Soviet Atomic Energy - Vol. 39, No. 1

Body: Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Russian Original Vol. 39, No. 1, July, 1975 January, 197 SATEAZ 39(1) 571-666 (1975) SOVIET ATOMIC ENERGY ATOMHAFI 3HEPITIR (ATOMNAYA iNERGIYA) TRANSLATED FROM RUSSIAN rc" q4.) CONSULTANTS BUREAU, NEW YORK Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 :pup: ATOM IC 'ENERGY' 't? , Soviet Atomic Energy is' abstrated, or in- ? - dexed in Applied? Mechanics Reviews, Chem- ical Abstracts, Engineering Index, INSPEC? Physias Abstracts and Electrical and Elec- tronks ? Abstracts, Current Contents, and Nuclear Science Abstracts: 7'1 ' Soviet Atomic Energy is a cover-to-cover translation of Atomnaya , Energiya,-a publication of the Acadeiny of Sqences of the USSR! , ? An agreement with the Copyright Agericy of the USSR (VAAP) Makes available, beth advance 'copies of Russian journal and original glossy photographs and artwork. This serves to decrease the necessary time lag between publicatkm of thepriginal ariad publicatiqn of the transletion and helps to improve the quality Of the latter. The transletion began with the first issue of the ? Russian journal. ' /Editorial Board of Atomnaya 'Energiya: ? Editor: M. D. Millionshchikov Deputy Director , I. V. Kurthatov Institute of Atomic Energy Academy of Sciences of the USSR. Moscow, USSR ,Associate Editor: N. A. Vlasov A. A. Bochvar N. A. Dollezhal' Fursov , , I. N. Golovin V. F. Kalinin A. K. Krasin ? A. P. Zefirov V. V. Matveev ? M. G. Meshcheryakov P, N:Palei V. B. Shevchenko V. I. Smirnov ' A.' P. Vinogradov ? Copyrigbt C) 1976 Planum Publishing Corporation, 227 West 17th Street, New liork, N.Y. 10911,? All rights reserved. No article contained kerein may ,be reproduced, stored in a retrieval system; or transmitted, in any form'br by any means, electronic? mechanical, photocopying, microfilming, recording or otherwise, without Written permission of the publisher. ? ' Consultant's Bureau journals' appear.about"six nionths after, the publication of the origirtal,,Russiari 'issue. ,cir bibliographic accuracy, the Engfish issue publithed consultants Bureau carries the same number?and date as the'original Russian frOm which it was translated. For example, a Russian issue published in December will- appear in a Consultants Bureau English translation about the following June, but the translation issue will car,ry the December date. Whe'n ordering any volume or particu- lar issue of a "consultants Bureau journal, please specify the date and; where appli- cable, the volume and issue numbers of the,origirial Russian. The material Vou will receiveWiltbe a translation of that Russian volume or issue. , Subscription $,871.50 per volume, (6 Issues) Pr'ices somewhat higher aCitside he United States. _ , CONSULTANTS BUREAU, NEW 227 West 17th Street New York, New York 10011 , Published monthly. Single Issue: $50 Single Article: $15 YORK AND LONDON . ? tia Lower John Street London W1R 3PD - England , Second-class postage paid at Jai=snaiCa, New York 11431. Declassified and Approved For Release 2013/09/24:' CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya January, 1976 Volume 39, Number 1 July, 1975 ARTICLES Investigation of the Physical Characteristics of the Reactor during Startup of the First Unit of the Bilibinsk Nuclear Power Station ? A. A. Vaimugin, CONTENTS Engl./Russ. V. V. Bondarenko, V. K. Goryunov, A. V. Gusev, B. G. Dubovskii, P. G. Dushin, A. N. Efeshin, L. D. Kirillovykh, I. M. Kisil', V. I. Kozlov, 0. V. Komissarov, E. V. Koryagin, A. G. Kostromin, N. I. Lagosha, M. A. Lyutov, M. E. Minashin, K. N. Mokhnatkin,A.P. Paniko, Yu. F. Taskaev, V. N. Sharapov, and A. I. Shtyfurko 571 3 BOOK REVIEWS V. G. Zolotukhin, L. R. Kimell, A. I. Ksenofontov,et al. The Radiation Field from a Point Unidirectional Source of Gamma Quanta ? Reviewed by B. R. Bergelyson 577 8 ARTICLES Some Problems of the Economics of a Research Nuclear Reactor ? V. I. Zelenov, S. G. Karpechko, and A. D. Nikiforov 579 9 BOOK REVIEWS A. A. Vorob'ev, B. A.Kononov, and V. V. Evstigneev. Betatron Electron Beams ? Reviewed by P. S. Mikhalev 583 11 ARTICLES Synthesis of a Digital System for Control of Neutron Flux Distribution ? E. V. Filipchuk, P. T. Potapenko, V. G. Dunaev, N. A. Kuznetov, and V. V. Fedulov 585 12 Absolute Measurement of the Radiative Capture Cross Section of 238U for 30 keV Neutrons ? Yu. G. Panitkin and L. E. Sherman 591 17 Heat-Transfer Crisis in a Steam-Generating Tube on Heating with a Liquid?Metal Heat Carrier (Coolant) ? A. V. Nekrasov, S. A. Logvinov, and I. N.Te"="6""""mg 595 20 --- X-Ray Diffraction Study of the Effect of the Temperature of Deformation in the Alpha Phase on the Quench Texture of Uranium Rods Containing Various Proportions of Iron and Silicon ? V. F. Zelenskii, V. V. Kunchenko, V. S. Krasnorutskii, N. M. Roenko, V. P. Ashikhmin, A. V. Azarenko, and A. I. Stukalov 599 24 A Loop Converter Channel for Testing Highly-Enriched Fuel Elements in a Research Reactor ? V. G. Bobkov, V. B. Klimentov, G. A. Kopchinskii, M. V. MePnikov, and V. A. Nechiporuk 603 28 Tests on Experimental Fuel Elements Containing Carbide Fuel,Irradiated in the LallS.1.) Reactor up to a Burn-Ups of 3 and 7% ? E. F. A. A. Maershin, 608 33- V. N. Syuzev, Yu. K. Bibilashvili, I. S. Golovnin, and T. S. Mentshikova Recollections of Professor Boris Vasil'evich Kurchatov, Doctor of Chemical Science, on His Seventieth Birthday ? S. A. Baranov, A. R. Striganov, and P. M. Chulkov 612 39 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 CONTENTS REVIEWS Problems in Shipment of Spent Fuel, from Nuclear Power Stations ? Yu.I. Arkhipovskii, (continued) Engl./Russ. 615 42 620 48 621 49 V. A. Burlakov, A. N. Kondratiev, E. D. Lyubimov, and A. P. Markovin ABSTRACTS Total Stability of a Nuclear Reactor with Connected Cores ? N. A. Babkin Frequency Criterion for the Stability of a Circulating-Fuel Reactor ? V. D. Goryachenko and V. V. Mikishev Estimation of the Effect of Physico-Geometric Factors on the Distribution of Delayed Fission Neutrons in a Borehole ? Yu. B. Davydov 622 49 Spatial Distribution of Fission Neutrons in a Breeding Medium, Crossed by a Drill Hole ? Yu. B. Davydov 623 50 Variable Mechanical Stresses, Induced in the Fuel Element Claddings of the IBR-30 Reactor by Power Pulses ? V. S. Dmitriev, L. S. Winskaya, G. N. Pogodaev, V. V. Podnebesnov, A. D. Rogov, V. T. Rudenko, and 0. A. Shatskaya 624 51 Correction of the Group Constants by the Results of Experiments on the BFS Critical Assemblies ? A. A. Van'kov and A. I. Voropaev 625 51 The Influence of Beam Noise on the Critical Current of Linear Electron Accelerators ? I. N. Mondrus 625 52 Model of Grouping of Low Energy Transfers in Calculating Electron Fields by the Monte Carlo Method ? A. V. Plyasheshnikov and A. M. Kol,chuzhkin 626 53 LETTERS TO THE EDITOR Use of a252Cf FissionChamber in Certain Physical Measurements ? V. F. Efimenko, V. K. Mozhaev, and V. A. Dalin 628 54 Eio.?11/sza2.1s,........anibuti of Neutrons Emerging from 13, R-10 Reactor Channels ? L. A. Trykov, 631 56 V. P. Semenov, and A. N. Nikolaev Track Detectors with an Extended Range of Measurements ? L. P. Roginets, 0. I. Yaroshevich, A. P. Malykhin, and I. V. Zhuk 636 60 ?-Detectors of the Radiation Typed Based on "Pure" Germanium ? V. K. Eremin, E. P. Dudnik, D. I. Levinzon, N. B. Strokan, N. I. Tisnek and O.P. Chikalova 638 62 Comparative Characteristics of Nal(T1) and CsI(T1) Detectors ? 0. P. Sobornov and 0. P. Shcheglov 640 63 Calculation of tremsstrahlung Spectra at Various Angles in the 1-30 MeV Range ? V. E. Zhuchko and Yu. M. Tsipenyuk 643 66 Monocrystalline Films of GaAs as Spectral Detectors of X-Rays and Soft y-Radiation ? V. M. Zaletin, I. I. Protasov, 0. A. Matveev, P. I. Skorokhodov, and A. Kb. Khusainov 646 68 Density, Surface Tension, and Viscosity of Uranium Trichloride?Sodium Chloride Melts ? V. N. Desyatnik, S. F. Katyshev, S. P. Raspopin, and Yu. F. Chervinskii 649 70 INFORMATION On the So-Called Cosmion ? N. A. Vlasov 652 73 INFORMATION: CONFERENCES AND MEETINGS Thirty-Seventh Session of the Academic Council of the Joint Institute of Nuclear Research (JINR) ? V. A. Biryukov 654 74 The European Conference on the Effect of Radiation on Materials for Fuel Element Cladding and Cores ? Yu. N. Sokurskii 659 77 Seminar on the Use of Thermal Nuclear Reactors in Ferrous Metallurgy ? E. F. Ratnikov 660 77 INFORMATION: NEW INSTRUMENTS AND APPARATUS Self-Contained Radioisotope Power Units for Navigation Equipment Systems ? Yu. B. Flekel', B. S. Sukov, and A. I. Ragozinskii 661 78 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 CONTENTS (continued) Engl./Russ. TOR-3 Reflecting Gamma Thickness Gage ? P. G. Lakhmanov, Yu. A. Skoblo, and V. B. Timofeev 663 79 BOOK REVIEWS S. M. Gorodinskii and D. S. Goltdshtein. Decontamination of Polymer Materials ? Reviewed by E. E. Finkel' 664 80 The Russian press date (podpisano k pechati) of this issue was 6/26/1975. 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 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 ARTICLES INVESTIGATION OF THE PHYSICAL CHARACTERISTICS OF THE REACTOR DURING STARTUP OF THE FIRST UNIT OF THE BILIBINSK NUCLEAR POWER STATION A. A. Vaimugin, V. V. Bondarenko, V. K. Goryunov, A. V. Gusev, B. G. Dubovskii, P. G. Dushin, A. N. Efeshin, L. D. Kirillovykh, M. Kistlt, V. I. Kozlov, 0 V. Komissarov, E. V. Koryagin, A. G. Kostromin, N. I. Lagosha, 1V1 A. Lyutov, M. E. Minashin, K. N. Mokhnatkin,A. P. Pantko, Yu. F. Taskaev, V. N. Sharapov, and A. I. Shtyfurko. UDC 621.039.524.2621.039.519 As already reported [1], the Bilibinsk Nuclear Power Station will consist of four units with reactors of the same type. In the period from 10 to 31 December, 1973, physical startup of the reactor of the first unit was effected* and on January 12, 1974 the Bilibinsk Nuclear Power Station produced electric current for the first time. During startup of the reactor of the first unit, detailed investigations were undertaken of the physical characteristics of the active zone in order to introduce, if required, any necessary changes in the loading of the reactors of subsequent units. The startup program, therefore, in addition to determining the characteristics of the reactor necessary for operation, provided for a number of other measurements, in particular, determination of the parameters of critical assemblies which was necessary for verification of the accuracy of the design procedures used in planning. *The second unit of the nuclear power station was brought on stream at the end of 1974. TABLE 1. Physical Characteristics of As- semblies Assembly Critical number of FC-3 Material parameter, m-2 experiment calcula- tion experiment calcula- tion I II III IV 38,3+0,2 55,2+0,2 50,8+0,2 63,5?0,2 39 55 46 58 6,5?0,6 4,7+0,5 4,4?0,5 ? 5,9 4,6 4,5 3,9 TABLE 2. Change of Reactivity on With- drawal of FC-3 with Water (ApFC) and with the Water Removed from It OpH2o) Coordinates ApFC.103 Assembly of cell* APH2o. 10' LII ?4,4?0,2 ?3,9?0,2 ?2,9?0,2 ?2,7?0,2 ?1,52?0,08 ?0,41?0,02 ?0,65?0,03 ?0,67?0,03 ?0,25?0,02 *Here and in future, the first two figures signify the number of the row and the next two figures signify the number of the cell in the row (see Fig. 1, a). Translated from Atomnaya Energiya, Vol. 39, No. 1, pp. 3-8. July, 1975. Original article sub- mitted September 13, 1974. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 571 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 N Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Number of row 21 20 19 18 17 16 15 14 13 12 11 10 09 08 07 06 05 04 03 02 01 1:1-> 0-2 ? -5 ? -4 o-5 A-6 P2111 Kir 01111CIMICIIIMMIC1111121100110 IIMMMIIMIIIIIIMMIIIIIIIIIIIII IIMDMICIIIIICIMC1111131110M11 111111111111111:111111MMMIIIII 121111B111:110111CIMEIMCJIMC MMIIIIIIIMMIIIIIIIMM1111111 121MCIIIICIMCIMCIIIICIMCIMEI IMEICIIIMMMEMIRMMIll CIMCIMCIIIDMOINCIIMOMO MMIIMMIIIIIMMIIIIIIIMMIM CIMILIMEMICIMCIMOMMIIIICI 111111111M111111121MIIMMIWIIIII IIMEIMCIIIICIWCIIIKIMallM MIIIIIIMMEMIIIIIIMMMIll P2MCIMIZIMCIMC111101111CW2 0802 03 04 05 06 07 08 09 0 1 2 13 415 16 17 18 19 20 21 Number of cell in row a 14 ? ? 15 ? 14 ? NM MM. 13 12' 11 10 13 12 ? 11 10 111111111MOM 11111?111MIIMIN ? ? 111?11111EMIll 09 09 os ? 08 MEM ? 07 ? ? ? 08 04 10 11 12 13 14 07 08 09 10 11 12 13 14 15 a 16 15 MA ME 15 14 1110111121M 14 0 ? 13 13 12 MCIMIZIMOM1211111 12 0 0 0 0 11 0 1111111111MMEM 11 10 111311113MOMOM 10 0 0 0 09 1111111111MMEMM 04 08 MCIMCIM 08 ? 0 0 ? 07 ME ME 07 06 07 08 09 10 .11 12 13 14 15 C 07 08 09 10 11 12 13 14 15 16 Fig. 1. Record chart of the reactor (a), transverse sections of a fuel channel (b) and a control and safety rod channel (c): 1, 2) Cells with fuel channels FC-3 and FC-3.3; 3, 4, 5) Automatic control rods (ACR), scram rods (SR) and manual control rods (MCR); 6) Neutron flux sensors; 7) Graphite brickwork;8) Fuel element;9) Steel tube; 10) Opening for control and safety rod. Fig. 2. Record charts of critical assemblies I-IV (a-d): 0) Cell with fuel channel FC-3; ?, ? , C) Channels for scram rods, ACR and MCR. TABLE 3. Change of Reactivity on With- drawal of the Control and Safety Rod Chan- nel with Water (ACS), and the Graphite Plug (NA) from cell 12-12 Assembly S ApC ? 103 Apg .103 III 2,8+0,2 ?0,74?0,04 IV 2,4+0,2 ?0,87?0,04 accordance with the characteristics Power station reactor. Measurements on critical assemblies were made for the purpose of investigating the physical prop- erties separately of the central and peripheral parts of the active zone. The difference between the prop- erties of these parts of the active zone is because the control and safety rod channels, which occupy separate cells, are located in the central part of the active zone and not in the peripheral section (Fig. 1). The following units were mounted in the center of the reactor for these measurements: assembly I repre- sented the control lattice (200 x 200 mm) of fuel channels filled with water and there were no control and safety rods in the active zone of the assembly; assembly II was similar to assembly!, but without water in The condition of the reactor and the emergency protection was monitored by means of a highly-sensitive startup equipment, the sensors of which-were located in the peripheral cells of the active zone. This equipment, with the presence in the active zone of a startup neutron source with a strength of ?107n/sec, enabled the neutron flux in the reactor to be monitored reliably from the instant of loading of the first fuel channels up to emergence at the power generating level. The effects of reactivity were measured by a "Pamir-M" analog reactimeter [21, whose kinetic simulator was made in of the effective groups of delayed neutrons of the Bilibinsk Nuclear 572 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 50 75 100 125 r, cm Fig. 3. Dependence of efficiency of water in the control and safety rod channels and the square of the neutron flux on the distance to the center of the assembly: CS *) Ap ; X) H20 Fig. 4. Dependence of change of reactivity Lp, by filling an opening for a fuel channel with graphite, on the distance to the center of the assembly:., ? ) Assemblies I and II. TABLE 4. Efficiency of Water in Control and Safety Rod Channels Position of rod in con- trol and safety rod channel Presence of waterinfuel channel Average change of reactivity on -- removal Fully withdrawn Yes No 6,1?0,3 8,0?0,4 Fully inserted Yes No 4,3+0,2 5,7?0,3 . the fuel channels; assembly III represented the fuel channel lattice, evacuated, with control and safety rod channels and fuel channels filled with water; assembly IV was similar to assembly III, but without water in the fuel channels. In order to form assemblies I and II, 12 regular control and safety rod channels were removed from the central part of the reactor and fuel channels were installed in their place. The entire assembly was loaded with fuel channels having a 3% uranium enrich- ment (FC-3). During the measurements on the assemblies, the minimum kinetic loadings, fuel channel efficiency and the efficiency of the control and safety rod channels, and the neutron flux distribution along the radius and height of the assemblies were measured. The measurement results are shown in Tables 1 to 3 and the record chart of the assemblies is plotted in Fig. 2. Comparison of the calculated and experimental data (see Table 1), shows their quite good agreement. The greatest difference (? 9%) in the critical number of channels is observed for assemblies III and IV. The effect of water in the cooling tubes of the control and safety rod channels on the reactivity of the assembly was determined. The data obtained (Fig. 3), show that the effectiveness of the water in the con- trol and safety rod channels (412S0), positioned at a different distance from the center of the active zone, to a first approximation is proportional to the square of the thermal neutron flux (41). This confirms that the effect of the water in the control and safety rod channels is due mainly to an increase of thermal neutron absorption. The overall decrease in reactivity, when the cooling tubes of the 12 central control and safety rod channels of assembly III are filled with water, amounted to 5.2 ? 10-3. In the experiments on the assemblies, the opening for the fuel channels, located outside the active zone, were not filled with graphite. In order to estimate the reduction of efficiency of the reflector due to the presence of these openings, the change of reactivity when certain openings were filled with graphite and located at a different distance from the boundary of the active zone (Fig. 4) was determined. The data showed that by filling all openings of the reflector with graphite, the reactivity is increased by 9.6 ? 10-3 and 7.6 ? 10-3 respectively for assemblies I and II. 573 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 N Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 TABLE 5, Efficiency of Fuel Channel and Control and Safety Rod Channels "8 a) 4-1 CtS Reactor change Measurable effect Change in reactivity, Ak/k. 104 cell 11-11 cell 11-14 cell 11-17 cell 11-19 cell 11-21 269 FC with FC-3 with water 7,3?0,4 7,4?0,4 5,5?0,3 6,1?0,3 5,8+0,3 water FC-3.3 with water 9,0?0,5 8,8?0,4 6,6?0,3 7,8?0,4 7,5?0,4 FC-3 filling with water 1,23?0,06 0,33?0,02 0,95?0,05 1,66?0,08 1,06+0,05 FC-3.3 filling with water 1,66?0,08 0,60?0,03 1,26+0,06 2,3?0,1 1,52?0,08 Removal of CSR channel with water ? 2,4?0,1 2,5?0,1 3,0?0,2 ? 269 FC FC-3 with water 10,0?0,5 10,2?0,5 5,6?0,3 4,6?0,2 4,2?0,2 without FC-3.3 with water 12,5+0,6 11,5?0,6 7,1?0,4 6,4?0,3 5,6?0,3 water FC-3 filling with water 2,5+0,1 0,40+0,02 1,39?0,07 2,2?0,1 1,13?0,06 FC-3.3 filling with water 3,3?0,2 1,00?0,05 1,72?0,08 2,9?0,2 1,46?0,07 Removal of CSRchannel with water ? 3,4?0,2 3,0?0,2 3,5?0,2 ? 1,00 0,95? go- 485- 460 0 20 40 . 60 . BO r, cm Fig, 5. Dependence of the neutron if on the distance to the manual control Measurements with the Full Reactor Charge, After carrying out the experiments on the assemblies, the reactor was loaded completely with fuel channels filled with water. A total of 217 FC-3 and 56 channels with 3.3%-enriched uranium (FC-3.3), which were installed in the peripheral cells of the active zone, were loaded into the reactor. The reactor was compensated at a minimally controlled power level by the total insertion of 40 manual control rods (out of 48) and 4 auto- matic control rods were located in the central position. In this state of the reactor, the efficiency of all the standard scram rods (8 rods) was Ak/k =1.3 . 10-2, which coincided satisfactorily with the design value of 1.24 . 10-2. On raising the scram rods and all the inserted manual control and auto- matic control rods, the subcriticality of the reactor was equal to Ak/k = -1.1 ? 10-2. The value obtained for the total reacti- vity reserve of the reactor (Ak/k = 0.11?0. 015) coincided with the calculated value. In the Bilibinsk Nuclear Power Station, reduction of the water density in the fuel channels leads to a drop in reactivity. It was determined in the experiments that the complete re- moval of water from all of the 273 fuel channels reduces the reactivity of the reactor by 3.1 ? 10-2. The corresponding calculated value is Ak/k = 2.75 ? 10-2. The experiment showed that water in the fuel channels at 30-50 cm distant from the boundary of the active zone and located in regions where there are no control and safety rods, has the greatest efficiency. The change of reactivity on removal of water from the fuel channels is due mainly to an increase of neutron leakage from the reactor. Analysis of the results of measurements of the efficiency of the rods in the case of a complete loading of the reactor with fuel channels without and with water, shows that the efficiency of the first case is greater by a factor of 1.4 approximately than in the second case. The removal of water from the tubes of the control and safety rod channels leads to a small increase of reactivity of the reactor. This effect was measured when the reactor was loaded with fuel channels without water, and fuel channels filled with water (Table 4). The data given show that the efficiency of water Ak/k in the control and safety rod channels depends on the position of the absorbing rods in these channels and the presence of water in the fuel channels. In the control and safety rod channels, the efficiency of the water on withdrawal of the rods, is higher by a factor of 1.4 approximately than in channels with inserted rods. During operation of the reactor at 100% power (62 MW), the average water density in the fuel channels is ? 0.6 g/c m3 and ? 0.9 g/c m3 in the tubes of the control and safety rod channels. As, at the start of the run, the number of control and safety rod channels with inserted rods is 30 (out of 60), the overall increase of reactivity as a result of the complete removal of water from the control and safety rod circuit in this case amounts to 3. 5 ? 104. 574 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 The increase of reactivity by the removal of water from the control and safety rod channels is caused mainly by a change of the multiplication properties of the active zone. The efficiency of the rods in this case is changed only very insignificantly. The effect of water in the control and safety rod channels on the efficiency of the rods was investigated for both complete loading of the reactor and by the critical assembly II. With complete loading of the re'actor, no effect of the water in the control and safety rod channels on the efficiency of the rods was detected, within the limits of measurement error (?5%). On assembly II, measurements were conducted in the. control and safety rod channels (12-06 and 12-04), located in the reflector at distances of 40-80 cm from the boundary of the active zone of the assembly. The measurements showed that, in this case, the presence of water in the tubes of the control and safety rod channels reduces the efficiency of the rod by 8-10%. The effects of reactivity on the setting up of the fuel channels (FC) and the control and safety rod channels were measured in cells of the eleventh row (see Fig. 1, a) for two states of the reactor: with the FC filled with water and without water (Table 5). The effects being considered are significantly dif- ferent for different cells which is due, in the first place, to the shape of the neutron field and to the presence of lattice inhomowneities created by the control and safety rod channels. On loading the reactor with fuel channels with water, the arrangement of fresh FC-3 (with water) increases the reactivity of the reactor on an average by 6.4 ? 10-4, and the arrangement of FC-3.3 by 7.9 ? 10-4. In this state of the reactor, the increase of reactivity on filling anFC-3.3 channel with water is greater by a factor of 1.4 approximately than on filling an FC-3 channel with it and the replacement of one manual control channel by an FC-3 channel increases the reactivity of the reactor by 9.2 ? 10-4 on an average. Replacement of FC-3 by FC-3.3 gives the least gain of reactivity in the cells located in the row with the control and safety rod channels and in the peripheral cells with two faces adjacent to the side of the reflector. The neutron flux over the height of the reactor and along the radius near the manual control rods and the empty cell was measured with miniature fission chambers when the reactor was loaded with fuel chan- nels without water. The measurements showed that when 8 absorbing rods are in the active zone in the central position (up to 4 scram rods and manual control rods), the neutron flux nonuniformity coefficient in the region of the active zone located around these rods and over the height Kz is equal to 1.50. Even when there is no rod in the intermediate position in the zone, Kz =1.34. These results confirm the design data concerning the significant increase of Kz in the presence of eight or more absorbing rods in the intermediate position in the zone. Measurements on the reactor showed also that the insertion of manual control rods into the zone reduces the neutron flux in the fuel channels located in series with the rod approximately 18% (Fig. 5). The presence of an empty cell leads to an increase of the neutron flux in the fuel channel adjacent to it by 7? 3%. The misalignment of the neutron flux in the fuel elements of the fuel channels located in a cell, which was situated between an empty cell and a cell with an inserted manual control rod, was determined experi- mentally. The maximum difference between the neutron fluxes in the fuel elements of this channel amoun- ted to 8 ?2%. Monitoring of the Energy Release in the Fuel Element Channels. A system for monitoring the heat release in the fuel channels is provided in the Bilibinsk Nuclear Power station, consisting of 24 rhodium sensors for the neutron flux [3]. The sensors are connected to an interpolating device and their currents are recorded by a multipoint pen-recording potentiometer. In addition to this, the energy release in the fuel element channels can be determined by measuring the efficiency of identical sections of the manual control rods as described in [4]. In these methods, the energy release in each fuel element channel is determined by linear interpolation of the neutron flux values at the measurement points and by the intro- duction of coefficients which take into account the change of the neutron flux in the vicinity of the command and control rods, empty cells and at points of installation of peripheral sensors, and also the uranium enrichment in the fuel element channels. In determining the energy release by the efficiency of sections of the manual control rod, a coefficient defining the reduction of the neutron flux at the periphery of the active zone also is used. In order to refine the values of these coefficients, the neutron flux was measured in all fuel element channels and at points of location of the sensors for monitoring the heat release, with a full reactor charge of fuel element channels without water, by fission chambers. On the basis of these mea- surements, values of the coefficients were chosen for which the values of the neutron flux in the fuel element channels obtained by linear interpolation, differed to the minimum degree from the values ottained by direct measurement in these fuel element channels. 575 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 For the coefficients chosen in this way, the mean-square error in determining the energy release in the fuel element channels by means of the regular system of monitoring the energy release amounts to ?79c, and by measuring the efficiency of sections of the manual control rods it is ?5%. The small error is due to the fact that the number of manual control rods is greater by a factor of two than those of the energy release monitoring sensors. With the combined use of both procedures, the error is reduced to ?4%: Measurements during startup showed that the principal physical characteristics of the reactor of the first unit of the Bilibinsk Nuclear Power Station correspond with the design values. The planned control and safety rods provide compensation of the reserve of reactivity equal to 119i and create the required subcriticality for the safe startup of the reactor. Therefore, the introduction of any corrections to the loading record chart of the active zone and to the reactivity compensation system is not required. By means of the manual control rods, the energy release field along the radius of the reactor can be smoothed. The coefficient of nonuniformity does not exceed the design value of 1.5. The reactor has a negative steam reactivity effect, which makes its operation stable and safe. In conclusion, the authors express their sincere thanks to all staff of the Bilibinsk Nuclear Power Station, participating in the preparation and execution of the physical startup. LITERATURE CITED 1. V. M. Abramov et al, Atomnaya Energiya, 35, No. 5, 299 (1973). 2. B. G. Dubovskii et al, Atomnaya Energiya, 36, No. 2, 104 (1974). 3. E. N. Babulevich et al, Atomnaya Energiya, 31, No. 5, 465 (1971). 4. I, Ya. Emeliyanov et al, Atomnaya gnergiya, 30, No. 5, 422 (1971). 576 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 BOOK REVIEWS V. G. Zolotukhin, L. R. Kimeli, A. I. Ksenofontov, et al. THE RADIATION FIELD FROM A POINT UNIDIRECTIONAL SOURCE OF GAMMA QUANTA* Reviewed by B. R. Bergellson In order to calculate the effects due to the interaction of y ?radiation with a substance, namely the quantity of energy released, dose intensity, instrument readings, biological effects, etc, it is necessary to have available comprehensive data on the space-energy and angular distribution of the radiation in every specific case. However, the problem of determining the y-radiation spectrum presents considerable com- putational difficulties. Even modern computers, in many cases cannot provide the required volume of calculations. An alternative for carrying out the cumbersome calculations is a procedure in which the required spectra can be set up by means of tabular data obtained for elementary sources. A point uni- directional source can be considered as the most elementary source, the radiation field of which has maximum information content. In the light of this approach, and also taking into account the applied nature of the problem, the monograph by V. G. Zolotukhin et al.should be considered; this is devoted to the com- putational-experimental investigation and tabulation of spectral data from point, unidirectional y -radiation sources in an infinite medium. The book consists of six chapters. In the first chapter, definitions are given of the principal characteristics of the sources and field and also methods of transforming the spectral distribution. The methodology for the experimental study of the energy and angular radiation spectra of a point unidirectional source is described in the second chapter. The authors pay particular attention to problems associated with the possibilities and special features of the use of the Monte Carlo method for calculating the transfer of y radiation to large distances from the source. These sections are written in a condensed form and, in contrast from others, postulate defined skills and knowledge of the subject by the reader. In the fourth chapter, the results are given of calculations of the differential characteristics of the radiation field of a point unidirectional source, and also data on the spectra, energy flux, absorbed energy and dose intensity for infinitely homogeneous media of H20, Al, Fe, Sn, W, Pb and U for source energies of 0.1 to 10 MeV. The present monograph is a unique publication in the diversity of the information given in the section on the characteristics of the y-radiation field of a point unidirectional source. In the sixth chapter, data are given which are essential for engineering calculations on building factors of scattered y radiation for anisotropic point sources. Unfortunately, the authors have confined themselves in this section to only a single medium ? water, although the necessary data for other media also were available. In conclusion, it should be emphasized that on the whole, the problem of finding the spectral distri- bution of scattered y radiation can be solved only by improvement of the appropriate algorithms and pro- grams for machine calculations. A knowledge of the entire problem for studying a unidirectional point source is scarcely feasible, in consequence of the diversity of the parameters of the source-medium- detector system, the complexity of presentation of the information in compact form and time consumption *Atomizdat, Moscow, 1975. Translated from Atomnaya Energiya, Vol. 39, No. 1, p. 8, July,1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 577 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 for reconstruction of the spectra. However, for the simplest case of point and surface anisotropic sources and a homogeneous infinite medium, this approach is competent and undoubtedly useful. The book reviewed will be used widely by engineers and scientific workers in their practical activities, 578 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 ARTICLES SOME PROBLEMS OF THE ECONOMICS OF A RESEARCH NUCLEAR REACTOR V. I. Zelenov, S. G. Karpechko, UDC 621.039.55:003.12 and A. D. Nikiforov Recently, more attention is being paid everywhere to problems of the economics of a research nuclear reactor and the planning of experiments on it [1-61. Procedures are considered in the papers for estimating the cost of an experiment on the reactor and certain efficiency indexes for its utilization. In determining the cost of the experiment, the total reactor costs are distributed between the experiments proportionally with the efficiency of the experimental apparatus [11, depending on the product of its volume and the neutron flux. This approach to the distribution of expenditure, in our opinion, is in need of refinement. For ex- ample: part of the total reactor costs, depending slightly on the power, is more suitable distributed equally between all experiments and the remaining part, which is directly dependent on the reactor power, is more suitably distributed proportionally with the efficiency of the experimental facility. In this paper, a procedure is proposed for assessing the cost of an experiment on a research nuclear reactor, taking account of this refinement. In addition, an attempt is made to explain the effect of the index of efficiency of utilization of a research reactor (in particular, the power utilization factor and the average operating power) on the cost of neutrons in the experimental facility. Procedure for Assessing the Cost of Experiments on a Research Nuclear Reactor We define the total costs on a research nuclear reactor in the following way: Etotal =Econst+ Evar where E is the total annual expenditure on the reactor; E total const is the constant component of the total (1) expenditure and Evar is the variable component of the total expenditure. We relate to the constant expenditure, those costs which depend only slightly on the reactor power : Econst= Earn+Esal Eint Eadrry (2) where Earn is the annual funding of amortization deductions on production buildings and power plant; Esai is the annual funding of working salaries of the staff; Et is the expenditure on the energy requirements for the intrinsic needs of the reactor and Eadm is the expenditure on salaries of the administration?manage- ment staff. We relate to the variable expenditure, the costs which depend directly on the power, in partic- ular the fuel costs : qCf ? Evar Nkpuf8760 , (3) where q is a coefficient which takes into account the consumption of fuel in the generation of one unit of thermal energy; Cf is the specific fuel costs in manufactured products; cp is the relative average burnup of the discharged fuel; N is the average operating power of the reactor and kpuf is the power utilization factor of the reactor, defined as the ratio of the time of operation of the reactor at any level of power to the calendar time. Translated from Atomnaya Energiya, Vol. 39, No. 1, pp. 9-11, July 1975. Original article sub- mitted August 8, 1974, ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, NY. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 579 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 ,1 42 43 44 45 46 47 48 49 PN/N Fig. 1. Variation of cost of neutrons in experimental facilities with in- crease of average reactor power. The cost of an experiment is determined by both the constant and variable components of the total expenditure on the reactor. As the constant component depends only slightly on the reactor power, then it should be distributed equally between the experiments without taking account of their individual charac- teristics (volume of the experimental facility, and the neutron flux in it). This distribution allows the minimum cost of an experiment in a given reactor to be obtained. We shall call it the support cost. The distribution of the variable component of the total costs between the experiments, which is proportional to their efficiency, allows individual special features of the experimental facility to be taken into account and the supplementary cost of the experiment to be obtained. We shall call this the physical cost. Thus, the total cost of the experiment will be determined by the support and physical costs. Taking account of what has been said, we write the total cost of a specific experiment on the reactor in the following form: Evar C e.. .___Econst _i_ (Its, n (4) where Ce. I. is the annual cost of the i-th experiment; 71,-1 is the unperturbed specific neutron flux in the cell of the active zone, intended for installation of the experimental apparatus [1]; Si is the working surface area of the experimental facility (immediately adjoining the active zone). Here, in place of the working volume in the expression for the efficiency of the experimental facility [1.1, the working surface is used, which equates the degree of effect of the physical (neutron flux) and geometric (diameter of the experimental facility) components of the efficiency, to the cost of the experiments. Cost of Neutrons in Experimental Facilities From the cost of an experiment in the reactor and the number of neutrons created by the reactor in the experimental facility during the year, it is easy to determine the cost of the neutrons for a given ex- perimental facility: _ Ce.i. _ (DiSiNkp1f8760 (5) where Cn.1. is the cost of neutrons in the i?th experimental facility. With calculation (4) Econst qCf _ (6) n N kpuf 87 MI ,S n ? E (DiSi The structure of the neutron cost is determined by the structure of the cost of the experiment. The cost of the neutrons also has two components: Econst (7) n. const nN kpuf 876001Si qCf 1 n ? (8) v. CD'S, 580 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 This definition of the neutron cost shows that, in experimental facilities of a different type (channel in a "catcher," in the heat release assembly, in the reflector and beyond the reflector),It is different. The constant component of the neutron cost in the i-th experimental facility depends on the utilization ef- ficiency index of the reactor (R, kpuf) and on the characteristics of the experimental facility Si) and ? the variable component depends on the productivity of the reactor (II Si, on the fuel cost and the depth of burnup of the discharged fuel. Let us consider the effect of the utilization factor of the reactor on the cost of the neutrons in experimental facilities. Simple mathematical transformations show that : ACjj.j, 2i Ln. Var TI 1+ AAT- E (-s? ) EconsttizSi (9) where 71-i Si is the efficiency averaged over the entire experimental facility; ACn. 1, /Cn.i. is the relative reduction of the neutron cost in the experimental facility for a relative increase of the average operating power of the reactor by AFT/FT with change of the other reactor characteristics. Figure 1 shows the dependence of ACn.i. /Cn. on AR/FT for various ratios of Evar/Econst with con- stant value of 41 Si/oli Si = 1.0. It follows from the figure that the maximum effect when the average reactor power is increased is achieved with small values of the ratio of these quantities. The absolute magnitude of the economic effect, when the average power of the reactor is increased, amounts to AN/A7 AC E -const 4/V ? 1 -' , (10) Consequently, the nature of the dependence of AC/Econst on AR/R. is identical for all reactors, and the economic effect is determined by the constant component of the reactor costs, other parameters being con- stant ( , Cf, cii, Si). It is obvious that an increase of the reactor utilization factor gives an economic effect which is determined by the expression: 41cpuf /kpuf AC =_Econst I Akpuf gr.puf ? The graph?of the dependence of AC/Econst on Akpuf/kpuf for all reactors will also be identical. Thus, in the case of costs distribution according to the proposed procedure, the concept of the sup- port cost of the experiment is introduced (defined by the constant component of the cost and the number of experimental facilities), which is the minimum cost of the experiment in the reactor. It is obvious that the cost of any experiment on a research reactor cannot be lower than the support cost, It has been shown that the ratio of the variable and constant cost components on the reactor, depends strongly on the re- duction of the cost of neutrons in the experimental facilities with an improvement of the utilization efficiency index of the reactor. In particular, it is most advantageous to increase the power characteristics of the reactor with respect to the variable and constant cost components within the limits of 0.1 to 0.5. The economic effect, when the utilization efficiency index of the reactor is changed, depends only on the constant component of the costs and is independent of the other reactor characteristics ((p , Cf Si). LITERATURE CITED 1. V. A. Tsykanov, Atomnaya t nergiya, 14, No. 5, 469 (1963). 2. A. S. Kochenov, Atomnaya Energiya, 21, No. 2, 97 (1966). 3. A. N. Erykalov and Yu. V. Petrov, Atomnaya Energiya, 25, No. 1, 52 (1968). 4. V. A. Tsykanov, Atomnaya Energiya, 31, No. 1, 15 (1971). 5. G. A. Bat', A. S. Kochenov, and L. P. Kabanov, Research Nuclear Reactors [in Russian], Atom- izdat, Moscow (1972). 581 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 6. V. V. Batov and Yu. I. Koryakin, Economics of Nuclear Power Generation [in Russian], Atomizdat, Moscow (1969). 582 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 BOOK REVIEWS A A. Vorob'ev, B. A. Kononov, and V. V. Evstigneev BETATRON ELECTRON BEAMS* Reviewed by P. S. Mikhalev Interest in accelerators as radiation sources for medicine, biology and industry continues to grow. One of the simplest and cheapest of the accelerators in operation over the range of energies 5-50 MeV is the betatron, which has been in series production for many years. The authors have attempted to eluci- date the buildup of experience in relation to the characteristics of electron beams during injection, accele- ration and, mainly, their extraction and utilization. There are five chapters in the book. The first two are devoted to the theory of the classical betatron with a time-variabie azimuthal-symmetrical magnetic field, and the dynamics of electrons on injection and during acceleration. In the third and fourth chapters, various methods are considered for extraction of the electron beam, problems of stabilization of the electron beam parameters and instruments used during operation with the beam. Problems of the practical application of betatronelectron beams are described in the fifth chapter. The authors have attempted to cover a quite wide circle of problems involving the motion of the elec- tron beam in the betatron. The principal value, in our opinion, is the experimental data assembled during the construction and operation of the betatrons in the Tomsk Polytechnical Institute and the engineering approach developed for the construction of beam extraction devices, The third and fourth chapters com- prise the basis of the book, However, these data are discussed very concisely and the theoretical part preceding it (chapters 1 to 3) is, in essence, an account of other well-known manuals on the theory of the betatron. It would be more valuable to give examples in more detail of the practical application of theory in approximate engineering calculations. The fourth chapter is overloaded with data which could be the subject of a separate and detailed consideration (this refers, first and foremost, to the sections on detec- tion and spectrometry, where only a brief listing is given of the methods and instruments for beam diag- nostics). Unfortunately, future improvements of the betatron beam characteristics are considered almost not at all in the book. It is true, the authors make an attempt to describe, from their point of view, future trends in this order: a betatron with a constant guiding field, a betatron with a spiral magnetic field, the application of superconductors, linear induction accelerators and a plasma betatron, Consideration of the prospects, in essence, reduces to the listing mentioned. However, a linear induction accelerator has been used for a long time in a number of investigations and, on the basis of the experience built up, judge- ment on the prospects for its utilization can be justified; an induction cyclic accelerator with a constant field, well-studied theoretically and on model experiments, permits the beam intensity to be increased by comparison with the normal betatron by an order of two to three, and now it should be possible to assess the prospects for its utilization. The other trends listed are in the experimental stage (plasma betatron) and the first theoretical proposals (spiral field), and no reference would be made in this book. Despite the shortcomings mentioned, the book can be useful to specialists in the development of beam extraction devices for betatrons of various applications. Moreover, the bibliography contained in the book *Atomizdat, Moscow, 1974. Translated from Atomnaya Energiya, Vol. 39, No.1, p. 11, July, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 583 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 (126 references) and the brief description of the experiments enables those so desiring to become acquainted in detail with the work carried out in this field. 584 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 ARTICLES SYNTHESIS OF A DIGITAL SYSTEM FOR CONTROL OF NEUTRON FLUX DISTRIBUTION E. V. Filipchuk, P. T. Potapenko, V. G. Dunaev, N. A. Kuznetsov, and V. V. Fedulov UDC 621.039.515 In recent years, data-transmission and data-handling systems based on "objective" computers are being used ever more extensively at nuclear power stations. The traditional path for computer use is the creation of a system of centralized control followed by its use as an "adviser" with the possibility of per- forming certain control functions in proportion to the accumulation of operating experience. It is consider- ed that one of the basic automatic control functions of such a system is control of the distribution of energy release in the core. ' We discuss several problems involved in the construction of a control system for a neutron field using an objective computer. Formulation of the Problem. For spatial control, we divide the reactor into m control zones with a detector and control rod in each zone. We introduce the local neutron fluxes vi and the settings (4.. The aim of control is minimization of the quadratic quality index [(p? (nT) ? w (nT)1T [cp? (nT)? (ID (nT)]. n=1 In contrast to [1, 21, we consider here fast motions of the system for a reactor with a stable power distribution. For a formulation of the problem, one can assume that the iodine and xenon concentrations are independent of time and the dynamics are satisfactorily described by a one-group diffusion equation. It is further assumed that the total power is stabilized by a supplementary high-speed automatic control system or by internal effects. A block diagram of the control system is shown in Fig. 1. Here, the Diare intrazone neutron flux detectors; K are commutators which provide a connection between the computer and each channel; the DMi are constant-velocity drive mechanisms. 11, r? D2 BM DM2 DMI Fig. 1. Block diagram of control system. Fig. 2. Structural diagram of pulse- height system. Translated from Atomnaya Energiya, Vol. 39, No. 1, pp. 12-16, July, 1975. Original article sub- mitted June 13, 1974. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, NY. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 585 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Kamp' 46 45 44 43 42 4015 4010 go 45 46 47 go oto 1,0 1,1 1,2 1,3 1,4 T, sec nT, sec Fig. 3. Dependence of threshold amplification factor of an open circuit on quantization time: ?) one channel; ---) two channels. Fig. 4. Transitional processes in the system parameters :T = 1.4 sec; rate of introduction of reactivity, 1.7?10-5 sec -1; static accuracy, 0.5k; correction coefficientb= 0(a), 1 (b), and 2 (c). A feature of this system is that the control computer, or machine for centralized control, is included in a closed control circuit, i.e., the controlled coordinates are quantized with respect to time and level. Introducing the appropriate lattice functions and considering that each of the m control channels is connected by commutators to the computer in a definite time period c = T/m after the (m-1) -th channel (quantization time T is identical for all channels), we obtain a structural scheme for this multi-dimensional relay, pulse-height system (Fig. 2). Here, W(q, c) is a matrix of transfer functions for the continuous por- tion shown, which joins detectors, drive motors, and object of control; x, f, and cl) are m-dimensional vec- tors for error, external effect, and characteristics of nonlinear elements. We assume that the digital control device, which is realized with a specialized or general-purpose computer, has a sufficiently large digital mesh and therefore quantization with respect to level in the com- puter can be neglected in the calculation. Stability. Since the range of possible perturbations is ordinarily rather large, there is particular interest in a study of the absolute stability of this system. We use theorems on absolute stability [31 for analysis of the stability of multiply-connected nonlinear pulsed systems. This method is similar to the well-known Popov method for the determination of the stab- ility of continuous nonlinear automatic systems and yields results which are associated with the concept of frequency response; what is more important, the method yields general sufficient conditions for stability which are applicable to systems of arbitrary order in this class. Note that the frequency criteria yield only sufficient conditions for absolute stability. However, these qualitative results are extremely useful in the initial design stage of the system. We consider an autonomous multiply-connected pulsed system (see Fig. 2). The system has m non- linear (in this case, relay) elements. 4)i (xi) is the output of the i-th nonlinear element; the input is the i-th component of the vector x[n, c1. We assume the nonlinear functions belong to the sector (0, K): (Di [0]=0; 0 0; (4) ei" 12 [Re (I + er4_? I a) WI` (j7, 0)] ? ei1? ?0,25 (e'?-1 1 a) (Fo, 8) + +11 + (I ?0,7) a] W` (? jw, V (0.o). Setting a. 0 (see [41), we obtain the simpler forms Re (Fo, 0)+-J->0; [Re WT (j(7), 0) + _7]2_ 0,25 I 147 e) e-i0+ + TV! (? jc,C. e) 12 > 0, V (0.. 0 and Tkj 0 the equilibrium state (3) of system (1) in the case of weakly connected cores (Oki 0) is "totally" asymptotically stable, if there exist for each of the subsystems in Eq. (4) an infinitely large Lyapunov function which is positively defined throughout all phase space in the form Vk = vik(nk) v2k (zik zmkk) and in addition the inequality 0 9vik/ank/nk < 6k -s 00 is satisfied. THEOREM 2. Let the function zkw satisfy correlqtion (2) and each subsystem in Eq. (4) be subject to the conditions in Theoreml, then for any arbitrary (Vk)(4) the evaluation Translated from Atomnaya Energiya, Vol. 39, No. 1, pp. 48-53, July, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 620 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Mh (Vk)(4) Wikk IZikk 123 Ik=1. M, will be true, in which yk and (pikk are positive numbers. If the inequalities yk/Ok > ck,k = 1,... M; M; j= k are satisfied, then for all akj > 0 and Tkj a 0 system (1) with nonlinear delayed connections is "totally" asymptotically stable. LITERATURE CITED 1. F. Bailey, SIAM Contr. , 3, 443 (1966). 2. N. A. Babkin and V. D. Goryachenko, in: Problems of Atomic Science and Technology, Series: Dy- namics of Nuclear Energy Reactors, Vol. 2, Izd. TsNllatominform, Moscow (1972), p. 67. 3. Y. Asahi, S. An, and A. Oyama, Nucl. Sci. Technol. , 4, No. 6, 49 (1967). Original article submitted March 4, 1974. FREQUENCY CRITERION FOR. THE STABILITY OF A CIRCULATING-FUEL REACTOR V. D. Goryachenko and V. V. Mikis.hev UDC 621.039.514 Frequency criteria for reactor stability have been obtained only for fixed-fuel reactors. We propose a frequency condition for the asymptotic stability of a circulating-fuel reactor. As initial equations we take the kinetic equations of a circulating-fuel reactor from [1] and the linear feed-back equations in integral form. Proper transformations reduce the initial system to a single nonlinear integro-differential equation dx ? (1 + f ? u) x (u) du 13L[E?x ? ki ? u) x (u) du] , v dt c (1) in which x is the relative deviation of the reactor power, T is the time measured in fractions of T*, the transit time of the fuel through the core, pi and j are the fraction and the importance [2] of the i-th group of delayed neutron emitters, p=1,1 pi, v = 1/pT*, 1 is the neutron lifetime, f(T) is the kernel for linear feedback which can be either lumped or distributed, and the k1(r) are the kernels generated by the equations for delayed neutron sources. The steady state of a circulating-fuel reactor is described by the solution x = 0 of Eq. (1). We denote by Ki(p) and F(p) the Laplace transforms of the kernels lq (r) and f (r). On the basis of the results of [2, 3] we prove the following. Suppose the function F(p) has no poles for Rep a 0 and F(0) > 0, and for all real values of co the inequality F(to) Re F > 0 v 2 K (P)1 (2) is satisfied. Then the zero solution of Eq. (1) is asymptotically stable for all initial conditions. This statement is analogous to the criterion obtained in [4] for fixed-fuel reactors and lumped linear feedback. 621 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 LITERATURE CITED 1. V. D. Goryachenko and E. F. Sabaev, Atomnaya Energiya, 23, No. 4, 295 (1967). 2. V. D. Goryachenko, Stability Theory Methods in the Dynamics of Nuclear Reactors [in Russian], Atomizdat, Moscow (1971). 3. V. D. Goryachenko, in: Problems of Atomic Science and Engineering, "Dynamics of nuclear power installations," No. 2 (6) [in Russian], TsNIIatominform, Moscow (1974), p. 75. 4. W. Baran and K. Meyer, Nucl. Sci. and Engng. , 24, No. 4, 356 (1966). Original article submitted April 3, 1974. ESTIMATION OF THE EFFECT OF PHYSICO-GEOMETRIC FACTORS ON THE DISTRIBUTION OF DELAYED FISSION NEUTRONS IN A BOREHOLE Yu. B. Davydov UDC 550.835 In order to determine the uranium content by delayed neutrons, the fission reaction of natural uran- ium nuclei under the action of primary neutron radiation is used [1, 2]. The purpose of this paper is to estimate the effectofphysico-geometric factors on the distribution of delayed fission neutrons in a borehole. A quantitative estimate of the effect of the hole diameter and the water saturation of the breeding medium on the distribution of fast and thermal delayed neutrons is obtained by a numerical method. The problem is solved concerning the distribution of delayed fission neutrons, induced by a point source of fast neutrons in a two-layered infinite medium with a cylindrical boundary of separation. The calculations are carried out for the case when the breeding medium is composed of porous rock of carbonate composition, the pores are filled completely with fresh water and the content of natural uran- ium in the rock is constant. The energy of the primary neutrons from the source is assumed equal to 14.1 MeV. The results of the calculation allow the following conclusions to be drawn: the flux of fast delayed neutrons decreases monotonically with increase of the hole diameter for probes of any length; the nature of the effect of the hole diameter on the magnitude of the flux of thermal delayed neutrons depends on the length of the probe. In the region of small probes of t 20 cm, an increase of the hole diameter causes a decrease of the thermal neutron flux. In the region of large probes, an increase of the hole diameter leads to the appearance of a local flux maximum of thermal delayed neutrons, which is attained when the depth of the water layer in the hole is equal to 2 to 3 cm. With further increase of the hole diameter, the buildup process is replaced by a process of absorption of thermal neutrons in the water and the flux de- creases. An increase of the moisture content of the breeding medium leads to a reduction of the flux of de- layed fast fission neutrons in the hole. The nature of the effect of water-saturation of the rock on the magnitude of the thermal delayed neutron flux depends on thebole diameter and the length of the probe. In the region of small probes, the magnitude of the thermal neutron flux decreases, for large diameters, with increase of the water-saturation of the medium and has a local maximum for small hole diameters when the moisture content reaches 10-20%. For large probes of / a 30 cm, an increase of the moisture content leads to a monotonic reduction of the magnitude of the thermal delayed neutron flux for any hole diameter. 622 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 LITERATURE CITED 1. S. Amiel and M. Peisakh, Atomnaya Energiya, 14, No. 6, 535 (1963). 2. Yu. B. Davydov, Izv. Vuzov. Gornyi Zhurn. , No. 6, 8 (1972). Original article submitted May 20, 1974. SPATIAL D.ISTRIBUTION OF FISSION NEUTRONS IN A BREEDING MEDIUM, CROSSED BY A DRILL HOLE Yu. B. Davydov UDC 550.835 The solution is considered of the problem concerning the spatial distribution of fission neutrons, in- duced by a point source of fast neutrons in an infinite homogeneous uraniferous medium, crossed by a drill hole [1-4]. The numerical calculation is carried out for the case when the breeding medium is composed of a dense rock of carbonate composition containing uranium ore of natural isotopic composition and with a density of the medium of 2.7 g/cm3. A source of primary neutron radiation is located in the hole, filled with fresh water ? a borehole generator of neutrons with energy equal to 14.1 MeV. The initial energy of the prompt fission neutrons is assumed to be 2 MeV and the neutron yield 2.5 n/event. In order to estimate the effect of the hole on the magnitude of the flux of fast 4>21k(r, z) and thermal 4)22k(r, z) fission neutrons, the results of the calculation are presented in units of magnitude of the fast 4321(0, 0) and thermal 4)22(0, 0) fission neutron fluxes in an infinite breeding medium, in the case of a negligibly small effect of the hole. The spatial distribution of the flux of fast and thermal fission neutrons is shown in Fig. 1. The re- sults of the calculation confirm that the fission neutron flux reaches a maximum magnitude in the breeding medium in regions subjected to the most intense irradiation. The moderation length of the fast neutrons exceeds the diffusion length of the thermal neutrons, and therefore data concerning the moderating proper- ties of the breeding medium are obtained from the more distant regions. Fig. 1. Spatial distribution of the flux of fast and ther- mal fission neutrons in units of the maximum flux mag- nitude in an infinite breeding medium: a) z) /.1321(0, 0); b) 4)22k(r, z)/422(0, 0). LITERATURE CITED 1. S. A. Igumnov, Izv. Vuzov. Gornyi Zhurn. , 2, 3 (1966). 2. Yu. B. Davydov, Izv. Vuzov. Gornyi Zhurn. , 6, 8 (1972). 3. Yu. B. Davydov and A. T. Markov, Atomnaya Energiya, 33, No. 1, 574 (1972). 4. J. Czubek, Report N. 732/PH, Cracow Institute of Nuclear Physics (1971). Original article submitted May 20, 1974. 623 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 VARIABLE MECHANICAL STRESSES, INDUCED IN THE FUEL ELEMENT CLADDINGS OF THE IBR-30 REACTOR BY POWER PULSES V. S. Dmitriev, L. S. Il'inskaya, G. N. Pogodaev, V. V. Podnebesnov, A. D. Rogov, V. T. Rudenko, and 0. A. Shatskaya UDC 621.039.55:621.039.526 During the development of power excursions, the active zones of pulsed fast reactors and boosters are subjected to the action of thermal shocks. This phenomenon, due to exceeding the rate of rise of tem- perature above the rate of expansion of the material, is accompanied by the stimulation in the fuel elements of alternating deformations and stresses, which are transmitted to the claddings and supporting structures of the core [1j. The action of the thermal shocks is aggravated by their high repetition frequency, which creates an accelerated wear of the active zone elements due to material fatigue. Damage of the fuel ele- ments might also occur when the tensile strength is exceeded, either during a single power pulse or from wave interference from the stresses of several pulses in the case of a too high repetition frequency. On the pulsed fast reactor (IBR) of the Joint Institute of Nuclear Research in Dubna, this phenomenon has been investigated over a number of years for the purpose of finding the optimum fuel element design and for determining the permissiblepulsed loading [2, 3]. - The paper describes the procedure and gives the results of measurements of the alternating me- chanical stresses which are induced by power pulses in the fuel element claddings of the IBR -30 reactor. The relative deformations were determined by means of high-temperature wire teuso-resistors with a base of 10 mm. Wire with a diameter of 30M made of NM23 x 10 alloy is used as the material for the tenso-sensitive lattice and VN-15T organo-silicon cement is used as the binding and insulating material. The principal measurements were conducted during operation of the IBR-30 in a cycle of widely-spaced pulses at a repetition frequency of 0.2 Hz and an average reactor power of up to 15 kW. Longitudinal and transverse oscillations were detected, with a frequency of ?5000 and ?1000 Hz respectively (Fig. 1). The amplitudes of the oscillations increased with increase of the pulse energy, and the time of damping did not exceed 10 msec. With power pulse energies of 2 ? 1015 fissions (average rise in temperature of the plutonium fuel elements of the active zone was 20?C per pulse), the maximum stresses in the cladding, created by the longitudinal and transverse oscillations amounted to 7.105 and 5. 105 N/m2 respectively. Fig. 1. Signals from the tenso-resistors installed on the cladding of a fuel ,element of the working fuel assembly; below: pulse power. The pulse energy was 2 ? 1015 fis- sions; pulse frequency 0.2 Hz. LITERATURE CITED 1. I. Randles and R. Laursma, EUR-3654-1 (1967). 2. V. T. Rudenko, Preprint OIYaI 13-764, Dubna (1971). 3. V. D. Anan'ev, OIYaI 13-4395, Dubna (1969). Original article submitted July 10, 1974. 624 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 CORRECTION OF THE GROUP CONSTANTS BY THE RESULTS OF EXPERIMENTS ON THE BFS CRITICAL ASSEMBLIES A. A. Van'kov and A. I. Voropaev UDC 621.039.519 The following problems are considered in the paper:correction of the group constants on the basis of integral experiments on critical assemblies (measurements of the ratio of the average cross-sections of different reactions and the ratio of the reactivity of samples at the center of the active zone for three BFS assemblies); determination of the constant error of the principal characteristics of a fast reactor of the OK-5 type (parameters ofcriticality and breeding, reactivity coefficients) before and after taking account of the Integral experiments; determination of the bias of the numerical values of these character- istics, due to taking account of the integral experiments. A statistical method has been used for the correction, in linear approximation, of the coefficients of sensitivity. Part of the results obtained is given in Table 1. The conclusion consists in, the following: the bias of the constants with a correction based on the integral data, depend significantly on the assumptions about the dispersions and also on the form and mag- nitude of the initial correlations for both the group constants and the integral data. In order to obtain physically plausible results of the correction, it is important to estimate correctly the error of the group constants and of the integral quantities associated with the approximations of the numerical model (re- quirement of adequacy of the conditions of the calculation and of the experiment). Taking account of cor- relations between the measured quantities is of considerable importance. The bias in the calculated reactor parameter is stable, if the choice of the integral quantities is sufficiently informative relative to this parameter. TABLE 1. Bias and Error Keff of the Total Coefficient of Breeding KB and of the active zone KBaz, the Doppler (DKR) and Sodium (NKR) Coefficients of Reactivity BFS-22 BFS-23* BFS-27t .0H-5 Keff KBaz Keff KBaz Keff Keff KB DKR NKR Go, % 1 5,5 3,7 9,2 4,1 4,2 6,2 80 60 all % 0,9 4,4 1,4 5,3 1,5 1,6 4,4 50 55 6, % 0,0 ?4,5 1,6 ?8,5 2;7 3,8 --6,0 35 15 *Model of fast reactor with uranium (BFS-22) and plutonium (BFS-23) fuel. t Assembly without U-238, with large dilution with graphite. *Breeder-reactor with oxide fuel, volume of active zone 5 m. Calculated values: KB = 1.39; DKR = ? 4.4.10-4 (Keff/AT). T? from 900 to 1500?K; NKR = 1.1% (Keff/Keff) with 50% of sodium removed from the reactor. 00 and oi are the errors, with and without taking account of the integral experiments respectively; .5 is the bias relative to the values calculated by the BNAB-70 system of constants; BFS = Fast Physics Assembly. Original article submitted August 23, 1974. THE INFLUENCE OF BEAM NOISE ON THE CRITICAL CURRENT OF LINEAR ELECTRON ACCELERATORS I. N. Mondrus UDC 621.384.64 This paper deals with the development of the transverse beam instability in a single section of a linear electron accelerator, taking into account the fluctuations in transverse displacement Y1 (s) and 625 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 Declassified and Approved For Release 2013/09/24: CIA-RDP10-02196R000400060001-4 transverse velocity Y2(s) of the center of mass of the beam bunch. At the section input the time-series Yi(s), Y2(s) constitute a normal stationary process and are described by the spectral density matrix f(co). The variance a2n and the mean An of the transverse displacement of the n-th bunch at the section output (Yn) can be expressed by the elements fik(w) of this matrix and the fundamental solutions of the transverse instability problem rii(s), rj2(s) by the relation [1] Y.= [Y1 (s) ? )72 (s) 112 (n s)] s=o The probability that the displacement of the last (n-th) bunch in the pulse does not exceed the section aperture a at critical current is given by P (lYni