 307 2 604 132 206 ' 137 32 events TABLE 3. Energy Spectra of Protons (%) beyond a Shield of Thickness 150 g/ cm2 at a Distance r from the Axis of a 660 MeV Primary Beam E, MeV r = 2 cm r = 16 cm r = 29 cm experiment theory experiment theory experiment theory 0--50 1,0 0,7 (0,7) 8,8 8,7 4,3 9,1 50-100 1,4 1,2 (1,2) 10,0 17,5 12,8 27,4 100-150 1,6 1,3 (1,3) 16,5 24,5 28,2 32,8 150--200 9,7 1,7 (1,7) 16,5 18,0 21,3 25,5 200--250 12,3 2,2 (2,2) 8,4 21,8 5,8 5,3 250-300 21,4 12,0 (3,0) 11,5 6,8 7,4 300--350 23,6 72,0 (89,7) 11,5 1,6 6,4 350-400 13,6 8,7 (0,1) 8,0 0,6 3,7 > 400 15,4 0,2 (0,1) 8,8 0,5 10,1 Total number of 281 1 215 272 183 188 55 events Note: The values in parentheses are calculated by neglecting multiple scattering and the energy spread of the primary beam. TABLE 4. Energy Spectra of Protons (%) beyond a Shield of Thickness 75 g/cm2 at a Distance r from the Axis of a 340 MeV Primary Beam E, MeV r = cm r = 16 cm r = 29 cm experiment theory experiment theory experiment theory 0-50 20,7 12,2 (6,3) 65,8 37,5 32,5 17 50-100 56,0 64,3 (93,2) 24,4 25,0 27.5 17 100-150 18,1 20,8 (0,2) 7,8 12,5 27,5 49 150-20(1. 4,8 2,5 (0,2) 1,0 25,0 2.5 17 200-250 0 0,2 (0.1) 1,0 2,5 250-300 0,4 0 (0) - - 7,5 Total number of events 229 3 786 193 8 40 6 Note: The values in parentheses are calculated by neglecting multiple scattering and the energy spread of the primary beam. the correct character of the spatial distribution of radiation beyond the shield and is in satisfactory quan- titative agreement with experiment. This agreement is somewhat less satisfactory for fluxes close to the axis of the primary beam (r 2 cm). It should be noted that the agreement between the calculated and measured proton fluxes results not only from the exact account of the beam geometry but also from the multiple scattering correction for those protons which passed clear through the slab without a nuclear interaction. Such "uncollided" protons make up the main part of the particle beam close to its axis but Coulomb scattering effectively increases its size. Thus, for example, in spite of the fact that the radius of the beam is only 1 cm almost 30% of the protons emergingfrom a target of thickness z = 150 g/cm2 at a distance of 5 cm from the axis of the primary beam 243 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 TABLE 5. Spectra of Protons (%) with Energies E > 10 MeV in a Given Range of Polar Angle beyond a Shield of Thickness 75 g/cm2 for 660 MeV Primary Protons E, MeV 0-30? 30--60? 60_900 0--90? experiment I theory experiment I theory experiment I theory experiment I theory 0-50 0,4 0,9 25,2 10,3 60,0 26,5 2,4 2,0 50-100 2,0 1,5 33,3 15,2 20,3 42,9 3,1 3,3 100-150 1,6 1,5 20,2 18,3 15,6 12,2 1,8 3,3 150-200 4,9 1,5 9,8 15,9 3,1 14,3 5,3 3,1 200-250 2,3 1,3 3,7 15,7 0,4 4,1 2,4 2,8 250-300 5,2 1,7 3,2 12,2 0,5 - 5,3 2,7 300-350 5,3 2,6 1,6 7,9 0,1 - 5,0 3,1 350-400 5,8 2,3 0,5 4,5 - - 4,7 2,5 >400 72,5 86,7 2,5 - - - 70,0 77,2 Total number of events - 7 437 - 732 - 77 824 TABLE 6. Spectra of Protons (%) with Energies E > 10 MeV in a Given Range of Polar Angle beyond a Shield of Thickness 150 g/cm2 for 660 MeV Primary Protons E, MeV ,0--300 30--60? 60--90? 0_900 experiment theory experiment theory experiment theory experiment theory 0-50 0,9 0,6 (0,6) 28,8 14,7 74,0 38,5 7,7 2,2 50-100 2,7 2,2 (2,2) 26,7 20,5 18,1 38.5 9,3 4,2 100-150 4,8 3,0 (3,0) 29,5 26,3 5,3 19,2 15,3 5,3 150-200 12,3 4,4 (4,4) 10,5 20,0 1,2 3,8 15,6 5,9 200-250 11,9 5,4 (5,4) 1,8 14,0 0,5 ___ 8,8 6,2 230-300 19,7 12,8 (6,3) 0,8 3,8 0,912,6 11,9 300-350 21,3 65,2 (77,9) 0,8 0,4 - _ 12,7 58,6 350-400 12,4 6,1 (0,1) 1,1 0 8,3 5,5 >400 14,0 0,2 (0,1) - 0,3 - 9,7 0,2 Total number of events 2766- - 293 26 - 3085 _Note; Values in parentheses were calculated by neglecting multiple scattering and the energy spread of the primary beam. for T = 660 MeV are particles which have passed through the slab without nuclear collisions. At a distance of 2 cm the fraction of such protons increases to 50%. The calculated values of the total radiation fluxes beyond the shield are shown in Table 1. The importance of the contribution of uncollided particles for small values of r can be clearly traced in the energy spectra (Tables 2-4). Even a small uncertainty in reproducing the experimental conditions has an appreciable effect on the calculated results. Thus if it is assumed that the primary beam is mono- chromatic a sharp maximum appears in the calculated spectrum (cf. numbers in parentheses in Tables 3 and 4). If it is assumed that the incident protons have a certain energy spread, e.g., 10 MeV for T = 660 MeV and 25 MeV for T = 340 MeV, the tables show that the maximum is spread out and the spectral shapes approach the experimental. In this case the energy spread taken into account is not that of the whole primary beam but only of those protons which have penetrated the whole slab without a nuclear interaction. It is clear that better agreement with experiment can be achieved by varying the dispersion and taking ac- count of the spread for the whole beam. Thus in the present cases (z = 75 g/cm2, T = 340 MeV, and z = 150 g/cm2, T = 660 MeV) the experimental conditions are known almost exactly. On the other hand for z = 75 g/cm2 and T = 660 MeV the energy loss in the slab is small and the effect of the energy spread of the prima- ry beam is beyond the limits of measurement of the spectrum. In this case the calculated spectra are in good agreement with the experimental (Table 2). All the above continues to hold in the analysis of the correlations between the angle of emergence and the energy of the secondary particles (Tables 5-7). Even here where the effect of the energy spread in the primary beam on the shape of the measured spectra is unimportant, agreement with experiment is adequate (Table 5). Both the correlations between the angle of emission and the energy of the particles and the spatial-angular distribution of charged particles beyond the shield are correctly given (Table 8). 244 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 TABLE 7. Spectra of Protons (%) with Energies E > 10 MeV in a Given Range of Polar Angle beyond a Shield of Thickness 75 g/cm2 for 340 MeV Primary Protons E, MeV O-30. 30_600 6O-9O? 0-90? experiment theory experiment theory experiment theory experiment theory 0-50 16,5 11,7 62 66,3 50,5 82 20,4 12,5 50-100 46,2 67,3 36 23,6 48,5 0 46,0 66,6 100-150 30,9 18,7 1,2 4,5 1,0 18 27,5 18,5 150-200 5,3 2,28 0,8 5,6 -- -- 4,7 2,32 200-250 0,6 0,02 -- -- -- -- 1,0 0,06 250-300 0,5 -- -- -- -_ 0,4 0,02 Total number of events -- 6935 79 -- 11 -- 7025 TABLE 8. Angular Distribution of Protons (%) Emerging from a Shield of Thickness 75 g/cm2 at Various Distances from the Axis of a 660 MeV Primary Beam Angle, deg r = 2 cm r = 16 cm r = 29 cm Total value (over all r) experiment theory experiment theory experiment theory experiment theory 0-15 15--30 30--45 45--60 60--75 75--90 84,6 10,6 3,3 0,7 0,6 0,2 87,0 7,1 3,0 2,0 0,8 0,1 9,1 49,2 22,7 12,1 6,1 0,8 0 17,8 58,0 18,7 5,5 0 8,8 27,0 35,7 21,9 5,1 1,5 0 15,4 53,9 23,0 7,7 0 . 76,9 12,2 6,3 2,9 1,3 0,4 77,7 12,3 6,3 2,8 0,7 0,2 Total number of events 2664 206 32 8246 Thus our method of calculating the penetration of high-energy particles through a slab of material by simulating pion- and nucleon-nuclear interactions with the Monte Carlo method can be used success- fully to calculate integral characteristics such as the total particle flux, and to obtain detailed information on the spatial structure of radiation beyond the shield, various spectral-angular characteristics, their correlations, etc. A comparison of the calculated and experimental results shows that at the present time the accuracy of the calculations depends mainly on a sufficiently complete knowledge of the experimental conditions. Some uncertainty in the values of individual parameters of the model is not very important. LITERATURE CITED 1. V. S. Barashenkov, N. M. Sobolevskii, and V. D. Toneev, Atomnaya Energiya, 32, 123 (1972). 2. V. S. Barashenkov, K. K. Gudima, and V. D. Toneev, OIYaI P2-4065, P2-4066, Dubna (1968). 3. V. S. Barashenkov, A. S. Thinov, and V. D. Toneev, OIYaI P2-5282, Dubna (1970). 4. V. S. Barashenkov et al., OIYaI P2-5507, P2-5549, Dubna (1970). 5. L. P. Abagyan et al., Group Constants for Nuclear Reactor Calculations [in Russian], Atomizdat, Moscow (1964). 6. V. E. Dudkin et al. Atomnaya Energiya, 22, 491 (1967). 7. V. E. Dudkin et al. Atomnaya 1nergiya, 23, 241 (1967). 8. A. I. Vikhrov et al. in: Dosimetry and Radiation Shielding Problems [in Russian], Vol. 11, Atomiz- dat, Moscow (1970) p. 70. 9. V. E. Dudkin et al., in: Dosimetry and Shielding Physics for Accelerators [in Russian], OIYaI 16- 4888, Dubna (1970), p. 75. 245 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 ABSTRACTS NEUTRON SPECTRA IN NOVO-VORONEZH ATOMIC POWER STATION REACTORS S. S. Lomakih, G. G. Panfilov, V. I. Petrov, L. I. Golubev, V. P. Kruglov, and V. A. Vikin UDC 621.039.524:621.039.51 During the start-up of the second reactor of the Novo-Voronezh atomic power plant the spectral parameters, the thermal neutron flux density distribution, and the absolute values of the neutron flux density over the height of the reactor core were measured with activation detectors. The experiments were performed at various coolant temperatures and reactor powers in instrument channels passing into fuel assemblies with 1.5 and 2% enrichment in U235. The spectral indices A were measured with 3 mm diameter detectors made of lutecium and europium metal on an aluminum base. Copper foils 50 /..( thick were used as 1/v detectors. Absolute thermal neutron flux densities were determined with gold foils 0.1 i thick. Copper detectors 50 /u thick and gold foils 10 tt thick, all 2 mm in diameter, were irradiated in 0.5 mm thick cadmium covers to separate the thermal and epithermal components of the neutron spectrum. The average values of Rcd, A, and Tn measured in the instrument channels are listed in Table 1. Analysis of the data obtained in these experiments and those on the first reactor [1] shows that the variations of such parameters as the spectral indices and the cadmium ratios over the height of the reactor core are of much the same nature for all measurements. The ratio of thermal to resonance neutrons in the central part of a fuel assembly (region 30-200 cm) is constant. The reflector causes an increase in the cadmium ratios at the edges of an assembly. The spectral index for lutecium over an assembly increases somewhat toward the top of the core. The neutron spectra for the first and second Novo-Voronezh atomic power plant reactors vary con- siderably over the assemblies. TABLE 1. Spectral Characteristics of Neutrons in Second Novo-Voronezh Atomic Power Station Reactor Channels Tempera- ture of coolant, ?C Uraniumen- richment, %u23' _ Rn RCu Cd ALu Cu neutrons/cm2. sec 30 108 218 260 1,5 2 1,5 1,5 1,5 2 1,75+0,05 1,57+0,05 1,69+0,05 1,65+0,05 1,54+0,05 1,51+0,05 6,50?0,19 5,33?0,16 6,10?0,18 5,70+0,17 5,14+0,15 4,31+0,13 1,00+0,03 1,01+0,03 1,15+0,03 1,37+0,04 1,37+0,04 1,42+0,04 375+19 396+20 462+23 607+30 642+32 713+35 (9,85+0,98)?10" (8,03?-0,80)-10" (7,18+0,72)?10" (7,32+0,73)?10" (2,88+0,29)?10" (2,36?0,24)?10" The absolute values of the thermal neutron flux density were measured at 129 cm from the bottom of the reac- tor core. Translated from Atomnaya nergiya, Vol. 32, No. 3, p. 223, March, 1972. Original article sub- mitted April 13, 1971; abstract submitted October 29, 1971. 246 O 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without permission of the publisher. A copy of this article is available from the publisher for $15.00. Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 LITERATURE CITED 1. S. S. Lomakin et al., Atomnaya Energiya, 30, 301 (1971). VARIATIONS IN COMPOSITION IN RADIOISOTOPE X-RAY FLUORESCENCE ANALYSIS OF POLYMETALLIC ORES Yu. P. Betin and I. A. Krampit UDC 543.09 Methods for coping with the effect of changes in the chemical composition of samples on the results of radioisotope x-ray fluorescence analysis of complex ores are discussed. The formulas for the intensities of the characteristic x-radiation of the element to be determined, and the backscattered radiation, in forms convenient for numerical calculations, are derived. Formulas are derived for the case of a semiinfinite flat surface. Absorption of both primary and secondary radiations on the forward and backward path is also taken into cognizance in the derivation of the formulas, in addition to geometrical factors. The computational results made it possible to evaluate the effect of changes in the contents of ac- companying elements on the intensity of x-rayfluorescence emission from the element to be determined in the analysis. The example of x-ray fluorescence analysis of lead in complex ores (lead?zinc and lead ?barium ores) is invoked to demonstrate that the relative error in lead determinations due to redistribu- tion of the accompanying elements may attain the level of several tenths of a percent (when secondary emis- sion is excited with the aid of a Cdin radioisotope source, and the L-emission from lead is recorded). The feasibility of recording the ratio of intensities of L-fluorescence in lead and singly scattered radiation of energy,?:z:20 keV, i.e., the ratio rpb = Npu/Ns, in order to minimize the dependence of the results of analysis of lead-containing samples on changes in the content of accompanying elements (barium, zinc, iron), is demonstrated on the basis of computational data. The computational results are in close agree- ment with experimental data obtained in measurements taken on artificial mixtures simulating lead?zinc ores in chemical composition. The possibility of minimizing systematic errors due to changes in the composition of samples to be analyzed, on the basis of records taken of the above ratio, was again confirmed in x-ray fluorescence anal- ysis of lead in lead?barium ores. Three distinct methods of dealing with the effect of changes in lead con- tent on the results of a zinc determination were pointed out in x-ray fluorescence analysis of zinc in lead ?zinc ores: 1) using one of the calibration graphs (corresponding to a specified lead content) in a family of TIZn = NZnK/NS = f(Czn) graphs as an aid in interpreting the analysis data; 2) determining zinc content from a calibration graph in the family of NznK = yo(Czn) graphs; 3) using the "compensation" technique. The first of these methods is preferred in this application. Translated from Atomnaya Energiya, Vol. 32, No. 3, p. 224, March, 1972; abstract submitted April 9, 1971. 247 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 THE SPECTRUM OF SECONDARY ELECTRONS IN MATTER IRRADIATED WITH y-RAYS A. M. Kollchuzhkin, V. V. Uchaikin, UDC 539.124.16:535.33 and V. I. Bespalov The secondary electrons accompanying the propagation of y-rays through matter produce ionization, radiation damage, and other effects. The electron energy spectrum is related to the differential y-flux by the expression (De (r, E)=-- dC dV' d12' dE'G, (r C, E; i', , E') dS2" (S2' , E'; 52", E") 01, (r' , 52", E"), where Zey is the differential cross section and Ge is the Green's function for the electron transport equa- tion. For photon energies below 10 MeV the variation of the y-flux over a distance of the order of a mean free path of the secondary electrons is negligible and therefore neglecting the variation of the y-flux in the integrand we obtain ape (r, E)-= f dE'G, (E; E') f dE"L (E'; E") (by (r; E"), ?where Ge(E; E') is the equilibrium electron spectrum. The equilibrium electron spectrum is calculated by the Monte Carlo method using the catastrophic collision model. A lexicographical scheme is used to construct and process the branching electron trajec- tories, and the flux is determined from the mean free path, since this method is more efficient than an estimate from the collision density. The program uses a new method of sampling the electron energy from a Moller distribution. An analysis of the catastrophic collisions model shows that the results of the calculations are prac- tically independent of the magnitude of the threshold energy separating the catastrophic collisions from con- tinuous energy losses, and the calculation time decreases significantly with an increase in threshold. Calculation of the secondary electron spectrum in lead for E0 = 1 MeV shows that scattered y-radia- tion can make a significant contribution to (13e(r, E), and the contributions of the photoelectric effect and Compton scattering can be equally important for unscattered radiation. The conclusion drawn in [1, 2] that unscattered y -radiation and Compton scattering are predominant is valid only for thin absorbers of low Z materials and high enough -y-energies. LITERATURE CITED 1. 0. Oen and D. Holmes, J. Appl. Phys., 30, 1289 (1959). 2. J. Cahn, ibid., p. 1310. Translated from Atomnaya knergiya, Vol. 32, No. 3, p. 224, March, 1972. Original article sub- mitted May 27, 1971; abstract submitted November 23, 1971. 248 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 Declassified in Part - Sanitized Copy Approved for Release 2013/03/01 : CIA-RDP10-02196R000300100003-8 LETTERS TO THE EDITOR FEW-GROUP COMPUTATIONAL MODEL OF SLOWING-DOWN FOR WATER ? ALUMINUM CORES E. A. Garus-ov and Yu. V. Petrov UDC 621.039.51 A program of constantly updated reactor calculations is greatly aided by the availability of a simple, but at the same time reasonably exact, theoretical model of slowing-down of neutrons. The simplest such model is the few-group diffusion model [1]. But the possibility of utilizing this model in the design of cores containing ordinary water is not at all obvious because the conditions for validity of the diffusion approxi- mation are violated. As a consequence of the steep decline in t