Soviet Atomic Energy Vol. 56, No. 5

Body: Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 SOVIET ATOMIC ENERGY Soviet Atomic Energy is abstracted or in- dexed in Chemical Abstracts, Chemical Titles, Pollution Abstracts, Science Re- search AbStracts, Parts A and 8, Safety Science Abstracts Journal, Current Con- tents, Energy .Research Abstracts, and Engineering Index. Mailed in the USA by Publications Expediting, Inc., 200 Meacham Ave- nue. Elmont, NY 11003. POSTMASTER: Send address changes to Soviet Atomic Energy, Plenum Publish- ing Corporation, 233 Spring Street, New York, NY 1 001 3. Soviet Atomic Energy is a translation of Atomnaya Energiya, a publication of the Academy of Sciences of the USSR. ? An agreement with the Copyright Agency of the USSR (VAAP) makes available both advance copies of the Russian journal and original glossy photographs and artwork. This serves to decrease the necessary time lag between publication of the original and publication of the translation and helps to improve the quality of the latter. The translation began with the first issue of the Russian journal. Editorial Board of Atomnaye tnergiya: Editor: 0. D. Kazachkovskii Associate Editors: N. A. Vlasov and N. N. Ponomarev-Stepnoi Secretary: A. I. Artemov I. N. polovin V. V. Matveev V. I. II 'ichev I. D. Morokhov V. F. Kalinin A. A. Naumov P. L. Kirillov A. S. Nikiforov Yu. I. Koryakin A. S. Shtan' E. V. Kulov B. A. Sidorenko B. N. Laskorin M. F. Troyanov E. I. Vorob'ev Copyright 01984, Plenum Publishing Corporation. Soviet Atomic Energy partici- pates in the Ccipyright Clearance Center (CCC) Transactional Reporting Service. The appearance of a code line at the bottom of the first page of an article in this journal indicates the copyright owner's consent that copies of the article may be made for personal or internal use. 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Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya November, 1984 Volume 56, Number 5 May, 1984 ARTICLES CONTENTS Engl./Russ. Quantitative Estimation of"NUclear Safety ? B.P. ShiShin. .... ? ? Emissions of " ilom . Co, Ag and 541ih at the Armenian'NuClear'Power 287 275 Plant, and Their Content in the Surrounding Atmosphere ? E. S. Saakov, A. A. Avetisyan, and K. T. Pyuskyulyaw. . . ... ? 291 278 Unloading Additional Absorbers from the RBMK1000 Core ? N. V. Isaev, V. E. Druzhinin, and Yu. V. Shmonin 294 280 Dissolution of Oxide Films on Constructional Steels ? Yu. G. Bobrov, G. M. Gueyanov, A. P. Kovarskii, Yu. P. Kostikov and A. V. Mbtornyi 298 282 Influence of Cold Deformation on the Behavior of Helium in Steel OKh16N15M3B ? A. G. Zaluzhnyi, M. V. Cherednichenko, O. M. Storozhuk, V. F. Reutov, and G. T. Zhdan 304 286 Isotopic Composition of Fuel in the Blanket of a Hybrid Thermonuclear Reactor with a Thorium Cycle S.-V. Marin and G. E. Shatalov 307 289 Recuperator with Inhomogeneous Electric and Magnetic Fields ? S. K. Dimitrov and Ya. A. MWnik 311 291 Calculation of Model High-Level Wastes in a Horizontal Apparatus ? V. V. Kulichenko, V. F. Savel'ev, V. A. Prokhodtsev, and A. A. Ryabova 314 293 Concentration Ratios for Radiogenic Lead and Uranium in Aureoles around Hydrothermal UraniumMineralization ? XT M. Ershov 319 298 Effect of Gamma-Neutron Radiation from a Nuclear Reactor on the Electrical Stability of Microlite 7 N. S. Kostyukov, M. I. Muminov, and V. M. Lanskov... . ... . . ......... 322 300 REVIEWS The New Generation of Highly Charged ion Sources ? K. S. Golovanivskii . . . . . .. ? ? ? ?, ? ? . ....... 326 303 LETTERS Small-Scale System for the Formation of a Field of Irradiation with Accelerated Electrons ? O. A. Gusev, S. P. Dmitriev, A. S. Ivanov, V. P. Ovchinnikov, M. P. Svin'n, and M. T. Fedotov. 336 311 Density and Surface Tension of Molten Mixtures of Uranium Tetrafluoride with Lithium and Sodium Fluorides ? A. A. Klimenkov, N. N. Kurbatov, S. P. Raspopin, and Yu. F. Chervinskii 339 312 Influence of Grain Size and Doping with Boron on the Behavior of Helium in Stainless Steel 16-15 ? A. G. Zaluzhnyi, M. V. Cherednichenko-Alchevskii, O. M. Storozhukm and A. G. Zholnin 341 314 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 CONTENTS Consideration Of the Decay of 23312u When Determining the Isotopic (continued) Engl./Russ. Composition of Uranium Fuel of a Hybrid Thermonuclear Reactor ? S. V. Marin and G. E, Shatalov,??????? OOOOOO "??? ? 6 343 315 Static Instability of Once-Through Steam Generators with Convective Heating ? I. I. Belyakov, M. A. Kvetnyi, D. A. Loginov, and S. I. Mochan 347 317? Cryogenic Loop for y-Ray Sources I. I Buzukashvili, and, G. S. Katumidze O ... ..? . ..... 351 319 Influence of Neutron Spectrum on Formation of .283U from 232Th Gerasimov, G. V. Kiselev, and A. P. Rudik 353 320 Spectral-Angular Distribution of yRadiation behind an. Instrumentation ? Unit? P. A. Barsov, V. M. SakharoV,? and V. G. SezenOv 357 322 The Russian press date (podpistuto k pechati) of this issue was 4/20/1984. Publication therefore did not occur prior to this date, but must be assumed to have taken place reasonably soon thereafter. Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 ARTICLES QUANTITATIVE ESTIMATION OF NUCLEAR SAFETY B. P. Shishin UDC 621.039.58 Ensuringnuclear safety of a critical stand or nuclear reactor, i.e., preventing un- controlled increase in the breeding coefficient of neutrons in a breeder system above unity, presumes the solution of a series of technical 'problems: the development of special control and protection systems, training of staff, and the production of documentation coordinating and limiting the functional activity of the staff and systems of the equipment. The de- signers, users, and inspection personnel must evaluate the nuclear-safety level of nuclear- physics equipment in the same way. In such conditions, it is important to formulate and employ objective methods of estimation as an instrument of, analysis, as a supplement to the Rules of Nuclear Safety. In' the present work, the possibility of quantitative estima- tion of the nuclear safety of apparatus, primarily critical stands, is demonstrated; it is reduced to the problem of analyzing the reliability of functioning of the systems and ele- ments of the equipment, using the apparatus of mathematical logic and probability theory. The nuclear installation takes the form of an active region in combination with systems for emergency protection, flooding the active region with water, parameter monitoring, etc. The service personnel are regarded as one of the elements of the systems. Each system con- sists of elements (sensors, valves, tubes, mechanisms, etc.), which may be in one of several possible States (operative, inoperative, emergency, etc.). The set of states of the elements determines the set of states of the system, which, in turn, are components of the set of states of the installation. The states if the elements may be taken in accordance with some value of the probability, i.e., the relative proportion of time in which the element is in a particular state. Transition of the system from one state to another in the course of the operating time is also characterized by a probability [1]. Emergency states of the installation correspond to some probability value, which is a measure of the "nuclear hazard," while its inverse is a measure of the "nuclear safety." Estimation of these quantities is an important problem of nuclear safety. Nuclear emergency of the installation may be regarded as the intersection of two events: the nuclear installa- tion is in a supercritical state, and increase in the nuclear flux above a specified limiting value in the active zone does.not trigger the emergency-control (EC) devices. The probability of nuclear emergency 35 (NE) is written in the form 35(NE.)=34(p>0)q(EC), (1) where g(p:>0) is the probability of a supercritical state of the reactor; o is the reac- tivity of the reactor; q(EC) is the probability of EC failure. The probability g4(1))..o) consists basically of two components. One -0,1(p>0) is determined by the operating program. In the course of this program, the reactor is sometimes in subcritical, critical, and super- critical states. The probability that the reactor will be in a supercritical state may be estimated if the operating program and the methods of its implementation are known. The second component 052(p>0) is determined by random unplanned processes in reactor control, when it becomes supercritical as a result of operator error or by breakdown or loss of func- tioning in the systems of the installation (systems of moderator and fuel filling, control and safety system, etc.). The probability of failure of the emergency-control system q(EC) is determined by the specific logic circuit and elements of the installation. If a limiting value of the nuclear- emergency probability is specified, then variation of the values of gi(fl>0) and q(EC) is permitted. The nuclear-safety level may be increased by decreasing the probability that the reactor will reach a supercritical state and increasing the reliability of operation of the EC system. Translated from Atomnaya Energiya, Vol. 56, No. 5, pp. 275-277, May, 1984. Original article submitted July 6, 1982. 0038-531X/84/5605-0287$08.50 01984 Plenum Publishing Corporation 287 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Pip,>nd 0,3 I ? 1 1 1 1 1 1 -5fi -10 47A rioo Fig. 1. Dependence of the probability of the reactor state on the reactivity. Fig. 2. System for flooding CP with water. In the present work, an example of estimating the nuclear safety of a critical stand (critical pile with a water moderator, a system for flooding with water, and a control and safety system) is considered, when experimental programs assocated with the attainment of a supercritical state are employed, with possible disruption of the functioning of the water- flooding system and EC failure. Estimating .951 (ID> 0). It is assumed that the critical pile (CP) in the flaw of working mixture first moves from a deep subcritical state (pp> ? (3 , it is approximately 0.1. In such conditions, the probability of random emergency flooding of the CP with water is estimated as (E) 0) --- pa pb+ pc? 840-5+5-10-7 +8-10-5=1,6.10-4. Estimating q(EC) In the given example, it is assumed that the EC system (Fig. 3) has two power-measure- ment channels (1-2 and 3-4), a reactivity-measurement channel (5-6), two groups of EC work- ing components, (8 and 9), elements of manual (10) and automatic (11) regulation, and a compensating unit 12. Sensors 1, 3, 5 send signals to the comparative instruments 2, 4, 6. When the signal exceeds a specified level, it is fed to commutating device 7, resulting in the triggering of all the working elements of the safety and control system (8-12). Failure of the EC system may occur in the case of failure of the following system ele- ments: 1 or 2, 3 or 4, 5 or 6, or even element 7, or element 8-12. The expression for the 289 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 probability of such an event is written in the form ( EC) - II (qi- qi+1) q7 -I- [I qi, (5) 3, 5 where qi is the probability of failure of the i-th element in the system; r is the product sign. The failure rate of the radioelectronic equipment, its components, the relays, drives, electric motors, is 10-4-10-6 h-1 [5-7]. The failure probability of such equipment in the course of a workingshift is 10-2-10-2. According to Eq. (5), the probability of EC-system failure in this case is approximately 10-2-10-3. For nuclear power plants, the approximate failure rate of the control and safety system is a few times 10-4 h-1 [5, 7]. Probability of Nuclear Emergency The probability of nuclear emergency in experiments associates with attainment of a supercritical state by the reactor, in the case of possible disruption of the functioning of the flooding system and RC system is calculated from Eq. (1) with - (p > 0) (p >0) + 1,2(p>0). ? With the values of fP (f) >0), g's (1) > 0) , and q(EC) determined above, the probability of such an emergency is p (NE) = 10-4-10-6. The maximum contribution comes from AC failure in planned movement of the reactor to a supercritical state in the ?course of ?the experiments. The next most probable emergency is associated with erroneous release of water from volume 5 and emergency discharge through channel 1-7 (Fig. 1). If V' (NE) is the probability that there will be an emergency in the course of a single operating shift, the probability of such an emergency in the course of a year (100 operating shifts) is 10-2-10-4. A similar analysis may be conducted with respect to other nuclear-installation systems. The development of quantitative methods of estimating nuclear safety will facilitate the adoption of more accurate solutions and the reduction in material costs in ensuring the safety of nuclear-physics installations. LITERATURE CITED 1. V. V. Frolov and V. I. Bulanenko, At. Tekh. Rub., No. 1, 3 (1981). 2. Handbook on Engineering Phychology [in Russian], Mashinostroenie, Moscow (1982). 3. R. Lloyd et al., Nucl. Technol., 42, 13 (1979). 4. S. M. Trunin et al., Reliability of Marine Machines and Mechanisms [in Russian], Sudostroenie, Leningrad (1980). 5. F. Ya. Ovchinnikov et al., At. &erg., 50, No. 4, 248 (1981). 6. Proceedings of Seventh, Eighth, and Ninth Symposium on Natural Reliability and Quality Control, Washington (1961-1963). 7. Proceedings of a Symposium an the Reliability of Nuclear Power Plants, IAEA, Vienna (1975). 290 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 .EMISSIONS OF "Co, 11?ThAg, AND 541111 AT THE ARMENIAN NUCLEAR POWER PLANT, AND THEIR CONTENT IN THE SURROUNDING ATMOSPHERE E. S. SaSkov, A. A. Avetisyan, UDC 551.510.721.614.73 and K. I. Pyuskyulyan Six years of operation of the Armenian nuclear power plant (NPP) with a VVgR-440 led to an increase in the contribution of radioactive corrosion products, in particular "Co, 1"1'Ag, and 54Mn, to the total activity of the gas?air mixture discharged into the stack (Table 1). According to the data for January-September, 1982, these radionuclides contributed 57% of the 31 total activity of the long-lived aerosols discharged into the atmosphere, neglecting 11. Consequently, the yearly emission of long-lived radioactive aerosols is due largely to ra- dioactive corrosion products. The "Co enters the coolant from abrasion resistant alloys containing this element (pump bushings, control and safety rod actuators, etc.), and from austenitic stainless 4 steels. The S Mn comes from the same structural materials, and the 110Ag from the silver in alloys used in the thermoelectric heater of the volume compensator and control devices. Most of the radioactive corrosion products are deposited on the inner walls of the primary loop system. As a result of changes in the coolant velocity and temperature and the water-chemical condition during transient operation there is a change in the relative amounts of corrosion products deposited on the walls of the equipment and in the coolant; during shutdown the amount of ammonia and boric acid in the primary loop is increased. Thus, it was determined that-with a drop in power the specific activities of "Co, 110mA8, and 54Mn at the Kola NPP increased by factors of 50, 80, and 110 respectively. According to data in [2], about 25% of the corrosion products in the coolant are in the form of large particles, about 75% are colloid, and the ion fraction is less than 1%. The concentrations of the nuclides mentioned reach a maximum in 't,46-48 h after a drop in power. Then the specific activity decreases, and the coefficient of precipitation of radioactive corrosion products becomes an order of magnitude, smaller ?than that obtained under steady reactor operation El]. In view of this, one should expect that during reactor startup and shutdown the emission of radioactive corrosion products would be appreciably increased. Using the daily measure- ments of the total activity of the discharge of long-lived aerosols during the period of planned preventive maintenance on the first block of the Armenian NPP, we plotted a graph (Fig. 1) of the time variation of the activity of the discharge. The figure shows that during cooling, hydraulic testing, and reactor startup, the total activity of the discharge increase sharply. The same conclusion can be drawn from spectrometric analysis (Table 2). TABLE 1. Contribution of Radionuclides to Total Activity of Discharge of Long-Lived Aerosols, % Period "Co I"mAg 'Mn 1918 2,4 1,9 1,7 January-September 1982 12,0 14,0 5,0 Translated from Atomnaya gnergiya, Vol. 56, pp. 278-280, May, 1984. Original article submitted September 20, 1983. 0038-531X/84/5605-0291$08.50 01984 Plenum Publishing Corporation Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 291 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 36 40 - III i I i 27 1 S 9 13 17 21 2329 2 5 10 July August Fig. 1. Total activity of discharge of long-lived aerosols into the stack during: I) cooling; II) unpacking main joint, opening Volume compensator; III) refueling; IV) hydraulic testing, output at power. TABLE 2. Rate of Discharge of Radionuclides in 1982, 107 Bq/month Period 61410 ".?mAg 541Wn January 1,65 1,30 '0,69 February 3,43 . ? 3,20 . 1,30 March 2,60 1;42 0,84 April ? 20,80 6,75 7,22 May 7,45 7,22 ? 2,37 June 5,21 7;58 ? 1;78 J 4,00 1,85 13,02 August 4.02 2,96 ? 0,91 September . 1.66 . 1,54 0,71 TABLE 3. Concentration of Radionuclides in Air At 1 km from NPP in 1982, 10-4 Bq/liter Period, . ? 6IiCo 110MAg -54Mn 1st quarter 0,62 -0,64 Traces 2nd quarter 1,40 3,00 1,10 July .1,70 4,80 1,00 August 0,80 .0,80 Traces :September Traces . Traces Traces The increase in the discharge of 60co, ilomAg, and "Mn coincides with the periods of planned preventive maintenance at the NPP. The time variations of the discharge of "Co and 5411a are identical, since they have a common source. The behavior of 11?mAg is somewhat different; In particular there was a sharp increase in the discharge of silver in July due to opening the manhole of the volume compensator and repairing it during planned preventive maintenance on the first block. 292 Air samples were taken at seven check points in various directions from the Armenian NPP Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 TABLE 4. Comparison of Calculated and Mea- sured Maximum Concentration of Radionuclides During July, 1982, 10" Bq/liter Radionuclide 6410 110mAg 542.1n' Cale. values 1,10 3,60 0,51 Meas. values 1,70 4,80 1,00 and at distances of 1, 5, 6, 11, 14, 15, and 50 km from it. The spectrometric analysis was performed by using a DGDK-32A-1 semiconductor detector of AI-4096 and UNO-4096-90 analyzers. The minimum detectable concentration of aerosols estimated by the formula in [3] is 0.4x 10-5 Bq/liter with an error of 30% (Table 3). Only traces of radionuclides were found in samples taken at 5 and 6 km, and none were found in the samples taken at the remaining points. The results are confirmed by analysis of sedimentation samples. Samples of top soil showed only insignificant amounts of 157Cs and 50Sr due to global fallout. Evidently the recording of these radionuclides in the air near the Armenian NPP is due to specific climatic conditions in the region. The climate of the Ararat valley is clearly continental, with an arid summer and a short rarely cold winter. The atmospheric precipita- tion is 250-300 mm. During the long calm summer with an abundance of sunny days, the air is heated up to 40?C. The wind velocity in the industrial region of the Armenian NPP does not exceed 3 m/sec. Under the strong solar heating of the earth's surface, the air temperature decreases with height, so that the vertical temperature gradient is larger than adiabatic. Such conditions lead to intense vertical displacement of the air. The maximum concentration near the ground for the present stack height is relatively large, but it decreases rapidly with distance from the source, depending on the wind direction. The minimum value of Kd, the average monthly dilution ratio of the contaminant, was calculated from metereological data obtained at stations in the NPP region, and the formula recommended in [4]. The calculated value of Kd was 1.4 x 106 m5/sec. For comparison we point out that the minimum yearly average dilution ratio for the center of our European territory is 3.4 x 106 m5/sec [4]. The maximum ground level concentrations of "Co, 116mAg, and 541111 were found form the measured values of their emission and the calculated value of Kd. Within the limits of error, the calculated values agree with the spectrometric analysis of air samples taken at 1 km from the NPP (Table 4). The data obtained justify the following conclusions and recommendations: the possibility of recording 60co, 110mAg, and 561n in the air at 1 km from the Armenian NPP for emission which does not differ from that of other NPP with VVgR, is due to the specific climatic conditions at its location; the emission of radioactive corrosion products during reactor shutdown and startup can be reduced by using filters to retain the deposits, after loosening them and scraping them off; in determining the isotopic composition of the radionuclides in the layer of the atmos- phere near the ground, and the radiation doses received by people living near NPP, it is necessary to take account of the contribution of radioactive corrosion products. LITERATURE CITED 1. V. B. Gall' et al., in: Radiation Safety and Shielding of NPP [in Russian], No. 5, Atomizdat, Moscow (1981), p. 5. 2. I. K. Morozova et al., Removal and Deposition of Corrosion Products of Reactor Materials [in Russian], Atomizdat, Moscow (1975). 3. S. M. Vakulovskii et al., in: Methodological Recommendations for Monitoring the Radio- active Content of Objects in the Environment [in Russian], Moscow (1980), p. 234. 4. N. E. Artemova et al., Admissible Emissions of Radioactive and Harmful Chemicals into the Ground Layer of the Atmosphere [in Russian], Atomizdat, Moscow (1980). 293 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 UNLOADING ADDITIONAL ABSORBERS FROM THE RBMK -1000. CORE N. V. Isaev, V. E. Druzhinin, UDC 621.039.539.1 and Yu. V. Shmonin In the initial loading of the RBMK-1000 reactor, the considerable reactivity margin is reduced by loading the channels only incompletely with fuel assemblies (FA), absorbers being inserted in the unfilled channels. The core then contains 234 additional absorbers (AA) and 44 unloaded channels containing water columns (WC). The RBMK-1000 design envisages a strategy for reloading the channels by dividing them up into periodicity cells (polycells) of size 4x 4 cells each [1]. One channel in each polycell is reloaded, and to maintain the loading symmetry and periodicity, one unloads channels identically placed in the polycell. This method has been used in reloading the reactors at Kursk nuclear power station. Here we examine the effects of AA unloading order on the economic parameters of the second-generation RBMK-1000 reactors in the steady state. As the safety state is attained at 7-8 years after commissioning, the optimum AA unloading strategy in the transition period can be examined only on theoretical models. The unloading of the. AA involves about 670-700 effective working days. Model Description. The REF program [2] is used in calculating the transient state in the RBMK and envisages a specially homogenized core. This program is widely used in de- termining RBMK characteristics, so the algorithm is not described here. The OPERA program for optimizing apparatus reloading [3] enables one to perform a detailed full-scale two-group diffusion calculation on the RBMK parameters under conditions of partial and ongoing reloading, including the adjustment of the power production field by varying the insertion lengths of the control and protection rods (CPR). The OPERA program is based on the OPTIMA one [4]. A difference from the OPTIMA algorithm, which is described in [4], is that the OPERA program uses an improved algorithm that instead of using the CPR to compensate the reactivity and equalize the power production (while retaining kef) enables one to bring the reactor to a given kef by automatic selection of the necessary rods in the CPR set. In simulating RBMK reloading, the power production field is profiled to a certain set field WTeg. The following constraints must be met in using the OPERA program: 1. The insertion depth hi of the CPR is in the range 0 Uo. The mechanism of the rise in efficiency with increasing Um has not been ascertained as yet. For comparison Fig. 3 gives the values found for the efficiency from the formulas presented by Timofeev [1]. Thus, the following conclusions can be made. For the conversion of the energy of plasma fluxes containing particles of both signs with a small longitudinal velocity component it is expeditious to use the recuperator described by Timofeev. But for nonmonoenergetic beams of Charged particles of the same sign with an arbitrary ratio cn /6,I the recuperator presented in this paper is more efficient (11:-.30.7 for e4 /el.:0.9 and RA 0.7). Inorder to increase its efficiency, it is necessary to increase the size of the apparatus, i.e., decrease the ratio Ro/R. The recuperator discussed here can be used in the neutral-particle injectors of thermo- nuclear machines, microwave devices (traveling-wave tubes, klystrons), electron-beam valves, etc. LITERATURE CITED 1. A. V. Timofeev, Fiz. Plazmy, 4, No. 4, 826 (1978). 313 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 CALCULATION OF MODEL HIGH-LEVEL WASTES IN A HORIZONTAL APPARATUS V. V. Kulichenko, V. F. Saverev, UDC 621.039.7.14 V. A. Prokhodtsev, and A. A. Ryabova In the processing of radioactive wastes by calcination [1], the wastes are dried and the salts are thermally decomposed, the final product consisting mainly of oxides. As the wastes contain many different compounds, a wide temperature range is used: first the nitrates of the rare-earth and other trivalent elements decompose, then the nitrates of the alkaline- earth metals, and finally at the highest temperatures and with the lowest rates the same applies to the alkali-metal nitrates. Calcination is considered as the first stage before vitrification, the production of ceramics, etc. Various calcination methods have been devised based on equipment of sprayer type, fluidized beds, and of horizontal type with a rotor or rotating body [2-4]. A horizontal plant is much less sensitive to the waste composition than is a fluidized-bed or sprayer one, and the volume of the gas discharged is determined only by the vapor-gas mixture formed during the drying and calcination. However, the design is more complicated, since there are moving mechanisms in the rotor (worm conveyer) or body. Also, the calcination is conducted in a thin layer and heat must be supplied through the wall of the apparatus, which makes it difficult to design a horizontal plant of large throughput. A horizontal calcinator of throughput 40 liter/h [3-4] has been operated for several years in France in a plant for vitrifying high-level wastes. The calcinator consists of a heated tube of diameter 27 cm and length 3.6 in rotating at 30 rpm, which is tilted at 1047. Within it there is a free rod to grind up the solid product. The liquid wastes contain mainly the nitrates of aluminum (up to 81 g/liter), sodium (19-23 g/liter), and iron (15-17 g/liter) together with other salts at much lower concentrations. The product from any apparatus must be as dry as possible and contain the minimum amount of nitrate (source of corrosive oxides of nitrogen) and should flow freely, in order to provide for transfer to the vitrification apparatus. We have examined the calcination conditions in a horizontal plant operating with wastes having elevated sodium nitrate contents. The drying and calcination were performed in a horizontal tubular apparatus containing a rotating worm conveyer to mix and transport the product through the working zone into the bunker (Fig. 1). The length of the working zone was 1400 mm, internal diameter 160 mm, gap between projections on the blades of the worm conveyer and the inner wall of the apparatus about 2.5 mm. The apparatus was heated by six external demountable heaters of power 2.5 kW each. The temperatures in the sections of the working zone were 250-800?C in accordance with the required conditions. The nitrate solutions were supplied by a dispensing pump to the input, where they were dodatered, denitrated, and calcinatedas theymoved thorughtheworkingzone. The solid product in the form of powder or granules was transmitted by the worm conveyer to the bunker. The steau gas mixture containing oxides of nitrogen passed successively through a filter (to remove aerosols), a cooling condenser (to collect condensate), and to a bubble tower con- taining alkaline solution to neutralize the oxides of nitrogen. The final product was analyzed for nitrogen, water, total carbon, carbon in the form of carbonate, and iron, and some mechanical properties were determined. The specific surface was measured by thermal nitrogen desorption; the data curves were recorded with an OD-103 instrument by the standard method. Table 1 gives data on the calcination of model solutions of various compositions in the presence of molasses and without them at temperatures below 800?C with throughputs of 4.0-17 liter/h and time spent by the product in the apparatus of 2 and 4 min. Translated from Atomnaya gnergiya, Vol. 56, No. 5, pp. 293-.297, May, 1984. Original article submitted August 18, 1983. 314 0038-531X/84/5605-0314$08.50 1984 Plenum Publishing Corporation Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 0 1== 0 0. As x2 + x3 increases the value of R grows insignificantly and the decreases for all the regimes considered, except 0=10 neutrons/cm2..sec and y = 0.4, its value being R< 1 at the maximum of x2 + x3 (this is attributed to the contribution of neutrons formed in the fission of 233U). The value of R increase substantially with the growth of 0 and y and the weakening of the blocking since in this case the 233Pa content rises and the 233U fraction decreases correspondingly and together with it, the contribution of 233U fission neutrons. This effect manifests itself most clearly in the regime with 0=1014 neutrons/cm2osec and y = 0.4: Thenumber of 233U fission neutrons is insufficient to compensate the external neutrons absorbed and R> 1 for all values of x2 + x3 (we note that in this case R increases as the blocking is intensified). Energy Release E. During the fission of the 233U formed energy is released in the 232Th charged. In Fig. 6 we give the dependence of E on x2 +x3 per ton of the initial 232Th. The energy released in one 232Th fission event was assumed to be 191 MeV [7]. For a fixed x2 + x3 the energy release E decreases as and y increase and the blocking becomes weaker; this is explained by the lowering of the fraction of 233U in the mixture of 233Pa and 2331J. Clearly, the energy release is proportional to the number of neutrons produced during the fission of 233U. Conclusion. The results of the calculations are of interest from two points of view: First, the main physical characteristics of the 232Th 233U process are presented as a func- tion of the flux density and the energy spectrum of thethermal neutrons, making it possible to evaluate the efficiency of one reactor design or other; second, the decrease in the 232Th consumption and the increase in the neutral consumption R as y grows give reason to hope that the problem of economic optimization for the 232Th 233U process will have a nontrivial solution. LITERATURE CITED 1. 2. V. M. Murogov,}4. F. Troyanov, [in Russian], Energoatomizdat, A. S. Gerasimov, A. K. Kruglov, and A. Moscow and A. N. Shmelev, Use of Thorium in Nuclear Reactors (1983). P. Rudik, At. nerg., 51, No. 4, 237 (1981). 356 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 3. S. Mughabhab and D. Garber, Neutron Cross Sections, BNL-325, 3rd ed. (1973), Vol. 1. 4. G. N. Gusev and P. P. Dmitriev, Radioative Chains (Handbook) [in Russian], Atomizdat, Moscow (1978). 5. L. V. Matveev and g. M. Tsenter, At. Tekh. Rubezhom, No. 4, 10 (1980). 6. T. S. Zaritskaya et al., At. gnerg., 48, No. 2, 67 (1980). 7. I. V. Gordeev, D. A. Kardashev, and A. V. Malyshev, Nuclear-Physical Constants Russian], Gosatomizdat, Moscow (1963): SPECTRAL-ANGULAR DISTRIBUTION OF y RADIATION BEHIND AN INSTRUMENTATION UNIT P. A. Barsov, V. M. Sakharov, UDC 621.039.538.7 and V. G. Semenov The weight fraction of the various equipment and devices is appreciable on transporting apparatuses with nuclear reactors; therefore taking correct account of the shielding proper- ties of these items is rather important. The possibility of using an effective attenuation coefficient to estimate the shielding properties of the equipment, which has been modeled in such a way that a binomial distribution of the units of material was realized, has been verified experimentally in [1, 2]. It has been shown that the dose is attenuated exponential- ly along the axis of a wide radiation beam with an effective coefficient determined by the binomial law. The possibility of using an effective attenuation coefficient in calculations by the Monte Carlo method of the differential characteristics of y radiation behind actual equip- ment units is discussed in this paper. With this goal we have performed experiments and cal- culations in the identical geometry. In the experiments monoenergetic y radiation of a point isotropic 137Cs source with a photon energy Ey = 662 keV was directed at a rectangular equipment unit with a geometrical thickness at the irradiation site of Z = 53 cm and an average density of p = 0.43 g/cm3. We used a scintillation spectrometer with a CsI(T1) crystal 16 mm in diameter and 40 mm in height placed in a conical collimator as the le-radiation detector. The isotropy of the de- tector is no better than ?5% within the limits of the collimation angle. The distance be- tween the equipment unit and the detector was selected so that all the y radiation emergent from the unit in the direction of the detector was recorded. Measurements were made for angles of 0, 30, 60, and 90? to the inner normal relative to the irradiated surface. The error of the instrumental spectra was q,30% in the region of Ey ' 130-250 keV and 15-20% for higher values of the energy. It did not prove possible to make measurements for Ey < 130 keV due to the high background. The calculations were performed by the Monte Carlo method using the BSERAD program [3], whose geometrical section simulated the geometry of the source, detector, and equipment unit made out of aluminum. The effective attenuation coefficient of the unscattered y radiation used to determine the mean free path of a photon by simulating the trajectory was found using the formula [4]: Translated from Atomnaya gnergiya, Vol. 56, No. 5, pp. 322-323, May, 1984. Original article submitted October 12, 1983. 0038-531X/84/5605-0357$08.50 ?1984 Plenum Publishing Corporation 357 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved ForRelease2013/09/14 : .a --, c...) 1 ' 60 - I 1s a c 1 ON ....-? ?. 0 0 cl?P+0-. 10 m 41':1.i..- r. v o 41 ?- s..a .4 - a -0 a?0 a. a J `1 r? ..., L. n .. ? I 1 b .., 0 CIA-RDP10-02196R000300040005-3 5"E a 1810 150 270 2/.7 bV E keV Fig. 1. Calculated and experimental instrumental spec- tra of the y radiation behind an equipment unit at angles to the beam axis of (a) 6 = 0, (b) 60, (c) 30, and (d) 900 with a photon energy of 662 keV. ---) ex- periment; - - - -) calculation for the quipment unit; ) calculation for a layer of aluminum 7.2 cm thick with a density of 2.7 g/cm3. Iii 11.i 0yAlc,, 714, (1) The The distribution function of the material thickness f(x, Z) was obtained experimentally by the method of y transillumination of the unit in different directions [5]. We used the sensitivity matrix determined experimentally as the sensitivity function of the detector in the calculations. The error of the spectra calculated for the indicated angles does not ex- ceed 15% over the entire energy range. The calculated and experimental instrumental spectra normalized per photon emitted from the source in the direction of the equipment unit are similar in shape for all detection angles (see Fig. 1). A certain systematic understating of the computational results in com- parison with the experiment;11 results in the region of the single-scattering peak is evidently associated with the use of an attenuation coefficient for aluminum, whereas the unit being investigated contains an appreciable amount of lighter elements (hydrogen, carbon, oxygen). The total mass attenuation coefficient for these elements is smaller than for aluminum, which leads to the indicated difference of 20-307.. The dependence of the y-radiation intensity on the recording angle is of an exponential nature. The value of the characteristic angle 6 = 47? turned out to be close to its value for a homogeneous layer of aluminum and a point isotropic 137Cs source [6]. This fact served as the reason for doing similar experiments and calculations for a homogeneous layer of alu- minum with PM, = 2.7 g/cm3 and Z = 7.2 cm, which corresponds to an optical thickness of the equipment unit, which is defined as the product of the effective attenuation coefficient by the geometrical thickness of the unit in the direction of the incident beam of y-photons. The thickness of the layer, which is calculated as pZ/PA1, is equal to 8.4 cm. The experiments and calculations for a layer of aluminum 7.2 cm thick have shown that the spectral-angular distributions behind the layer and behind the equipment unit are prac- tically identical. At the same time the calculations for a layer of aluminum 8.4 cm thick give results which are understated by a factor of 1.5 in comparison with those given for the equipment unit over the entire energy range. 358 Declassified and Approved For Release 2013/09/14 : CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 The calculations performed for an optical thickness of the unit of px = 4 have shown that the use of a homogeneous layer with thickness peffx leads to an understating by a factor of 1.5 of the flux of y-photons in the region Ey > 400 keV and to some softening of the spec- trum in contrast to the direct calculations, in which the distance between successive in- teractions is determined with the use of veff. The investigations carried out have led to the conclusion that the passage of y radiation in equipment with a thickness Z < 30 g/cm2 (px,2) can be calculated as in a homogeneous mate- rial with the use of the usual interaction constants, determining the optical thickness of the layer with the help of the effective attenuation coefficient. For equipment of greater thickness it is necessary to use the effective attenuation coefficient. The latter asser- tion requires further experimental verification however. LITERATURE CITED 1. V. V. Bolyatko et al. in: Problems of Dosimetry and Shielding from Radiation [in Rus- sian], No. 8, Atomizdat, Moscow (1968), p. 80. 2. P. A. Barsov et al., in: Abstracts of Lectures at the 2nd All-Union Scientific Conference on Shielding from Ionizing Radiations of Nuclear Engineering Facilities [in Russian], Atomizdat, Moscow (1978), p. 40. 3. P. A. Barsov et al., in: Radiation Safety of Nuclear Power Plants [in Russian], No. 26, Moscow (1979), p. 75. 4. A. V. Kolomenskii et al. At. Energ., 44, No. 6, 517 (1978). 5. V. V. Bodin et al., in: Abstracts of Lectures at the 2nd All-Union Scientifiec Con- ference on Shielding from Ionizing Radiations of Nuclear Engineering Facilities [in Russian], Atomizdat, Moscow (1978), p. 103. 6. Yu. A. Kazanskii et al., Physical Investigations of Reactor Shielding [in Russian], Atomizdat, Moscow (1966). 359 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 Declassified and Approved For Release 2013/09/14: CIA-RDP10-02196R000300040005-3 MEASUREMENT TECHNIQUES lzmeriternaya Tekhnika? 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