Soviet Atomic Energy Vol. 41, No. 6

Body: Declassified and Approved For Release 2013/09/23: CIA-RDP10-02.196R000700080005-5 Russian Original Vol. 41, No. 6, December, 1976 SATEAZ 41(6) 1037-1136 (1976) ? 'ATOIVIHAR 3HEPhill (ATOMNAYA iNERGIYA) TRANSLATED FROM RUSSIAN CONSULTANTS BUREAU NEW YORK Declassified and Approved For Release 2013/09/23 : CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23 : CIA-RDP10-02196R000700080005-5 SOVIET ATOMIC ENERGY Soviet Atomic Energy is abstracted or in- dexed in Applied Mechanics Reviews, Chem- ical Abstracts, Engineering Index, INSPEC? Physics Abstracts and Electrical and Elec- tronics Abstracts, Current Contents, and Nuclear Science Abstracts. Soviet Atomic Energy is a cover-to-cover translations 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 Atomnaya Energiya: Editor: 0. D. Kazachkovskii Associate Editor: N,. A. Vlasov A. A. Bochvar N. A. Doll'ezhar V. S. Fursov N. Golovin V. F. Kalinin A. K. Krasin V. V. Matveev M. G. Meshcheryakov V. B..Shevchenko V. I. Smirnov. A. P. Zefirov Copyright ? 19/7 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. All rights reserved. No article contained herein 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. ? Consultants Bureau journals appear about six months after the publication of the original Russian issue. 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Single Issue: $50 Single Article: $7.50 CONSULTANTS BUREAU, NEW'YORK AND LONDON , 227 West 17th Street New York, New York 10011 - Published monthly. :Second-class postage'tiaid at Jamaica, New York 11431: Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya June, 1977 Volume 41, Number 6 December, 1976 CONTENTS Engl./Russ. ARTICLES An Investigation of Resonance Absorption of Neutrons in an RBMK-Type Grid - L. N. Yuroba, A. V. Bushuev, A. F. Kozhin, M. B. Egiazarov, and P. M. Kamanin 1 037 387 Two-Dimensional Kinetic Calculation of Nuclear Reactor by the Finite-Elements Method - N. V. Isaev, I. S. Slesarev, N. E. Gorbatov, and A. P. Ivanov 1 042 31 Start-up Tests on the Efficiency of the Biological Protection on Nuclear Power Stations Equipped with Water-Moderated Water-Cooled Power Reactors - A. S. Iz'yurov, A. S. Kuzhil', V. N. Mironov, A. I. Rymarenko, and S. G. Tsypin 1 046 395 An Experimental Study of the Way in Which the Internal Moderators of Annular Fuel Elements Affect Resonance Absorption in the Uranium - I. M. Kisir, V. F. Lyubchenko, I. P. Markelov, V. V. Orlov, V. V. Frolov, and V. N. Sharapov 1 051 399 Formation of Vacancy Micropores during Bombarding of Nickel by Similar Ions with Energy up to 300 keV - N. P. Agapova, I. N. Afrikanov, V. G. Vladimirov, V. M. Gusev, V. D. Onufriev, and V. S. Tsyplenkov 1 055 402 Effect of Reactor Radiation on the Susceptibility of Austenite Steel to Intercrystallite Corrosion - S. N. Votinov, Yu. I. Kazennov, V. L. Bogoyavlenskii, V. S. Belokopytov, E. A. Krylov, L. M. Klestova, and L. I. Reviznikov 1 058 405 Coherent Beam Instability in the IFVE Accelerator - V. I. Balbekov and K. F, Gertsev 1-061 408 DEPOSITED ARTICLES Multiorbit Induction Accelerators - A. A. Zvontsov, V. A. Kas'yanov, and V. L., Chakhlov 1 066 413 Pressure Change in a Vessel with Saturated Water on Being Unsealed -A. V. Alferov, V. V. Fisenko, and A. D. Shcherban' 1 067 413 Quantitative Estimates of the Energy of X-Ray Field Backscattered from Air - F. L. Gerchikov 1068 414 Computation of the Radiation Field of a Unidirectional Point Source of Fast Electrons by the Monte Carlo Method - A. V. Plyasheshnikov and A. M. Kol'chuzhkin 1 069 415 Determination of Neutron Spectrum from Measurements with a Small Number of Detectors - G. M. Obaturov and A. A. Tumanov 1 070 416 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 CONTENTS (continued) Engl./Russ. Floating Control of a Power Reactor with Respect to a Parameter with Time Lag - B. G. Ogloblin and K. N. Prikot 1071 416 1072 417 1073 418 1076 420 1078 422 1080 422 1083 425 1085 427 1087 428 1089 430 1091 431 1094 434 1095 435 1097 438 1100 440 1101 441 1102 442 1104 443 1106 444 LETTERS Sputtering of Metallic Surface by Fission Fragments - B. M. Aleksandrov, I. A. Baranov, N. V. Babadzhanyants, A. S. Krivokhatskii, and V. V. Obnoskii The Composition of the Radiolysis Products of the System CO2-H20(D20)-Oil Formed in a KS-150 Reactor - M. I. Ermolaev, A. K. Nesterova, and V. F. Kapitanov ...... . Determining the Thermal Power Outputs of Small High-Temperature Nuclear Power Plants - A. I. EPtsov, A. K. Zabavin, Yu. A. Kotelynikov, A. A. Labut, E. P. Lorin, I. P. Sviridenko, and Yu. L. Shirokovskii Influence of Irradiation on the Oxidation Kinetics of the Alloy Zr +2.5% Nb - M. G. Golovachev, V. V. Klyushin, and V. I. Perekhozhev Influence of Boron on Radiation Embrittlement of Low-Alloy Steel - V. A. Nikolaev and V. I. Badanin Measuiement of the Ratio o-f(239Pu)/o-f(235U) for Neutron Energies of 0.27-9.85 MeV - E. F. Fomushkin, G. F. Novoselov, Yu. I. Vinogradov, and V. V. Gavrilov Nuclear y Resonance Method for Investigating EI-69 Austenite Steel Irradiated with y Quanta or Fast Neutrons - I. M. V'yunnik, P. 0. Voznyuk, and V. N. Dubinin Effect of Temperature on the Porosity of Nickel Irradiated with Nickel Ions - S. Ya. Lebedev and S. D. Panin Numerical y-Ray Albedo from Limited Sections of the Surface of Reflecting Barriers - D. B. Pozdneev and M. A. Faddeev Yields of 200 201T1, T1, and 204T1 during Proton and Deuteron Irradiation of Mercury - P. P. Dmitriev, G. A. Molin, Z. P. Dmitrieva, and M. V. Panarin Albedo of a Cylindrical Rod - V. V. Orlov and V. S. Shulepin Operative Monitoring of Fission Products in Sodium Coolant of Fast Reactor - V. B. Ivanov, V. I. Polyakov, Yu. V. Chechetkin, and V. I. Shipilov CONFERENCES AND MEETINGS Regeneration of Fast-Reactor Fuel - A. F. Tsarenko Meeting of Four Nuclear Data Centers - V. N. Manokhin Meetings on the Compilation of Nuclear Data from Reactions with Charged Particles and Data on the Structure of the Atomic Nucleus - L. L. Sokolovskii IAEA Symposium on the Design and Equipment of "Hot" Laboratories - B. I. Ryabov Second Seminar on Computer Simulation of Radiation and Other Defects - Yu. V. Trushin International Conference on "Ion-Exchange Theory and Practice" - V. V. Yakshin Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 CONTENTS (continued) BOOK REVIEWS A. M. Petros'yants. From the Scientic Quest to the Atomic Industry. Contemporary Problems of Atomic Science and Engineering EngL/Russ. in the USSR ? Reviewed by Yu. I. Koryakin 1108 446 V. A. Zuev and V. I. Lomov. Plutonium Hexafluoride 1109 447 ? Reviewed by N. P. GW1iT INDEX Author Index, Volumes 40-41, 1976 1113 Tables of Contents, Volumes 40-41, 1976 1119 The Russian press date (podpisano k pechati) of this issue was 11/23/1976. 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 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 ARTICLES AN INVESTIGATION OF RESONANCE ABSORPTION OF NEUTRONS IN RBMK-TYPE GRID L. N. Yuroba, A. V. Bushuev, A. F. Kozhin, M. B. Egiazarov, and P. M. Kamanin UDC 539.125.5.173.162.3 The object of the present work is to obtain experimental data that would characterize the absorption of resonance neutrons in grids of RBMK type of the Leningrad Atomic Power Station. The computation of resonance absorption in such strongly hetrogeneous grids is a very complicated prob- lem. The difficulties are associated with the consideration of not only energy but also the spatial distribution of the flux of retarded neutrons. The presence of strong intrachannel retardation leads to a noticeable spatial inhomogenity of the field of resonance neutrons in the cell. One of the possible directions in searching for the solution of this problem is an experimental measurement of these dependences on the models of the active zone of a reactor. The variation of the concentration of hydrogen in a hetrogeneous water film may have an effect on the physical process in the intermediate thermal region of the neutron energy. Therefore, the integral para- meters p", (48)/ of, which are sensitive to this region of the neutron spectrum and also the effective res- onance absorption integral I2e5ff were measured. In the present work we describe the experiments carried out by two experimental groups on RBMK-type grids and present their results. The dimensions of the experimental assemblies, parameters of the fuel cas- settes, and the procedures of measurements somewhat differ but the results were close, which permitted us to arrive at consistent conclusions. The height of one of the experimental assemblies of 25 cells in an RBMK-type reactor was 2 m. Cas- settes of cylindrical form were placed in graphite stack with 25-cm steps. The fuel elements were prepared from aluminum tubes (13.5 x0.65 mm), and filled with tablets of natural uranium dioxide with a density of 10.2 g/cm3 and diameter 12.15 mm (see Fig. 1). A neutron beam from the horizontal experimental channel of the IRT-2000 reactor served as the source for the subcritical assembly. The beam was directed along the hollow channel into the depth of a 1-m-high graphite prism, which served as the base of the assembly and was used for forming spatial?energy distribu- tion of neutrons entering into the assembly. The axial distribution of neutrons in the assembly was measured by copper foils and Si-235U semiconduc- tor detectors with cadmium coating and without it. The upper and lower boundaries of the region of asymptotic spectrum of neutrons were determined and the location for subsequent experiments along the height was deter- mined. The cadmium ratios for reaction 238U (n, ,y) in cassettes placed at the center of the assembly in the second and the end rows were determined at this height. The values of cadmium ratios coincided within the errors of measurements (? 1.5%). This permitted the conclusion that the leakage of neutrons does not effect the results of measurements of Rd, p28, Cd, I? If I28 at the center of the assembly and the obtained values correspond to the parameters of an infinite array within the indicated errors. The parameters were measured with the use of indicators of about 1 mm in thickness prepared from standard UO2 tablets filling the fuel elements. During the experiments, the indicators were placed in detachable fuel elements, which were mounted in an experimental cassette in the place of the ordinary fuel elements. In the measurements of the rate of reaction 238U(n,7) for resonance neutrons, a cadmium screen of 0.5 mm thick- Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 387-391, December, 1976. Original article sub- mitted July 16, 1975; revision submitted April 30, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1037 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Fig. 1. The transverse cross sec- tion of the channel with cassette. TABLE 1. Values of the Parameters Oc- curring in Formulas (2) and (3). Parameter Value Source of information ivo b Kbl St S i (d ? 1mm) Sst (d = 0.1mm) Cpst /(Di f (E) 275+5 0,3150+0,0054 1,335+0,005 1,045T-0,004 0,993+0,003 (w/o water) 0,996+0,003 (with water) 9,983(w/o water) 0,995(with water) [51 131 Experiment [31 Experiment Computed by N. I. Belousov MIFI TABLE 2. Results of Experiments Medium fuel element Arrangement of fuel elements K K. "2 zelf, b x. Kei= .Tir: Kei (with water) Kei (w /o water) Graph- ite Single fuel element 7,50+0,22 20,6+0,7 1,0 ? Row 1 5,52+0,16 15,2+0,5 0,736+0,007 Row 2 4,51+0,13 12,4T0,4 0,601+0,014 ? Row 3 4,60+0,12 12,7+0,4 0,613+0,020 Air Value average for cassette 5,16+0,15 14,2+0,5 0,688+0,010 ? Row 1 5,65+0,13 15,5+0,5 0,753+0,028 1,02?0,04 Row 2 4,90-T0,13 13,5-T0,4 0,653T-0,026 1,09+0,04 Row 3 4,93T-0,12 13,5+0,4 0,657+0,025 1,07T0,04 Water Value average for cassette 5,37+0,13 14,8+0,4 0,716+0,027 1,04+0,04 ness was used. The results of [1, 21, in which means of decreasing the effect of cadmium on the results of measurement are indicated, were taken into consideration in choosing the scheme for arranging the indicator. The standard method was used for determining If. The desired quantity was obtained by comparing the rates of reaction in the indicator irradiated in the fuel elements of the cassette and in a uranium metallic foil standard irradiated in the retarder at the distance of 10 cm from the cassette where the spectrum of epicad- 1038 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 TABLE 3. Results of Measurements with the Second Assembly Medium around fuel'elf, element Arrangement of fuel elements ,28 28 'eff Peffst b Row 1 0,770+0,015 16,2+0,7 Row 2 ' 0,667T0,015 14,1+0,6 Row 3 0,648T-0,014 13,7T0,6 Air Av. value for cassette 0,731+0,015 15,4+0,7 Row 1 0,828+0,016 17,5+0,7 Row 2 0,717T0,014 15,1T0,7 '.ow 3 0,703T0,015 14,8T0,7 Water A,:. value for cassette 0,786+0,016 16,6+0,7 TABLE 4. Values of Some Parameters in Formula (7) Parameter Value Source of information (18th 6 2,72+0,02 151 o2s 6 fth, 582,2+1,3 [5] g25 0,976+0,002 [7] y28 v25 Ce ' Ce 0,816+0,010 [8] 628* 0,034+0,02(with water) Expt. 0,045+0 , 02 (w /o water) *Ratio of number of fissions in 238U to 235U. TABLE 5. Values of (1) / (45) and p28 Parameter Experi- mental system Without water With water logs), \GP/ Assembly in IRT (7,50+0,13)?10-3 (6,16+0,04)? 10-3 Assembly in F-1 (7,23+0,16)?10-3 (6,04+0,16)?10-3 228 Assembly 0,670+0,025 in IRT 0,441+0,009 1 mium neutrons is close to 1/E. The self-blocking coefficients for the standard t were determined experi- mentally in [3]. The method of determination is based on the measurements of cadmium ratios for the standard Rst and a thin uranium sample Ro: IPsit = (Ro? 1)/(Rst ?1). (1) The thin samples were prepared from aluminum foil of 0.1 mm thickness with a layer of natural uranium of 0.25 mg/cm2 thickness. The corrections for the self-blocking of thin samples and for the difference of the neutron spectrum in the retarder from 1/E spectrum were determined computationally. The measurements of the intensity of reaction 238U (n, y) are based on recording of y radiation of 239Np with an energy of 277 keV. A measuring system with a Ge(Li) detector was used for this purpose. The energy resolution of the system at 277 keV was 2.4 keV. Due to the unequal distribution of the radioactive nuclei in the sources, i.e., the indicators and the stan- dards, errors may arise due to the different efficiency of recording of y quanta emitted from different seg- ments of the source. In order to eliminate this effect an absorbing filter of variable thickness was placed be- tween the source and the detector so that the efficiency of recording of y quanta with an energy of 277 keV was equalized. From the measurements of the activity of irradiated foil we obtain ai/ a st, the ratio of activities of the indicator irradiated in one of the fuel elements of the cassette and the standard. The effective resonance inte- gral of the fuel element of the i-th row of the cassette was determined from the formula /eqf L.= MCI, (2) where e is the true resonance absorption integral of 238U; Ki are coefficients that take account of the self- blocking of the fuel element and the mutual screening of the fuel elements in the cassette, which were computed from the formula bl Si Ost Nst (E). Ki= Ks ast Sc7t (Di N, Here K131 is the coefficient of self-blocking of the standard; t'stAD i is the correction taking account of the dif- (3) ference of the neutron fluxes at the locations of the i-th fuel element and the standard caused by the macrodis- 1039 Declassified and Approved For Release 2013/09/23 : CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 tribution of neutron flux in the assembly; Nst/Ni is the ratio of the number of nuclei 238U in the standard and in the indicator; f(E) is the computed correction for the departure of the spectrum of epicadmium neutrons in the retarder from 1/E; Si and Sst are the correction for the self-absorption of 7 quanta with an energy of 277 keV in the indicator and the standard. The absorption coefficient of such y quanta in UO2 for the given geom- etry of the measurements was determined experimentally and was 0.60 mm-I. The correction coefficient for the self-absorption was computed from the formula given in [4]. The values of the parameters occurring in formulas (2) and (3) are given in Table 1. The values of 128 were determined for fuel elements of all the rows of the cassette, and In was deter- effi effo mined for single fuel element placed at the boundary of the cell. The ratios of gifo and 122 determine the co- efficient of resonance blocking of the capture cross section of 238U in the fuel; Ki and K0 are, respectively, the coefficients of mutual screening for the fuel elements in the fuel cassette. The effective resonance integral of the fuel cassette was determined from the formula E e ffi /Of 7 ni , (4) where ni is the number of fuel elements in the 1-th row of the cassette; 1tt 2,8?. is the effective resonance integral ei of the fuel element of the i-th row of the cassette. The results are given in Table 2. Another series of experiments were conducted on a subcritical assembly of 49 cells mounted in a wide neutron beam (150 x150 cm) of the F-1 reactor. The height of the assembly was 1.8 m. Cassettes with diam- eter of the fuel core of the fuel elements equal to 11.0 mm (casing 13.5 x1.0 mm) were investigated. The re- gion of the asymptotic spectrum was determined by axial and radial measurements with indicators made of 238u, 1.15in, 239pu9 and 235U in cadmium filters and without them. The experiments showed that within the assembly a region with 120 x120 x 100 cm dimensions has the asymptotic spectrum of neutrons that is charac- teristic for this grid. In these experiments the resonance integral was determined from measurement of the relative rate of absorption of epicadmium neutrons in the rods of the cassette and in a single rod placed in the retarder at the boundary of the cell. The value of the resonance integral for the single rod was calculated from the Hellstrand formula. Metallic foil of 10-fold depleted uranium of 0.09 mm thickness were used for the measurements. The rate of capture reaction in 238U was measured on a NaI(T1) spectrometer from the y ra- diation of 239U with an energy of 74 keV. In the analysis of the results corrections were introduced to take ac- count of the small background from the radiation of the fission products. The effective resonance integral for different fuel elements of the cassette was determined from a formula similar to (3): /28 eff a st /28 ? ast (Di effst (5) where Iafst =21.1 ? 0.8b for a rod of 11 mm diameter (computed from Hellstrand formula). The resonance integral of the cassette was determined from formula (4). The results are shown in Table 3. Thus, the following conclusions can be made; 1. The effective resonance integrals for the two investigated types of cassettes differ insignificantly. The somewhat larger value of the resonance integral in Table 3 is due to the smaller diameter of the uranium cores of the fuel elements. 2. The filling of the fuel channels by water has a weak effect on the effective resonance integral of the cassette. In the experiments for the determination of p28 and (o-/ (of) in the assembly on IRT reactor the above- mentioned indicators made of uranium dioxide and Ge(Li) spectrometer were used. The value of p28 was de- termined from the measurements of the cadmium ratio Rd for the reaction 28U (n, 7) in the fuel elements; p28 1/(1 - nd). (6) The determination of Bpd is based on the measurement of radiation of 239Np with an energy of 277 keV from the indicators irradiated in a cadmium screen and without cadmium. The irradiation was carried out in two diametrically opposite experimental fuel elements of the cassette. 104,6 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 The parameter ( crP)/ ( cri5) was determined from the measurements of -1/ radiation of 239Np with an energy of 277 keV and the decay products of 143Ce with an energy of 293 keV (the procedure is described in [61). (an 028 cth aNp ith (ar) 025 g25 aceC28 aaT.:KpR (7) where crPth and a-43h are the cross sections of radiation capture in 238U and in the fission of 235U for thermal neutrons; g25 is the Wescot factor; aNp and ac e are the radiation intensities of 239Np and143Ce indicators irradiated in the fuel element; akep and at are the intensities of emission of 239Np and 143Ce indicators irradiated in the thermal column; C28 ? 28 20 is a correction taking into consideration the contribution of the fission of 1+628 Yee/Yce 238U to the activity of 143Ce; Yee and lie are the yields of 143Ce in the fission of 238U and 235U. The indicators were calibrated in the thermal column of the F-1 reactor of the I. V. Kurchatov Atomic Energy Institute. The values of the parameters occurring in formula (6) are given in Table 4. The ratio (48)/ (43) was measured also in the assembly on reactor F-1. The rate of reaction of cap- ture in 238U was measured by depleted metallic uranium foil of 0.09 mm thickness; the y radiation of 239U with an energy of 74 keV was recorded with a NaI(T1) spectrometer. The background of the fission product under the peak of 74 keV was taken into consideration in accordance with [9] and it comprised not more than 2% in the fuel and the thermal column. The rate of fission reaction of 235U was measured by foils of a dispersion alloy of aluminum and uranium enriched to 90% in 235U (mass con- tent of uranium 17%) and having a thickness of 0.07 mm. The integral activity of the fission products was re- corded. The mean values of (gen)/ (op \ / for the cassette are shown in Table 5. The results of the measurements show that the filling of the thermal channels by water leads to a reduc- . tion of (o8)/(of) and specially of p28 due to softening of the neutron spectra. e f In conclusion, one should note the increased reliability of the results presented here since they were ob- tained on systems of different dimensions and with the use of different techniques and instruments. LITERATURE CITED 1. A. V. Bushuev, L. N. Yurova, At. Energ., 27, No. 4, 334 (1969). 2. Y. Hachya, J. Nucl. Science and Technology, 9, 629 (1973). 3. L. N. Yurova et al., At. Energ., 38, No. 4, 24-5- (1975). 4. D. Watt and D. Reamsden, High Sensitivity Counting Techniques. Pergamon Press (1964), p. 277. 5. BNL-325, 3 Ed. V. I. N. Y. (1973). 6. L. N. Yurova et al., At. Energ., 32, No. 5, 412 (1972). 7. S. I. Sukhoruchkin, Atomic Technology Abroad, No. 8, 34 (1976). 8. L. N. Yurova et al., At. Energ., 36, No. 1, 51 (1974). 9. L. N. Yurova et al., At. Energ., 31, No. 6, 628 (1971). 1041 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 TWO-DIMENSIONAL KINETIC CALCULATION OF NUCLEAR REACTOR BY THE FINITE-ELEMENTS METHOD N. V. Isaev, I. S. Slesarev, UDC 621.039.51.12 N. E. Gorbatov, and A. P. Ivanov In theoretical investigations of nuclear reactors, it is often necessary to calculate the neutron-physical characteristics of media consisting of heterogeneous regions of complex spatial configuration. Known calcula- tional models of this type include, in particular, two-dimensional systems comprising an irregular set of hex- agonal compartments with different physical properties. In recent years, methods and procedures for the calculation of nuclear reactors consisting of hexagonal compartments have been developed using both diffusional [1, 2] and kinetic [3] approximations. The use of a compartmental approach is extremely convenient for nuclear-reactor designers, since it allows the necessary information to be obtained for every compartment in the reactor. Natural refinements of the compartmental model of a reactor are to use a more complex function to represent the neutron flux inside the compartments (e.g.), by using series expansion with respect to some system of polynomials) or to adopt a more universal triangular grid, which allows the cells of the grid to be reduced in size. The present work proposes an approximate method of multigroup kinetic calculation of nuclear reactors using a regular triangular grid. The calculation scheme is based on the finite-elements method [4, 5]. The use of the kinetic equation allows the conditions at the outer boundary of the reactor to be accurately realized. In the present paper, in contrast to [5], the operators of the Boltzmann equation are taken in non-self-conjugate form, which is more convenient for computer realization. The use of a regular triangular grid leads to a sim- ple three-point scheme for the calculation of a reactor consisting of hexagonal compartments. The proposed method allows the solution to be found not only at the nodes of the calculation grid, as is usually the case with numerical methods, but at any point within the compartment, since the function that approximates the neutron flux inside the cell is known. The method is also suitable for precision calculations of nuclear reactors con- sisting of hexagonal compartments. Finite-Difference Grid We consider a two-dimensional Itknetic equation in the form [6] where 1 i 7)(114)(x, y, ,(1)(g)(x, y, (P)= K ff LO(g) (x, y, Ix, (ID) = 1 g +1211 =1/-1 ? 1112 [cos cp 76T? + sin cp Vt,Vti CD(g)(x, y, 2 p=1 ?1 0 G 2. X (x, y, VP )(x, y, T'); -01:1:"}(x, y, p., (p)=V?i) dp! d' vi x X (x, VP) (x, y, (V); /I =cos tu; tto=cos w cos w' +sin cc, sin tot cos(co ? go 9; F =1 , 2, ..., G is the number of the energy group; cu and (19 are directions of flight of the neutrons;7(g) (x, y), Zp (x, y) are the total macroscopic cross section and the ?tot (1) Translated from Atomnaya Energiya., Vol. 41, No. 6, pp. 391-395, December, 1976. Original article sub- mitted March 4, 1975; revision submitted December 26, 1975. This material is protected by copyright registered in the name of 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 $7.50. 1042 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Fig. 1 Fig. 1. Numeration of points in triangular cell. Fig. 2. Compartmental model of a BFS-16-10 assembly (1/6). Size of hexagonal compartment ',under key" 13.5 cm: 1) compensating absorber; 2-5) low-enrichment zones; 6) high-enrichment zone; 7) lateral shield. 44.1 OS% 4011 Akpaddit AAWAYA AYAYWAWAVAVVAYAYA Fig. 2 1,0 0,4 10 20 310 40 50 60 r,t cm Fig. 3. Radial distribution of heat emis- sion in the BFS-16-10 assembly. The dashed lines show the diffusional calcula- tion; the circles are experimental values; and the continuous curves correspond to the kinetic calculation; I) compensating absorber; II) low-enrichment zone; III) high-enrichment zone. TABLE 1. Macroconstants of Compartments of BFS-16-10 As- sembly Type of0) COMPart ment Itdt mi - i d ,,(020) i f v(2) -tot ? (2)):12) v 1 -. 2 1 0,1913 0,1575 -- 0,3359 0,2968 --- 0,0266 2 0,1814 0,1683 0,0061 .0,2952 0,2877 0,0077 0,00877 3 0,1818 0,1686 0,00608 0,2952 0,2877 0,0077 0,00887 4 0,1818 0,1686 0,00608 0,2950 0,2877 0,0077 0,00887 5 0,1811 0,1680 0,00613 0,2942 .0,2868 0,00764 0,0087 6 0,1759 '0,1621 T0,00843 0,2894 .0,2800 0,0118 A,00842 7 0,2126 0,1967 0,00255 0,3324 .0,3267 0,00327 - 0,0128 1043 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 fission cross section for neutrons of the g-th group. EN (x, y, ?0) is the macroscopic cross section for neutron transfer from the p-th to the g-th group; pco is the number of secondary neutrons formed in a single fission episode; x (g) is the fraction of the fission spectrum consisting of neutrons of the g-th group; Keff is the effective neutron multiplication factor. As the boundary condition for Eq. (1) we take the condition of zero input of neutrons to the reactor Vg)(x, y, p, (1))11" = --1 Q. We note in this connec- tion, that with a maximum pressure of 4.3-10-6 mm Hg an instability of the 9th harmonic of the radial oscilla- tions was observed sometimes; however, the increment in view of its small value could not be measured. Elec- tron instabilities seem to be of low probability, as the buildup of an appreciable number of electrons in the bunched beam with the stated parameters is impossible. In determining the dependence of the increment on the chromaticity, the latter was adjusted by changing the correction current in the square pole faces (Fig. 7). The current in the windings which focus and defocus the units was chosen so that the chromaticity of the radial betatron oscillations remained constant. With a change of current, the betatron frequency and the beam intensity were changed somewhat. The increment is re- duced to the average frequency Qz =9.825 by the formula (see Fig. 5) (10?Qz)/0.175. (4) Normalization with respect to intensity was carried out by formula (3). Figure 7 coincides completely with the previous results. Actually, the wall and ion instabilities are multireversible. For the oscillation mode of the bunch shown in Fig. 2 and formula (1), the dependence of the increment on the chromaticity should be deter- mined by the factor QQ' )2 --2 dQ 172 if 9)2 A sy_1? (? CP ? .!!'" ? (p ? v , 01' aP (5) where rrp2 is the mean-square phase length of the bunch. The first two terms of the expansion in series are written here, so that this formula can be used when A 0.5. It follows from formula (5) and Fig. 7 that 117p-2 0.4. Processing of the oscillogram in Fig. 2 gives 1 (17 2 0.5. Let us compare quantitatively the results obtained with the theoretical results. The increment of the vertical instability can be written in the form rpRAz (N? N) r e2gEz 172(4.0p n'VQz (k Q z) L 4nadb3 (72-1) az (ar+ az) (4T2 -)- 1) f? (6) The numerical values of the variable parameters refer to an energy of 0.97 GeV: r=e2/mc 2 =1.54 ? p 10-flcm is the classical radius of the proton; R =2.36- 104 cm is the average radius of the accelerator; N=2 ? 1012 is the beam intensity; Nthr =0.4.1012 is the threshold intensity; Az is a factor which takes account of the dependence of the increments on the chromaticity and is determined by formula (5) or from Fig. 7; -y =2.03 is the relativ- istic factor; Q=9.825 is the betatron frequency; k=10 is the number of the harmonic; c =3 ? 1010 cm/sec is the velocity of light; g=1.57 is the coefficient of linear expansion of the vacuum chamber due to corrugation; Ez = 0.85 is a factor which takes account of the ellipticity of the chamber cross section; o- =1.3 ? 1016 sec-1 is the conductivity of the chamber wall; d= 0.04 cm is the thickness of the chamber wall; b=5.75 cm is the half-height of the chamber; :7= 0.82 ? 105 mm Hg-1 ? sec-1 is the :lumber of ions formed by a single proton per second at a pressure of 1 mm Hg; P = 6.9 -10-7 mm Hg is the pressure; az =1 cm is the half-height of the beam; ar =2 cm is the half-width of the beam; wz =1.94 .105 sec-1 is the coherent oscillation frequency and T =5 -10-6 sec is the effective time of interaction of the ions with the proton beam. The first part of this formula, which takes into account the wall effect, differs from the well-known ex- pression for the wall instability increment [4, 5] by the factors g, Ez, and 6 /d, where the latter is introduced because the thickness of the skin layer. 6 is considerably greater than the wall thickness d. The effect of the ions is taken into account in the second part of formula (6) in accordance with [3]. It should be noted that the parameter T depends on the model used and the calculation is carried out with difficulty; the value taken is a quite rough estimate. When Az = 1, formula (6) gives Az =450 sec-1, and the wall contribution amounts to 510 sec-1; the ion con- tribution is 60 sec-1. The corresponding experimental values (see Figs. 6 and 7) are: Az =360 20 sec-1, wall contribution 400 sec-1 and the ion contribution is 40 sec-1, which are ,== 30% less than the calculated values. It is possible that the discrepancy is explained by the scatter of intensity of the bunches. It was mentioned above that scatter weakens the coupling of the bunches and leads to a reduction of the increments. In the calculation, all the bunches were assumed to be identical. The cause of the increase of the increments with increase of energy when W < 0.65 GeV (see Fig. 3) could not be established. Although with a low energy the damping effect of the ions increases, this is inadequate for explaining the observed relations. The chromaticity of the betatron oscillations in this region is small 1064 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 ???-? 4-5), so that its effect is almost absent. It is possible that here there is strong cubic nonlinearity of the magnetic field and a correspondingly higher threshold of instability, but the experimental verification of this hypothesis was not undertaken. Radial instability has much in common with vertical instability. Mainly the 10th harmonic is excited; the form of the radial oscillations of the bunch is the same as in Fig. 2; the increments depend approximately iden- tically on the energy (see Fig. 3). The radial increment increases with approach to complete resonance and decreases with increase of pressure of the residual gas (see Fig. 6). The experimental and calculated values of the increment differ by no more than a factor of two. However, the characteristics of the radial instability are considerably less stable than the vertical instability. This can be seen, for example, from Figs. 3 and 6. According to the first, the increment at 57 msec amounts to -100 sec-I, and according to the second, it is -200 sec-1, although the principal parameters of the accelerator were unchanged. Certain data confirm that the cause should be found first and foremost in the instability of the threshold. Thus, the cause of coherent instability in the IFVE accelerator is the interaction of the beam with the chamber walls, and the interaction with positive ions makes a relatively small stabilizing contribution. For vertical instability, the calculated and experimental data coincide with an accuracy of up to 30%. In respect of radial instability, the conclusions should be considered as preliminary, because the experimental data are in- complete, although there are no factors which contradict the supposition of a single excitation mechanism for both types of oscillations. The authors thank V. L. Ushkov for assistance given by him during the measurement of the dependence of the increment on the pressure. LITERATURE CITED 1. C. Pellegrini, Nuovo Cimento, 64-A , 447 (1969). 2. K. F. Gertsev et al., IFVE. Preprint, SKU 74-85 (1974). 3. G. Hereward, CERN MPS/Int., DL-64-8 (1964). 4. V. L Balbekov and A. A. Kolomenskii, At. Energ., 19, No. 2, 126 (1965). 5. L. Laslett, V. Neil, and A. Sessler, Rev. Scient. Instrum., 36, 436 (1965). 1065 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 DEPOSITED ARTICLES MULTIORBIT INDUCTION ACCELERATORS A. A. Zvontsov, V. A. Kaslyanov, UDC 621.384.6 and V. L. Chakhlov It is well known that the space charge of the beam limits the number of particles accelerated in a cycle. Their number can thus be increased by simultaneously accelerating several beams in the one emitter unit. Consequently, we must have several equilibrium orbits in the one emitter unit. We shall call accelerators of this type multiorbit accelerators. The equilibrium orbits in the form of concentric circles are situated in the one plane, while several such planes situated one above the other can be contained in a single unit. The field strength at orbits of radii r01, r02, roi must satisfy the conditions IT; (t)=21/z01 (t) for r =roi; frz (t)> 2115(0 for r < roi; 175 (t) < 2Hz (t) for r > rob 0 < n (ro,) 1.1 MeV). The specimens were screened from fission fragments and were weighed on a VLAO-100 analytical balance. ?During the tests (maximum duration 590 h) all the specimens became covered in a black film which ad- hered to the metal. The dependence of the oxidation kinetics of the alloy on the test conditions is shown in Fig. 1. The results of a mathematical analysis of the data are listed in Table 1. Translated from Atomnaya Energiya, Vol. 41, No. 6, p. 422, December, 1976. Original article submitted December 19, 1975; revision submitted May 7, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1078 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 50 40 20 255 100 200 300 400 Test duration, h Fig. 1. Oxidation kinetics of alloy Zr + 2.5% Nb in moist nitrogen, at various tem- peratures. a) No irradiation; b) under ir- radiation. MG ? 600 TABLE 1. Oxidation Kinetics Constants Oxidation conditions - Temp-. ?C Am = kt 1g T; ki (Am)n = ii2s k2 n 255 0,3 -- -- 310 -- 0,1 1,70 Underir- 330 -- 0,2 1,72 radiation 350 0,8 1,89 405 1,8 1,91 425 -- 3,5 1,96 290 0,7 -- -- 330 0,7 -- -- Noir- 350 1,0 -- -- radiation 400 -- 1,3 1,92 425 -- 2.2 1,96 Analysis of the data after the tests shows that irradiation of the alloy Zr +2.5 mass % Nb by a flux of neutrons in moist nitrogen increases the mass of the specimens by 20-50%. This effect may be due to the in- fluence of neutron irradiation on the structure of the oxide film and on the composition of the medium. 1079 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 INFLUENCE OF BORON ON RADIATION EMBRITTLEMENT OF LOW-ALLOY STEEL V. A. Nikolaev and V. I. Badanin UDC 621.039.531 Interest in the influence of boron on the properties of ferrite?pearlitic steel for the casings of water- moderated water-cooled power reactors has been stimulated by the following factors. In the first place, alloy- ing with 0.002-0.005% of boron increases the hardenability, and hence the strength, of the steel, without appre- ciably impairing its weldability [1]. Furthermore, information on the marked radiation embrittlement of steel containing boron [2] reveals the necessity of assessing the possible role of this element as an impurity in steel (,-10-3% or less). In this connection we attempted to study the radiation embrittlement of steel 15KUMFA withabout 0.004% of added boron; to elucidate the causes of the observed effects, we used boron of various isotopic compositions. Method. The investigations were performed on metal from a 100-kg batch induction-melted in the labo- ratory. When casting this into 16-kg ingots, a master alloy of Fe +10% B with a natural mixture of the isotopes 18B and 11B (18.4 and 18.6%), or with 95% enrichment in 18B or 11B, was added to the ladle. To prepare the master alloy with enriched boron, we mixed elementary boron with powdered iron car- bonyl; from the mixture we pressed tablets which were then sintered in vacuum. The melt was made with Armco iron as charge. The chemical composition of the steel is shown in Table 1. The ingots were drop forged and rolled to 10-mm sheet. The sheets were hardened from 960?C in oil and tempered at 690?C. After heat treatment the material had the structure of finely dispersed sorbite with a pri- mary austenitic grain size of 4-5. The mechanical properties were determined on impact specimens 5 x5 x27.5 ram in size with a V-shaped notch 1 mm deep and 0.25 mm in radius, and also on fivefold tensile specimens 3 mm in diameter. The liabil- ity to radiation embrittlement was estimated [3] from the difference between the critical brittleness tempera- tures Tc of the steel in its initial state and after neutron irradiation. The materials were irradiated in the core and reflector of a VVR-M reactor with flux density ratios of 1:1 for fast neutrons (E> 0.5 MeV) and 1:10 for thermal neutrons. The method was described in detail in [3]. Experimental Results. Figure 1 shows the dependence of the rise in Tc for the steel on the isotopic com- position of the boron and the fluence during irradiation in the core. After irradiation at about 50?C by a fluence of 5. 1018 neutrons/cm2, the change in Tc for the steel was the greater, the greater the enrichment with 18B. For the two extreme contents of 18B (about 5 and 95%), the shift in Tc differed twofold. A comparison with data on the influence of irradiation on the yield point (Table 2) reveals that for radiation hardening of steel with the same fluence there is no dependence on the isotopic composition of the boron. When the fluence is increased to 5 ? 1028 neutrons/cm2 Tc increases by 190-200?C in materials of all com- positions. From the results it follows that for steel with added 11B the graph of Tc vs fluence has the usual ratio ATc adAF/ 3 (where F is the fluence in units of 1018 neutrons/cm) for the value A ?24. For the remaining materials this relation does not obtain, showing that the saturation effect is attained earlier if 18B is present in the steel. When the irradiation temperature is raised to 300-350?C, the shift in Tc is much less, but in this case also the embrittlement increases with the 18B content (Fig. 1, curve 3). The influence of irradiation temriera- ture was investigated in more detail for steel with added natural boron. The results (Fig. 2) were compared with the data for steel of similar chemical composition ',Table 1) but containing not more than 0.0005% B. The Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 422-425, December, 1976. Original article sub- mitted January 6, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1080 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 TABLE 1. Chemical Compositions of Mate- rials Boron Qontent, Other elements, lc, mass le 0,004 "B ? 0,003 B 0,14 C; 1,72 Cr 0,005 1013 0,2 Ni; 0,81 Mo; 0,32 V; 0,2 Cu 0,0005 B 0,009 P TABLE 2. Influence of Irradiation by a Fluence of 5. 1019 neutrons/cm2 at 50?C on Yield Point of Steel* Boron content, mass `10 00.2, kgfimm2 before irradiation after irradiation 0,004 "13 0,003 0,005 10B *Tested at 20?C 75 75 76 96 95 97 AUC 80 40 JUG 120 80 40 0 100 200 300 400 0 100 200 300 400 Irradiation temp., 'C Annealing temp., 't Fig. 2 Fig. 3 Fig. 1. Dependence of rise in Tc of steel on isotope composition of boron after irradiation. 1) 5 ? 1029 neutrons/cm2, 50?C; 2) 5. 1019 neutrons/cm2, 50?C; 3) ?1- 1029 neutrons/cm2, 300-350?C. Fig. 2. Rise in Tc for steel with 0.003% boron (1) or no boron (2) vs irradiation temperature in core [fluence (0.5-1) ? 1029 neutrons/cm21. Fig. 3. Influence of annealing temperature on restoration of Tc for steel containing 19B after ir- radiation with a fluence of thermal neutrons of 6.8.1029 neutrons/cm2 at 50?C. extra embrittlement due to the boron falls rather rapidly as the irradiation temperature rises, especially be- tween 150 and 250?C. This relation does not differ appreciably from that observed in steels with no boron ad- ded [3, 5]. To elucidate the influence of boron on the embrittlement we studied the possibility of restoring Tc of ir- radiated steel by postirradiation annealing. Experiments were performed on steel with added 19B; to increase the contribution of this isotope to the embrittlement we irradiated the specimens mainly with thermal neutrons (in the reflector). Fig. 3 shows that after annealing for 1 h Tc is partially restored even at 200?C, while at 300?C it reaches 85-90%. The temperature range of restoration of Tc for steel with boron is typical of most steels of the fer- ritic?pearlitic class [4, 5]. , The data show that the influence of boron merely intensifies the normal embrittlement, the external mani- festations of which are well known. 1081 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Fig. 4. Microstructure (a) and autoradio- gram of boron distribution (b) of steel 15Kh2MFA. Increase of embrittlement under the influence of boron probably occurs as a result of formation of frag- ment nuclei with high kinetic energies (about 1.8 or 0.6 MeV) by the reaction 10B(n, a)7Li. Their retardation in the matrix forms additional defects in the lattice. Owing to the large reaction cross section, the 10B is rapidly burnt up (for a fluence of thermal neutrons of 5 ? 1020 neutrons/cm2 the burn-up is about 90%). Therefore, its influence is manifested at comparatively low doses and is hardly appreciable in the conditions for the satura- tion effect. Radiation strengthening of steel does not depend on the 10B content. This is evidently because of the na- ture of the distribution of boron, which can be seen by track autoradiography [6]. To study the distribution of boron* we used steel specimens with added "B. To obtain a coarse grain the specimens were heat treated by heating under hardening to 1200?C, followed by tempering, as in the other cases at 690?C. The resulting autoradiograms when compared with the microstructure (Fig. 4) showed that the den- sity-of tracks is very nonuniform and is maximal at the boundaries of the former austenitic grains, which are thus also sites, of preferential concentration of boron. The distribution of tracks left by alpha particles in the autoradiograms simultaneously reflects the distribution of zones in which retardation of alpha particles and lithium ions leads to an increase in the number of atomic displacements. Naturally, a macroscopic character- istic such as the yield point, which is determined by the deformation resistance of the whole bulk of the metal, cannot be influenced by such a local increase in defect concentration. Therefore the yield point depends almost entirely on the uniformly distributed defects due to fast neutrons. At the same time, the most damaged zones in the crystal lattice, adjoining grain boundaries, can apparently serve as sites of preferential formation of crack nuclei, and consequently the appearance of such sites may influence the breaking strength of the steel. Thus the role of boron in embrittlement of 'steel is essentially equivalent to an increase in fluence. It is known that Te increases practically linearly with the fluence as the Ni, Cu, or P content of the steel increases [3]. Therefore we can expect that the presence of boron should be more markedly manifested, the higher the *The track autoradiography experiments were performed by N. V. Mishina and N. B. Odintsov. 1082 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 content of these elements in the steel. This means that when steel is alloyed, for example, with nickel, it may be necessary to limit and control the boron impurity content in order to ensure high radiation stability. LITERATURE CITED 1. C. Cottrell, J. Brit. Weld., 1, 315 (1954). 2. H. Myers and M. Grounes, in: Proc. of Third Intern. Conf., Geneva, 1964, A/conf. 28/P/420. 3. V. A. Nikolaev and V. I. Badanin, At. Energ., 37, No. 6, 491 (1974). 4. A. D. Amaev et al., Fourth Geneva Conference, Report No. 705 [in Russian], (1971). 5. C. Serpan and J. Hawthorne, Trans. ASME. J. Basic Engng., 89, No. 4, 877 (1967). 6. R. Fleischer, P. Price, and R. Walker, Science, 149, No. 3682, 383 (1965). MEASUREMENT OF THE RATIO 7f(239Pu)/crf(235U) FOR NEUTRON ENERGIES OF 0.27-9.85 MeV E. F. Fomushkin, G. F. Novoselov, UDC 539.173.4:621.039.9 Yu. I. Vinogradov, and V. V. Gavrilov The energy dependence of the ratio of the fission cross sections of 239PU and 25U for neutrons has been studied by many investigators by various methods. However, at present there is a rather large disagreement between the results. Even in the appraised data in the compilations of Davey [1], Byer [2], Greene et al. [3], Sowerby et al. [4], and Kontshin et al. [5], it reaches 7-8%. The curves characterizing the ratio differ in shape, especially at neutron energies of about 1-10 MeV. To refine our ideas of the structure of the graph of af(239Pu)/ 7f(235U), it is advisable to carry out investigations by the same method over a wide range of neutron energies. The preliminary results of measurements of the ratio of the fission cross sections of 239Pu and 235U, given here, were obtained by the time-of-flight method with an underground nuclear explosion as the pulsed neutron source. Similar measurements of comparatively well-studied characteristics also enable us to assess the feasibility of using the method in other nuclear physics investigations. The method of measurement was described in [6]; we shall merely mention its essential features. As the fission-fragment detector we used a film of polycarbonate with a molecular mass of 90,000. Time-of-flight scanning was effected by an electromechanical apparatus: One of its units was a drum with the polycarbonate film cemented on, rotating at about 104 rpm at the moment of the neutron pulse. Layers of fissile isotopes were arranged near the film in the neutron flux; between each layer and the film there was a slit-type fission- fragment collimator. With this method of measurement, the time resolution is governed by the rate of rotation of the drum, the flight distance, and the width of the fragment collimator slit, At/L-=-- Ax/5L, where L is the flight distance, v is the linear velocity of the film relative to the collimator, and x is the col- limator slit width. In our measurements, the time resolution was 7.5 nsec/m (total width by half height). The influence of scattered neutrons outside the direct flux was accounted for by additional collimators with layers of 235U and 239pu. Calibration of the layers, i.e., measurement of the ratio of the effective numbers of fissile nuclei in the layers (with allowance for the probability of passage of fragments through the collimator) was effected by Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 425-426, December, 1976. Original article sub- mitted January 21, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1083 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 7,25 NZ' 1,00 0,75 - 450 01,7 0,5 5 10 EmMe V Fig. 1. Energy dependence of ratio of fis- sion cross sections of 239Pu. and 235U. Present authors' data; - - - -) estimate of Sowerby et al. [4]. TABLE 1. Ratio af(239Pu)/crf(235U) vs Neu- tron Energy > > 0 .,s,' `,1.. .--. b --, > ') >m . tj ---- Error, 010 cud e 11 d d z- - b 0,27 0,36 0,81 14,4 1,42 1,62 1,39 5,1 0,36 0,44 1,07 12,1 1,62 1,85 1,63 4,9 0,44 0,50 1,15 10,9 1,85 2,15 1,47 4,6 0,50 0,59 1,31 8,4 2,15 2,50 1,56 4,4 0,59 0,69 1,34 7,3 2,50 3,00 1,57 4,6 0,69 0,83 1,37 6,3 3,00 3,60 1,62 4,7 0,83 1,01 1,55 5,5 3,60 4,45 1,67 4,9 1,01 1,13 1,33 6,6 4,45 5,60 1,37 5,1 1,13 1,26 1,39 6,1 5,60 7,28 1;25 4,8 1,26 1,42 1,24 5,4 7,28 9,85 1,23 6,2 means of fission by thermal neutrons. The experimental unit, including the layers of 239Pu. and 235U, the col- limator, and a small piece of detector film, was installed in a graphite prism 100 x100 X 100 cm in size. This prism was irradiated by fast neutrons from a 235U critical assembly [7]. At the site of the test specimen, the temperature of the thermal neutrons was 309 ? 18?K. The fission cross sections for 0.0253 eV neutrons were 742.5 ? 3.0 b for 239Pu and 582.2 ? 1.3 b for 235U [8]. The g factors were, respectively, 1.066 ? 0.014 [9] and 0.973 ? 0.003 [10]. The polycarbonate detector films were processed in identical conditions with 6.25 N NaOH solution. The films were examined and the tracks counted in classes with an optical microscope. The results showed that the total background of scattered neutrons in the range of times of flight under examination did not exceed 1.5%. We made a correction for the contents of 239PU, 235U, and other fissile isotopes in the specimens. Since the prob- ability of passage of fragments through the collimator depends on the angular distribution, we also made a cor- 1084 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 rection (about 1%) for the difference between the energy dependences of the angular anisotropy of fission of the nuclei under investigation. The results of the measurement of the ratio crf(238Pu)/o-f (235U) for 20 neutron en- ergy ranges (En.min and En. ) are shown in Fig. 1 and listed in Table 1. The quoted statistical errors cor- respond to a confidence probability of 68% and were determined from the number of tracks due to fission frag- ments in the corresponding intervals and the corrections. The calibration error (? 1.8%) was due to the error of measurement for thermal neutrons and the error in the values used for the g factors and the fission cross section for 0.0253 eV neutrons. For comparison, Fig. 1 shows the curve recommended by Sowerby et al. [4]. We see that the results ob- tained by the authors in the neutron energy range 0.5 -_En -4.5 MeV agree with the curve of the cross-section ratio. The m arked c' viation from the recommended curve at higher energies may be due to the inadequate energy resolution. However, there are certain results, e.g., those of Savin et al. [11], which agree with this curve at high neutron energies (En> 4.5 MeV) as well. The discrepancy at low energies is apparently due to background sources which were not taken into account. Improvements to the method and further research will improve the accuracy and reliability of measure- ments of the fission characteristics of heavy nuclei. LITERATURE CITED 1. W. Davey, Nucl. Sci. Engng., 32, 35 (1968). 2. J. Byer, Atom. Energy Rev., 10, No. 4, 529 (1972). 3. N. Greene, J. Lucins, and C. Craven, Rep. ORNL-TM-2797 (1970). 4. M. Sowerby, B. Patric, and D. Mather, Ann. Nucl. Sci. Eng., 1, 409 (1974). 5. V. A. Kontshin et al., in: Nuclear Constants, [in Russian], No. 16, Atomizdat, Moscow (1974), p. 329. 6. E. F. Fomushkin et al., At. Energ., 39, No. 4, 295 (1975). 7. Yu. M. Odintsov, A. S. Koshelev, and A. A. Malinkin, At. Energ., 38, No. 4, 209 (1975). 8. S. Mughabhab and D. Garber, Neutron Cross Sections, Vol. 1, Resonance Parameters, BNL-325 (1973). 9. C. Wagemans and A. Deruytter, Ann. Nucl. Sci. Eng., 2, 25 (1975). 10. G. Hanna et al., Atom. Energy Rey., 7, No. 4, 3 (1969). 11. M. V. Savin et al., At. Energ., 29, No. 218 (1970). NUCLEAR 7 RESONANCE METHOD FOR INVESTIGATING EI-69 AUSTENITE STEEL IRRADIATED WITH y QUANTA OR FAST NEUTRONS I. M. VIyunnik, P. 0. Voznyuk, UDC 548-162 :539.16.04 and V. N. Dubinin As is known, radiation damage produced by y and neutron irradiation affects the decomposition of aus- tenite in the annealing of EI-69 steel [1, 2]. It was of interest to investigate more thoroughly the state of aus- tenite in hardened steel immediately after irradiation. For this purpose, irradiated specimens were investi, gated by using the nuclear y resonance (NGR) method (1VIOssbauer effect), which provides additional informa- tion on radiation defects. Hardened polycrystalline specimens of EI-69 steel (0.42% C, 13.35% Cr, 13.68% Ni, 2.08% W, 0.33% Mo and 70.14% Fe) with a uniform composition (carbon-supersaturated austenite) were irradiated at room tempera- ture with 1.2 MeV y quanta from radioactive cobalt 88Co (irradiation dose, 8.8. 1018 quanta/cm2) and fast neu- trons from the VVR-M reactor at 60?C (mean fluence, 3.5 ? 1018 neutrons/cm2). Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 427-428, December, 1976. Original article sub- mitted January 21, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1085 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Number of pulses Source velocity. Fig. 1. Absorption spectra of iron atoms. a) In a steel matrix; b, c) in steel irradiated with quanta and fast neutrons, respec- tively: d) in steel irradiated with fast neutrons and annealed at 500?C over a period of 30 min. The specimens for investigations based on the NGR method were prepared by electrolytic dissolution of sheets of irradiated and nonirradiated steel. We obtained deposits with high concentrations of the carbide phase and other nonmetallic high-dispersion phases, which cannot be observed by means of an x-ray diffracto- meter. The thickness of the specimens (absorbers) was equal to 2.10-4 g/cm2 with respect to 57Fe. The pre- pared specimens were investigated in an electrodynamic 1Vlossbauer spectrometer operating under constant- acceleration conditions. The gamma radiation source was 57Co in a chromium matrix. The source and the ab- sorber were kept a-, room temperature. The results of investigations based on the NGR method are shown in Fig. 1. The velocity of the radiation source relative to the absorber is laid off on the axis of abscissas, while the number of pulses per analyzer channel is laid off on the axis of ordinates. The spectra were processed by means of a computer, using the method of least squares. It was assumed that the lines had the Lorentz form. The spectrum of hardened steel contains the usual austenite line with the width I' =0.53 0.03 mm/sec, shifted by a =0.21 ? 0.03 mm/sec with respect to the line of sodium nitroprusside in Fig. la. Only negligible line broadening, r =0.48 ? 0.03 mm/sec, occurs apparently as a result of y irradiation, which is seen in Fig. 1086 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 lb. In steel irradiated with fast neutrons, the NGR spectrum consists of the asymmetric line (Fig. lc) resul- ting from the superposition of two lines (dashed curves): the singlet of the steel matrix with the width r= 0.52 ? 0.03 ram/sec and (5 =0.21 ? 0.03 mm/sec and a line with the width r =0.76 ? 0.05 mm/sec, the center of which is shifted by (5 =0.62 ? 0.05 mm/sec with respect to the line of sodium nitroprusside. Radiation defects develop in hardened EI-69 steel as a result of irradiation; most iron atoms are located in normal regions of the crystal lattice that are not distorted by defects, so that the singlet of the steel matrix is observed in experiments (Fig. 1 b and c). The broadening of this singlet may be connected with the concen- tration nonuniformity caused by radiation defects in irradiated austenite. The appearance of the second, greatly broadened line (c) indicates that some iron atoms enter the new phase formed under the action of neutron ra- diation. A much larger number of defects appears in hardened EI-69 steel irradiated with fast neutrons than in steel irradiated with 7 quanta (in our case, 5.1020 and 1.2. 1017, respectively). For instance, such defects may consist of vacancy clusters [3], which promote the development of precipitation particles in hardened steel. At the initial stage of irradiation, an iron atom, bound to two carbon atoms, constitutes the basic precipitation nu- cleus [4]. It can be assumed that this precipitation also takes place in hardened EI-69 steel under the action of neutron radiation. Then, the line with the isomeric shift (5 =0.62 ? 0.05 mm/sec may be caused by the meta- stable carbide phase MemCn, the isomeric shift of which is close to Fe5C2 [5]. As a result of annealing at 500?C over a period of 30 min, the metastable carbide phase in EI-69 steel ir- radiated with fast neutrons passes into the stable phase Me23C 6, which is indicated by the appearance of lines with the isomeric shift (5 =0.96 ? 0.03 mm/sec (Fig. 1d) in the spectra [6]. Precipitations of the carbide phase Me23C were not observed in unirradiated specimens at this annealing temperature. LITERATURE CITED 1. I. M. Vtyunnik, I. D. Konozenko, and M. P. Krulikovskaya, At. Energ., 37, No. 3, 245 (1974). 2. I. M. Viyunnik and M. P. Krulikovskaya, KIYaI-74-18 Preprint, Kiev (1974). 3. V. M. Raetskii and S. N. Votinov, Fiz. Met. Metalloved., 29, 284 (1970). 4. A. Damask et al., Philos. Mag., 22, 549 (1970). 5. Chemical Applications of Mossbauer Spectroscopy [Russian translation], Mir, Moscow (1970), p. 164. 6. P. 0. Voznyuk, L M. Vtyunnik, and V. N. Dubinin, Fiz. Met. Metalloved., 36, 1310 (1973). EFFECT OF TEMPERATURE ON THE POROSITY OF NICKEL IRRADIATED WITH NICKEL IONS S. Ya. Lebedev and S. D. Panin UDC 621.039.51 The effect of helium on the formation of pores in nickel and the dependence of the radiation porosity on the dose in irradiation with nickel ions at a constant temperature of the specimen (500?C) were investigated in [1, 2]. We consider here the development of porosity at different temperatures, of the target. The method used for preparing the specimens and the irradiation method were similar to those used earlier. Specimens of com- mercially pure nickel, which had a thickness of 0.15 mm, were first annealed in a vacuum at 800?C over a pe- riod of 1 h. Irradiation with 46-keV nickel ions was effected with a current density of 3 pA/cm2 to a dose of 1.6.1017 ions/cm2, which corresponded to approximately 40 displacements per atom [3]. The mean duration of irradiation was equal to 2.3 h. The specimen temperature varied in the 350-700?C range. Electron-microscope investigations of irradiated specimens have shown that vacancy porosity occurs at specimen temperatures >400?C. The results obtained in processing the photomicrographs are given in Table 1. Translated from Atomnaya. Energiya, Vol. 41, No. 6, pp. 428-429, December, 1976. Original article sub- mitted March 9, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1087 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 TABLE 1. Investigation Results for Irra- diated Specimens 4v/v17. Ternp.,1:. Pore density, . cm3 Pore size, A Swelling ?I? 400 4,5.101' 20 0,22 450 1,85.1016 70 1,25 500 1,40.1016 100 2,45 550 1,05.1016 165 8,2 600 3-100 340 12,1 650 7-1013 550 5,6 700 5,75.1012 1600 . 1,25 2 10 .., 10-2 .? 5 no .9 2 'Ei 5 10 a o 5 2 400 450 500 550 600 650 700 Ternp., Fig. 1 100 300 500 700 900 1100 Pore dimension, A Fig. 2 1300 1500 1700 Fig. 1. Temperature dependence of the swelling of nickel irradiated with Ni + ions to a dose of 1.6.1017 ions/cm2. Fig. 2. Distribution function of pore dimensions in nickel irradiated with Ni+ ions at different temperatures to a dose of 1.6. 1017 ions/cm2. A large number of small pores with the mean dimension < dv> 20 A is observed at temperatures above 400?C. Moreover; there is a large number of small dark spots, which, like the pores, are uniformly distributed over the specimen's area under observation. The pores increase with temperature, and we find < dv > 1600 A at 700?C. At the same time, the density of pores < Nv> drops sharply with an increase in temperature. This in- dicates that, at the low-temperature limit of pore formation, the incipient vacancy clusters are probably ther- mally stable in irradiation with a constant flux of bombarding particles. The low mobility of vacancies is the cause of the high density of small pores in the matrix. As the temperature rises, the vacancy mobility in- creases, which results in more intensive pore development. Our experimental data agree with the theory of uniform generation and development of vacancy porosity in irradiated metals [4]. Figure 1 shows the temperature dependence of swelling (AV/V) for nickel. The rather narrow swelling peak has a maximum at 600?C. The maximum volume change corresponds to 0.5Tme?K. The swelling maxi- mum observed at 600?C is in good agreement with data on nickel irradiation with 500-keV nickel ions [3]. The noticeable swelling at temperatures below 500?C can be explained by the high density of small pores. Figure 2 shows the distribution functions of pore sizes, obtained by processing histograms for different temperatures. It is evident that the function broadens with an increase in temperature, and its maximum shifts toward larger sizes. We observe a certain asymmetry of the curves, which indicates that the pores grow as a result of fusion of small pores. This behavior agrees with the dependence observed for cold-rolled M-316 stainless steel, irradiated in the DFR reactor [5]. The distribution function indicates that, with an increase in the irradiation temperature, the pore size in- creases, while the pore density diminishes. This confirms the assumption that generation of new pores does not occur; rather, there is a continuous growth of the pores formed at the initial stage of irradiation. 1088 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 LITERATURE CITED 1. S. Ya. Lebedev, S. D. Panin, and S. I. Rudnev, At. Energ., 38, No. 6, 426 (1975). 2. S. Ya. Lebedev, S. D. Panin, and S. I. Rudnev, At. Energ., 39, No. 5, 362 (1975). 3. J. Delaplace, N. Aram, L. Le Naour, J. Nucl. Mater., 47, No. 3, 278 (1973). 4. Yu. V. Konobeev and V. A. Pechenkin, in; Problems of Atomic Science and Technology, Radiation Damage Physics and Radiation Science of Materials Series [in Russian], Vol. 1, Atomizdat (1974), P. 41. 5. C. Cawthorne et al., in; Proc. Reading Conf. on Voids Formed by Irradiation of Reactor Materials, Harwell, BNES (1971), p. 35. NUMERICAL y-RAY ALBEDO FROM LIMITED SECTIONS OF THE SURFACE OF REFLECTING BARRIERS D. B. Pozdneev and M. A. Faddeev UDC 539.122:539.121.72 Data about the distribution of reflected y quanta of a point isotropic source on the surface of a semi-in- finite scatterer are given in [1-4]. However, no such information is available about barriers of finite thickness, although it is of considerable interest for radioisotope instrument manufacture, radiometry, and other areas of applications. prEd 8 7 6 5 4 31_ 2 a 6 145 279 662 1250 Ea, keV 13(E) 1,2- 1,0 ? 48 5 C6 ? 6 44 0 145 279 662 1250 Ea key Fig. 1. Values of fl monodirectional (a) and isotropic (b) sources: 1) Pb; 2) Sn; 3) Fe; 4) concrete; 5) C; 6) Be. Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 430-431, December, 1976. Original article sub- mitted March 3, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1089 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 79 20 18 18 16 70 8 4 2 a et. 41 42 0,3 V 7,0 2 5 10 al 0,2 0,3 16 14 2 I ta 8 16 14 12 10 8 6 4 2 42 43 0, 1 2 5 10 0,1 0,2 43 45 1 2 5 10 (f.p.1.) Fig. 2. Values of a at E =1.25 (a, b) and 0.279 MeV (c, d): a, c) iso- tropic source; c, d) monodirectional source; 1) Pb; 2) Sn; 3) Fe; 4) con- crete; 5) C; 6) Be. From analysis of data on the 7-ray albedo, obtained from Monte Carlo calculations according to a pro- gram described in [3], it follows that the numerical 7-ray albedo a (r, d) from a circular region of radius r on the surface of a barrier of thickness d can be described to within t 10% by the empirical formula a (r, d)=-- a (co, co) ?exp (?Or)] (i?exp (?a (r)(d? c)11, (1) where a (00, 00) is the asymptotic value of the albedo from a semi-infinite scatterer made of the same material when the primary-quanta source has a fixed energy of E0, and 13, a (r), and c are empirical quantities (Figs. 1 and 2). Formula (1) was obtained from the results of calculations for point isotropic sources and monodirectional beams normally incident upon the surface of a barrier at the point r =0 for E 0 =0.145, 0.279, 0.662, and 1250 MeV and for barriers made of Be, C, concrete, Fe, Sn, and Pb of varying thickness ranging from 0.1 to 5 free- path lengths (f.p.1.). The value of c for an isotropic source is 0.5 free-path lengths of primary quanta of energy E0, and c =0 for a monodirectional source. Using this formula, it is not difficult to find the number of reflected 7-ray quanta emitted from a surface portion of unit area, bounded by radii ri and r2 of concentric circles with center at the point r=0. LITERATURE CITED 1. B. P. Bulatov et al., Gamma-Ray Albedo fin Russian], Atomizdat, Moscow (1968). 2. 0. S. Marenkov, At. Energ., 21, No. 4, 297 (1966). 3. D. B. Pozdneev and M. A. Faddeev, Kernenergie, 16, No. 4, 105 (1973). 4. T. Nakamura and T. Hyodo, J. Nucl. Sci. Technol., 6, No. 3, 143 (1969). 1090 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 YIELDS OF zoon, 201T1, 202T1, AND 204T1 DURING PROTON AND DEUTERON IRRADIATION OF MERCURY P. P. Dmitriev, G. A. Molin, UDC 539.172.12 Z. P. Dmitrieva, and M. V. Panarin Irradiation of mercury with protons and deuterons produces 200T1 (T1/2=26.1 h), 201T1 (T1/2=73 h)3202T1 (T1/2=12 days), and 204T1(3.78 yr), whose half-life makes them convenient for use in applied and research prob- lems. In this paper, the 200-202T1 yield is measured as a function of the proton and deuteron energy during ir- radiation of thick mercury targets and theoretical yield curves are calculated for 204T1. The irradiated samples were prepared from mercuric oxide, the conversion factor for the yield for Hg0 to the yield in metallic mercury is 1.17. The proton and deuteron energy was varied with copper stopping foils. The methods employed to irradiate the samples and to measure the integrated irradiation current and the ac- tivity of the isotopes were the same as in [1, 6] which reported on work also done on the cyclotron of FL (Obninsk). Isotopes 200-2?2T1 decay through electron capture (EC) and therefore their quantum radiation contains strong components of IOC and LX rays produced by K and L capture as well as K and L conversion. Isotope 204T1 experiences decay (97.46%) and electron capture (2.54%) to the ground state of 44Pb and 2?4Hg, respec- tively, and hence does not emit y quanta. Table 1 gives the most intense components of quantum radiation accompanying the decay of 2?0-2?2T1 and 2?4T1. The gamma yield was obtained from the decay scheme given in [2, 3], taking the conversions into ac- count. The yield of 7, LX, and KX quanta from 200-202T1 is calculated by the formulas n(ICX)=-464,+04, ny alf,; n (LX)=61,,wi,+ L (ny nicLny ictK i) Here wk, WL. and ak,ctL are the fluorescence yield and the K- and L-conversion coefficients, respectively; nici, is the number of L vacancies freed per K vacancy. The probabilities of K and L capture, ek and CL, were calculated by the formulas and data of [4]; ? k, WL, and nKL are also given there. The values of ak and a L were taken from [2, 3, 5], and the LX quanta were assigned the energy of the most intense transition Lai = L3 ? M3. The yield of 10C and L quanta of 204T1 were obtained in [4]. The experimental yield curves for 200-202T1 and the theoretical curves for 204T1 are given in Figs. 1 and 2. The reactions producing 200-202n and 204T1 are given in Table 2. ? Table 3 presents measured yields for given proton and deuteron energy. The calculation of the theoreti- cal yield curves for 204T1 was similar to that in [6]. No data are available in the literature about the yield and cross section for nuclear reactions producing zoo-zozn and 204Ti. As is seen from Table 3, when the protons and deuterons are slowed down the absolute error of the en- ergy value increases. It thus follows that the path and energy of the particles are related by R ?E7/4, whence dE/E=K dR/R or dE =KE dR/R, i.e., the absolute error is proportional to the absorbed energy. Let us consider the production of 200-202T1 and 204T1 of high radioisotopic purity. Radioisotopically pure 204T1 and zozn, free of 200-202n, are obtained after appropriate cooling. Irradiation of mercury with protons and Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 431-434, December, 1976. Original article sub- mitted April 16, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1091 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 TABLE 1. Energy and Quantum Yield of 7, KX, and LX Quanta Iso- tope Energy of K, Quantum K LX yield, 'To References 200 TI 1205,7 367,97 Ka 70,14 Ko 80,71 LX 9,99 30,7 89,3* 71,3 20,7 34 12] Present paper It 2011 167,43 8,8* [3] Ka 70,14 69,8 Present paper 1(0 80,71 20,2 /1,X 9,99 45 n.?2 TI 439,4 92* [3] Ica 70,14 71,3 Present paper K6 80,71 20,7 /-X. 9,99 32 Ot? It 204 Tl Ka 70,14 1,14 [4] Ko 80,71 0,33 [4] L.,V 9,99 0,76 [4] *7 -ray peaks used to measure the activity. TABLE 2. Energy Thresholds of Reactions 2oo-2o2n and 204n, of Production of MeV Reaction Isotope 20011 201T1 2021'1 204T1 pn. 3,26 1,20 1,91 1,16 p2n 9,53 9,02 p3n 17,31 15,45 dn d2n 5,51 3,45 4,16 3,41 d3n 11,80 11,3 d4n 19,65 17,75 *The isotope is not produced in the given reaction. reaction. TABLE 3. Yield of 200-202n and 204T1 Origin and energy of particles, MeV Yield, ? Ciip A ? h 200T1 201T1 2021'1 204T1 * (IlgJrp): 22,4+0,43 1020-F130 860-F110 21,9-F2,8 0,0077 20,3-F0,47 765-F98 685-F87 16,4-F2,1 0,0077 17,1-F0,52 450+57 410+63 10,8-F1,4 0,0071 14,3-F0,56 195-F25 160-F20 7,2-F1,0 0,0060 10,4-F0,61 15-F1,9 20,0-F2,5 1,6+0,2 0,0020 (1-11?-1-d): 22,5+0,41 1065-F135 595-F76 61,0-F7,8 0,10 21,1?0,43 790-F100 460+59 51,0A-6,5 0,098 20,2+0,46 725+93 400+51 46,4-F5,9 0,092 18,8A-0,49 530-F68 245-F31 43,8H-5,6 0,085 17,4-F0,50 380-F49 165-F21 36,5-F4,7 0,075 14,3+0,55 135-F17 55,0-F7,0 19,5-F2,5 0,043 10,1-F0,62 10,0-F1,3 5,0-F0,7 1,5+0,2 0,008 *The values of the 204T1 yield were calculated. deuterons produces 198T1(T/ /2 = 5.3 h) and 199T1(T1 /2 = 7.4 h) whose activity becomes insignificant 3 days after the ir- radiation. The content of the admixture 294T1 in the case of proton irradiation is smaller than in the case of deuteron irradiation by a factor of - 10-20 and upon completion of the irradiation is 0.0008% for 200T1, 0.001% for 291T1, and -0.04% for 292T1. Radioisotopically pure 299T1 can be obtained by the reaction 197Au (a, n) 299T1. When gold was irradiated with 44-MeV alpha particles the 299T1 yield was 32 ACIJAA ? h, i.e., smaller by a factor of -30 than when mercury was irradiated with protons and deuterons. 1092 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 1200 1100 1000 - 800 1. 700 ,-- 600 500 400 300 200 100 28 24 20 15 12 8 4 8 12 16 20 24 Proton energy, MeV Fig. 1 Yield 202T1, 204T1, liCi/pA ? h 1100 _ 1000 - 900 ,z 800 U 700 600 500 ?-o a) 400 300 200 100 4 8 12 16 20 Proton energy, MeV Fig. 2 72 64 56 48 40 32 24 16 a 24 Yield 202T1, 204T1, ?Ci/tiA Fig. 1. Yield of 20o-202T1 and 204T1 vs proton energy for thick mercury targets: 0) 200T1;.) 20111; 46,) 2o2n; _ _ __.) 204T1 (x 1000). Fig. 2. Yield of 255-252T1 and 254T1 vs deuteron energy for thick mercury targets: 0) Mom I) 201n; 46,) 202n; _ _ .._) 204T1 (X200). Isotopes 255-252T1 free from 2134T1 are obtained by irradiating mercury with alpha particles. Then, (axn) PT, reactions produce 205Pb (To 201pb =21.5h), t.,_ ih .9.4 h), 202mpbco i =3.62 h), 252Pb (T1/2= 3 . 105 yr) which are parent isotopes of 20 255-252T1. The isotope 2111Pb decays as follows: 2o2mpb_.202T1 (9.5%), 2o2mpb _.202pb (90.5%). After -1.5 days only 255Pb and 25iPb remain in the lead fraction, and after -, 4 days, only 20opb, which is in dynamic equilibrium with 255T1. Because of the long half-life, the 252Pb activity is negligible. High-purity 251T1 can be obtained by the reaction 253T1(p, 3n) 201pb, the reaction threshold being 16.5 MeV. The proton en- ergy should not exceed the threshold of the reaction 253T1(p, 4n) 200pb, which is 25 MeV. Enriched mercury isotopes can be used to obtain 255-252T1 of extremely high radioisotopic purity. The authors thank G. N. Grinenko for his assistance in the work. LITERATURE CITED 1. P. P. Dmitriev et al., At. Energ., 31, No. 2, 157 (1971); 39, No. 2, 135 (1975). 2. T. Comppa et al, Nucl. Phys., A163, 513 (1971). 3. R. Auble, Nucl. Data Sheets, 5, 561 (1971). 4. M. Martin and P. Blichert-Toft, Nucl. Data Tables, A8, Nos. 1-2 (1970). 5. R. Hager and E. Selzer, Nucl. Data Tables, 4, Nos. 1-2 (1968). 6. P. P. Dmitriev et al., At. Energ., 32, No. 5, 426 (1972). 1093 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 ALBEDO OF A CYLINDRICAL ROD V. V. Orlov and V. S. Shulepin UDC 539.125.52:621.039.51.12 A system of multigroup albedo equations was obtained in [1] for the probability of reflection of neutrons from a cylindrical rod or from an external spherical medium. As shown in [1, 2] in the two limiting cases (a cylinder of very small and very large radius) the equations of [1] lead to a solution that is close to the exact one. The objective of the present paper is to verify the exactness of the equations of [1] for a cylinder of interme- diate radius. An idea of the exactness of the system of equations in [1] can be obtained by solving the one-velocity albedo equation df3idR = s-20 (Z s +2s) +s2? (1,R) (1) where 3 is the albedo, zs is the neutron scattering cross section, ze is the absorption cross section, and R is the radius of the rod. The initial condition when R =0 is f3(0) =1. In the case zs =0 Eq. (1) is of the form 13 = [1 - exp ( - 4RZc)]/4REc? (2) When Es 0 the solution of Eq. (1) can be rewritten as = 1- raoR/(1? aiR ?a2R2-Fa3R3 I- ? ? ?)i ? (3) Substitution of Eq. (3) into Eq. (1) makes it possible to determine the coefficients an. The radial relation (3) is analogous to the formula for the albedo of a cylindrical rod, obtained in [3]. Numerical calculations showed that the solution (3) displays good convergence right up to RZ =R (Es +Ec) =2. The exactness of Eq. (1) was verified by comparing the results of calculation by Eqs. (2) and (3) with the exact results given in [4]. The calculation was performed for Rz over the interval [0; 2] for different ratios h=Zs/E (Table 1). According to Table 1, the exactness of Eq. (1) is satisfactory for all h when R 0.5 or for h> 0.7 when R 2. As noted above, Eq. (1) also leads to correct results in the interval R for any h. Equation (2) is inexact for large RZ and small h because under these conditions a large contribution to the albedo is made by neutrons which do not experience collision in the rod and whose angular distribution is highly anisotropic. Equation (1) can be made more exact by introducing the albedo of unscattered neutrons. TABLE 1. Albedo of Rod for Different RZ Exact solution Solution of Eq. (1) h 2,0 1,5 1,0 0,5 0,25 0,17 2,0 1,5 1,0 0,5 0,25 0,17 0,0 0,053 0,095 0,186 0,404 0,623 0,726 0,125 0,166 0,245 0,432 0,632 0,73) 0,2 0,115 0,165 0,263 0,482 0,683 0.773 0,179 0,233 0,317 0,504 0,689 0,775 0,4 0,199 0,256 0,363 0,574 0,749 0,823 0,270 0,319 0,407 0,590 0,753 0,825 0,6 0,324 0,389 0,499 0,686 0,823 0,877 0,386 0,437 0,529 0,695 0,825 0,878 0,8 0,542 0,604 0,694 0,824 0,906 0,936 0.568 0,623 0,705 0.827 0,907 0,936 1,0 1,0 1,0 1,0 1,0 1,0 100 1,0 1,0 1,0 1,0 1,0 1,0 LITERATURE CITED 1. V. V. Orlov, At. Energ., 38, No. 1, 39 (1975). 2. Yu. N. Kazachenkov and V. V. Orlov, At. Energ., 18, No. 3, 226 (1965). Translated from Atomnaya Energiya, Vol. 41, No. 6, p. 434, December, 1976. Original article submitted March 9, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1094 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 3. A. Amauyal, P. Benoist, and J. Horowitz, J. Nucl. Energy, 6, 79 (1957). 4. J. Stuart, in: Problems of Nuclear Energy [in Russian], No. 6, Izd. Inost. Lit., Moscow (1958), P. 71. OPERATIVE MONITORING OF FISSION PRODUCTS IN SODIUM COOLANT OF FAST REACTOR V. B. Ivanov, V. I. Polyakov, UDC 621.373/374 Yu. V. Chechetkin, and V. I. Shipilov To monitor the state of the active zone of fast reactors during operation and to predict the radiation sit- uation when attending to the equipment it is necessary to know the level of the radioactivity of the fission prod- ucts and the rate at which they are accumulated in the loop. Existing systems for monitoring the hermetic ity of jackets with respect to delayed neutrons and radioactive gases do not directly yield such information. The choice of radionuclides whose activity can be measured to solve the problem posed is determined by the half-life, the character of their leakage through defects in the fuel jackets, and the contribution of their radiation to the dosage rate from the equipment. The principal difficulty in measuring the activity of fission products is determined by the fact that during the reactor operation this activity is smaller by a factor of 103- 10 6 than that of the short-lived activation isotope 24Na. To ensure operative monitoring of the variations in the activity of selected nuclides in the loop, it is nec- essary to carry out measurements right in the sodium channel, the duration of these measurements not ex- ceeding several hours. Measurement by the usual one-detector and Compton suppression spectrometer sys- tems with such ratios of the radioactivity monitored and interfering nuclides is possible only with very long measuring times. Compton (CS) and summing Compton (SCS) spectrometers yield a decrease in the Compton distribution by a factor of 100-1000 but in doing so they also have a significant loss in the efficiency of record- ing the total absorption peak. Because of their more compact geometry contiguous semiconductors (duodes) in the SCS mode make it possible to attain a lower efficiency loss than in the case of the SCS mode with dectors separated by a considerable distance. Studies of various types of detectors showed that thin composite planar detectors in the SCS mode are more sensitive than are detectors of the same total volume in the one-detector mode [1, 2]. The block diagram of an SCS, which was tested in measurements in a sodium loop bypass connected to the first loop of a BOR-60 reactor, is given in Fig. 1. The coolant flowed through a 16-mm tube. It took 100 sec for the sodium to arrive from the active zone. The detector was set up 2 m from the tube behind a 2-cm collimator. The detector used in the experiment was a Ge(Li) duode consisting of two planar detectors, D1 and D2, put together with a thin beryllium spacer. The total sensitive region was 12 cm3. The overall energy reso- lution for an energy of 662 keV with input loads of 102 and 1 ? 105 was 5 and 6 keV, respectively. Energywindows of 120-220 keV and 220-460 keV, respectively, were installed in the time channels of detectors D1 and D2 to record the spectra of the sum of coincident energy in the 340-680-keV range. The resolving time of the coincidence circuit was 75 nsec. The choice of the energy windows was determined by the desire to simultaneously monitor nuclides 137Cs (662 keV) and 1311 (364 and 637 keV). The y-ray spectrogram obtained with the spectrometer operating in the SCS mode (Fig. 2) in 150 min exhibits total absorption peaks with an energy of 364.5, 511, 637, and 662 keV. The specific activities found from the results of this experiment for 1311 and 137Cs, respectively, were (3 ? 0.6) ? 10-2 and (2 ? 0.4) ? 10-3 Ci/kg sodium with a 24Na radioactivity of 48 ? 5 Ci/kg. In a one-detector mode with a detector of the same volume no sought 7-ray peaks were resolved. Thus, the SCS technique with a composite semiconductor detector allowed the fission-fragment nuclides to be distinguished against the background of the 24Na activity in the first loop with a useful-to-interfering ac- Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 435-436, December, 1976. Original article sub- mitted April 20, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1095 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 /05' A 3 Dl D2 5 2 - 3 4 7 No. of pulses detected 104 103 102 101 511 keV 300 400 500 Channel No. 700 Fig., 1 Fig. 2 Fig. 1. Block diagram of summing Compton spectrometer with Ge(Li) duode and measuring ge- ometry: 1) piping with sodium coolant; 2) Ge(Li) duode; 3) fast charge-sensitive preamplifier; 4) logic circuit for selection and sampling of energy windows; 5) circuit for amplification and con- trolled shaping based on two delay lines (1 x 1 ?sec); 6) summator-integrator; 7) pulse-height analyzer. Fig. 2. -y--Ray spectrum of sodium coolant in the first loop of a BOR-60 reactor. tivity ratio of the order of 104. Optimization a detector size and construction, improvement of the loading properties of the spectrometric channel with lower counting loss due to pulse pile-up, and the choice of energy windows in the spectrometric and measuring channels will make it possible to obtain a lower limit of measure- meant for the summing Compton spectrometer and to determine the activity of other fission products. The development of the SCS technique, in the opinion of the authors, will make it possible to ensure reli- able continuous monitoring of the behavior of fission products in a sodium coolant and the development of de- fects in fuel element jackets. LITERATURE CITED 1. D. Walker and J. Palms, IEEE Trans. Nucl. Sci., NS-17, No. 3, 296 (1970). 2. V. Ivanov and V. Shipilov, Nucl. Instrum. and Methods, 119, 313 (1974). 1096 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 CONFERENCE AND MEETINGS REGENERATION OF FAST-REACTOR FUEL A. F. Tsarenko A meeting of IAEA experts on the regeneration of fast-reactor fuel was held in Leningrad from May 17 to 21, 1976, with the participation of representatives of the USSR, France, Great Britain, the USA, the Federal German Republic, Italy, Belgium and Japan. The agenda included a discussion of the state of the art in the re- generation of fuel from active zones: storage and shipping to a processing radiochemical plant; cleansing fuel assemblies from sodium; preparation for regeneration and the actual regeneration; waste handling. The papers by the experts and the discussion showed that the level and scale of research on fuel regener- ation in various countries are closely bound up with the level of development of the reactors themselves. The fullest reports of all on the topics under discussion were presented by specialists of the USSR and France. The following reactors are at present in successful operation in the world: BR-10, BOR-60, BN-350 (USSR); from 1967-1970 "Rhapsodie-Fortissimo" (France, 24-37 MW), from 1963 DFR [60 MW(T)] andfrom 1974 PFR [600 MW(T), Gt. Britain]. At the present time the following reactors are under construction or in the planning stage: in the Soviet Union BN-600 and BN-1500; in the United States FFTF [440 MW(T)], 1979, Clinch River [350 MW(T)], 1983, PLBR [1000-2000 MW(E)]; in France, jointly with the Federal German Republic and Italy, "Super Phoenix" [1200 MW(E)]; and Italy PEC [130 MW(T)], 1979. The Federal German Republic planned to start construction in 1976 of reactor KNK-2 [58 MW(T)] and in the same year Japan was to start construc- tion of JOY? [100 MW(E)], and in 1983, MONHU [300 MW(E)]. The development of reactor BN-1500 in the USSR, and the rate at which plutonium is accumulated during the regeneration of fuel from atomic power plants with thermal reactors (water moderated water cooled power reactor, VVER, RBMK) permit the conclusion that the widespread introduction of fast reactors will begin no earlier than 1990. The atomic power industry of the USSR will present atomic power plants with both thermal and fast reactors, with the relative proportion of the latter gradually increasing. It is advisable to centralim the re- generation of the fuel from such reactors in a radiochemical plant; according to calculations one plant should handle fuel from atomic power plants with a total rating of 6-12 million kW, i.e., e.g., from 4 to 8 BN-1500 reactors. This requires the plant site be chosen in keeping with the conditions of long-term storage of the radioactive waste as well as the safety of the population in the vicinity of the plant and protection of the envi- ronment from radioactive contamination. If one proceeds from the premise that the power input will be deter- mined by the plutonium build-up, reducing the cooling time of the fuel after discharge from the reactor from 3 yr to only 1 yr makes it possible for the rating of atomic power plants with fast reactors to be increased from 20 to 55 million kW, or 2.5 times, by the year 2000. United States experts discussed the principal parameters of scientific-research and experimental-design work on the regeneration of fuel from fast reactors with liquid metal coolants (LMFBR). These reactors are looked upon in the USA as one means of satisfying the national energy requirements starting from 1990. A "hot" experimental facility for regenerating fuel is to come into operation in 1988, and a pilot plant in 1995-2000. France, which has been developing fast reactors for the past 18 years, is at present the only country with industrial experience in the regeneration of spent mixed uranium-plutonium fuel (department AT-1 of the plant in Cape Ag, and the experimental department of the plant at 1VIarcoule). In Gt. Britain the research is concen- trated at the experimental plant at Dounreay. The research in Italy has been directed at adapting the JTREC experimental plant at Rotondella with a capacity of 10 kg oxide fuel for the regeneration of PEC reactor fuel. Similar work is being done in the Federal German Republic on the MJLLJ experimental facility with samples of DFR reactor fuel. Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 438-440, December, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1097 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 The Belgian research program envisages the construction of an experimental plant in the nuclear re- search center in Mole by 1980; this plant is intended for experiments on cutting fuel elements on the thermal oxidation processing of fuel, and on dissolution and gas scrubbing. In the late 1980s Japan plans to build an experimental plant for developing a technology for the regenera- tion of fuel from an experimental and a pilot reactor. In all countries at the present time the aqueous extraction technology (extraction recovery of uranium and plutonium from aqueous nitrate media with a solution of tri-n-butyl phosphate in inert solvents) is recog- nized to be technically feasible, but it is necessary to solve problems associated, in the first place, with the storage and shipment of mixed fuel, prepared for the operation preceding regeneration. In this plan the princi- pal effort should be to create effective methods for removing metallic sodium from fuel assemblies, decladding the fuel, dissolution, clarification, as well as for storing and shipping spent fuel elements. Storing, Cleansing Spent Assemblies From Coolant Residue, and Shipment Of all the possible methods of heat removal during storage of spent fuel from fast reactors (cooling with gas, sodium, bismuth-lead alloy, fused salts, organic materials) preference is given to sodium. Upon being discharged from the reactor, assemblies are held in a sodium medium in a storage vessel inside the reactor, and then for longer cooling are transferred to a sodium- and inert-gas filled (intermediate) storage vessel outside the reactor, and finally put into a water pond. The assemblies of the Rhapsodie and Phoenix reactors, for example, are held in the vessel inside the reactor for two months (the heat release drops to 0.4 and 6.0 kW, respectively, per assembly), and then in a medium of argon (first) or sodium (second) are put into a storage vessel outside the reactor with an inert gas or sodium, respectively. Assemblies of the PFR reactor are cooled for 9 months in the internal storage vessel of the reactor, and are then shipped to the reprocessing plant. Assemblies of the BN-350 reactor remain in internal storage for two months until the residual release is reduced from 20-30 to 6-7 kW per assembly. In the external storage vessel, consisting of a rotating drum with sodium, the assembly moves in an inert-gas medium. During the transfer (-4 min) the temperature of the fuel element cladding reaches 480?C. The holding period does not exceed the time between rechargings. When the heat release has been reduced to 3-4 kW per assembly, the assembly is cleansed from sodium and then transferred to a water-filled cooling pond. The most common method of cleansing assemblies from sodium is to treat them with water vapor with nitrogen, argon, or carbon dioxide gas as gas carrier. An experiment using a lead bath to remove sodium and to contain defective fuel elements (USSR) has demonstrated this method to be promising. However, some tech- nical problems remain to be solved. To prevent radioactivity from entering the storage vessel from assem- blies with defective fuel elements, it is deemed advisable to place the assemblies in hermetically sealed con- tainers. Problems due to the high level of heat release and radioactivity arise during shipment of spent fuel. The heat release of one assembly may reach 5-10 kW, depending on the assembly size, the irradiation conditions, and the cooling time. Present-day technology makes it possible to construct a shipping container designed for a heat removal of about 40 kW, which means one container can take 6 to 10 assemblies. The mass of such a container is 50-60 tons. It was achnowledged at the meeting that the shipment of spent fuel in a sodium medium, as well as the use of other heat-transfer media such as lead, have not yet been investigated sufficiently. Decladding the Spent Fuel Alongside the mechanical method of cutting fuel elements (France, Gt. Britain, USA) other methods have been developed: laser cutting (USSR, Britain), melting the claddings of corrosion-resistant steel (USSR), etc. Work is under way along such lines as constructing equipment for cutting individual fuel elements after dis- mantling the assembly, as well as for cutting assemblies as a whole. The decladding of spent fuel requires the solution of complex problems involved in the high heat release, cleansing from gaseous and fugacious fission products, and nuclear safety. Experts of some countries (France, USSR) voiced the opinion that the holding ("cooling") period for assemblies should be no less than 6-12 months. 1098 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Dissolution of Spent Fuel and Clarification of the Solution Dissolution of uranium-plutonium oxide fuel in concentrated (8-12 M) nitric acid proceeds in two stages; rapid dissolution of the bulk of the fuel and the slow dissolution of the residual fraction, constituting highly dis- persed particles (from > 1 to 10 consisting of the fuel composition (UO2- Pu02), as well as alloys containing fission products (Ru, Mo, Rh, Pd, Tc) add plutonium. The quantity of undissolved plutonium evidently depends on the conditions of the fuel fabrication, irradiation, and dissolution. Therefore, the fuel is dissolved into two stages, HF additives being used in the second stage (USSR, France). For this purpose apparatus of periodic and continuous operation is being developed. Most of the experts emphasized the need for thorough clarification of solutions prior to extraction. Work is under way to develop reliable filters and filtering centrifuges; filtration and centrifuging can be employed as two stages in the clarification process in any order. In Dounreay a high-speed centrifuge (20,000 rpm) is used for clarification;and in department AT-1, a pulsating filter is employed for that purpose. Soviet specialists propose flocculants in the solution clarification stage. Technology of Fuel Regeneration It was acknowledged at the meeting that in the near future fuel from the active zone will be regenerated by an aqueous-extraction technology of the "Purex process" type. The feasibility of highly efficient quantitative regeneration of fuel by such methods is regarded as having been demonstrated experimentally. Experience has been gained in regenerating mixed uranium-plutonium fuel on the experimental and pilot scale. The most important experience has been accumulated in department AT-1 which has already regener- ated about 1 ton of mixed oxide fuel from the Rhapsodie reactor with a burnup of up to 100,000 MW-days/ton. The experimental department at Marcoule has also started regenerating mixed fuel from the active zone. Sev- eral cycles have also been carried out at the MJLLJ plant in Karlsruhe (Federal German Republic) and at the Dounreay plant. A serious problem with adapting aqueous-extraction technology to the regeneration of active-zone fuel is that of radiation decomposition of the organic extractant and nitric acid in the first extraction cycle. This prob- lem is solved by using extraction apparatus with a short phase-contact time. Pulsating columns, and especially centrifugal extractors, are just such apparatuses. A very high plutonium content (up to 200 kg/ton) presents problems of accomplishing the complete.simul- taneous extraction of plutonium and uranium, preventing the formation of a third phase, salt-free separation of plutonium from uranium, monitoring and computing the plutonium content in solutions in conformity with the safety requirements and the system of guarantees, and attaining discharge concentrations of plutonium in the technological waste. An interesting method is that of electrochemical selective reduction and re-extraction of plutonium in the stage of separation from uranium. This method has been successfully tested in the regeneration of fuel samples at the MJLLJ plant in the Federal German Republic; there are plans to introduce the method in the WAK plant (Federal German Republic) and in Marcoule. Work onthe fluoride method of fuel regeneration is under way at present in the USSR, France, and Japan. The state of the development of this method is such that it cannot be introduced into industry. The basic prob- lems which must be solved are connected with attaining a high degree of separation of uranium and plutonium, deeply purifying plutonium from fission products, and completely extracting uranium and plutonium from the final product. Putting the fluoride method into industrial use requires an extensive program of scientific re- search and testing and design work and technological investigations. Fuel from the BOR-60 and Rhapsodie reac- tors with a short holding time (3-6 months) has been regenerated in single operations in the Fregat (USSR) and Attila (France) pilot plants, demonstrating the basic feasibility of the gaseous fluoride technology. Method of Decontaminating Waste Gases It is known that gaseous wastes of radiochemical reprocessing plants are the major source of radioactiv- ity entering the environment. The need to reduce the radioactive discharges into the atmosphere requires the introduction of efficient gas scrubbing systems in the plant. The state of the technology for trapping iodine, krypton, xenon, tritium, and 14C was discussion at the meeting in these terms. The iodine trapping methods are 1099 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 in the pre-industrial stage at present but none of them has found universal application. These are the "wet" methods [alkaline scrubbing, scrubbing with mercury nitrate solution, "Iodox process" - scrubbing gasses with fuming (concentration of more than 20 M) nitric acid] and dry methods based on the use of filter media in which silver salts are the active reagents. In eliminating krypton, more attention is paid to methods of cryogenic distillation in comparison with methods of absorption with fluorocarbons. As for the elimination of tritium, only attempts are being made to solve the problem. A method is being developed in the Federal German Republic and France for localizing (and building up) tritium in recirculating aqueous solutions of the main regeneration processes. Even if tritium is successfully concentrated in a small volume of waste, the problem of storing and eliminating waste will remain. In plants regenerating fast-reactor fuel the content of radioactive isotopes and plutonium will be signifi- cantly higher than in thermal-reactor fuel. Therefore, the ratio of the radioactivity discharged into the en- vironment to the activity in plants after the regeneration of fast-reactor fuel should be one-thousandth of that in plants regenerating thermal-reactor fuel. To attain this level it is necessary to make a substantial im- provement in the methods of localizing radioactivity in all stages of regeneration, to prevent losses of pluto- nium, and to develop effective methods for trapping gaseous and fugacious fission products. Waste Disposal The general consensus is that the safe disposal of waste from the regeneration of fast-reactor fuel will be based on methods which have already been accepted or are being developed for the fuel cycle of thermal reactors. Most programs are based on vitrification, which is recommended for solutions with a heat release of less than 5 ? 104 W/m3. For fast reactors it is necessary either to increase the holding of fuel for regenera- tion to more than a year, to mix the waste from fast and thermal reactors, or to develop hard matrices allowing storage at higher temperatures. MEETING OF FOUR NUCLEAR DATA CENTERS V. N. Manokhin Four nuclear data centers held their regular (12th) meeting in Vienna on April 26-27, 1976. It was at- tended by representatives of the National Neutron Cross Section Center (Brookhaven, USA), Neutron Data Com- pilation'Center (Saclay, France), Nuclear Data Center (Obninsk, USSR), Nuclear Data Section (IAEA, Austria), as well as Rumania and Poland. Brief reports on the work of each center over the previous year were read at the meeting. The National Neutron Cross Section Center is engaged in work on Series V of the evaluated nuclear data file ENDF/B. The second volume of the new BNL-35 atlas has been completed. In 1975 evaluated data on fis- sion products and reactor dosimetry were turned over for general use in the IAEA in 1975. The Neutron Data Compilation Center has done a great deal of work on creating a SINDA bibliographical catalog on collecting numerical data and on writing a program for data format conversion. The Nuclear Data Section prepared the SINDA-76 catalog for publication, having previously eliminated redundant and erroneous information. The CINDU-11 catalog of nuclear data possessed by the NDS has been published. And WRENDA-76, a register of inquiries for nuclear data, is being prepared for publication. Since April 1975 the Nuclear Data Center (NDC) has recorded 75 papers on magnetic tape. In all, to the present time it has recorded some 300 out of a total of 450 papers, containing numerical data, published in 1959-1975. The evaluated data (complete files) of a number of isotopes have been handed over to the IAEA. Four "Nuclear Constants" compilations were published and 110 inquiries were satisfied in 1975. The NDC has prepared for publication the proceedings of the Third All-Union Conference on Neutron Physics. The evaluation of all cross sections of nickel and chromium has been completed. Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 440-441, December, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1100 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 The 19th meeting discussed the state of the computer library of experimental neutron data in exchange format (EXFOR). At the present time this library contains more than 1,700,000 lines of information on mag- netic tape - numerical results and brief descriptions of the conditions of many experiments on neutron physics. The meeting also considered problems pertaining to the completeness of the data library, data exchange, errors in the magnetic tape recordings, and the introduction of changes suggested earlier in the EXFOR format. Some problems of the SINDA bibliographic catalog were also considered (ensuring the completeness of the abstracts, the connection with the INIS system, the periodicity of publication, etc.). The meeting recommended that all centers inform each other about the existence of evaluated data, de- scriptions of evaluations, and results of comparison of the evaluated data of the various libraries. The results of the meeting showed that the international cooperation on the exchange of experimental neutron data is developing. The exchange of evaluated neutron data is increasing. The next meeting of the four nuclear data centers will be held in Obninsk in April, 1977. MEETINGS ON THE COMPILATION OF NUCLEAR DATA FROM REACTIONS WITH CHARGED PARTICLES AND DATA ON THE STRUCTURE OF THE ATOMIC NUCLEUS L. L. Sokolovskii The meeting of consultants on data from reactions with nuclear particles, organized by the Nuclear Data Section of the IAEA, was held in Vienna from April 28 to 30, 1976, with the participation of representatives of Gt. Britain, Poland, Rumania, the USSR, the USA, France, the Federal German Republic, and Japan. The agenda was purely technical: the index of bibliographical data, the range of compilation, computer file, keywords, dictionaries, improvement and coordination of dictionaries, and the principal rules of EXFOR for reactions with charged particles. As a result of the work of the Meeting an international network was established of Centers and groups participating in the exchange of numerical, bibliographical, and evaluated material on data from reactions with charged particles. It was decided that the next meeting on the compilation of data from reactions with charged particles would be held in the USSR after tha All-Union Conference on Neutron Physics and the Meeting of Four Neutron Centers in 1977. The Meeting on Data on the Structure of the Atomic Nucleus and Radioactive Decay was also organized by the Nuclear Data Section of the IAEA and took place in Vienna from May 3 to 7, 1976. The meeting was at- tended by representatives of Austria, Belgium, the Federal German Republic, Gt. Britain, Holland, Hungary, Italy, Japan, Poland, Rumania, the USA, the USSR, and Sweden. Some of the subjects brought up for discussion were: definition of systems of exchange of nuclear data (bibliographical, numerical, and evaluated) and the format of the exchange, the general rules and terminology (dictionaries, methods of evaluation), the interna- tional file of evaluated data on the structure of the nucleus and decay (contents, structure, format, distribution), and international cooperation on data compilation and evaluation. In the USA the work on the evaluation and compilation of data on the structure of the atomic nucleus and radioactive decay is concentrated in four laboratories (Brookhaven, Berkeley, Oak Ridge, and Idaho) and in the University of Pennsylvania. Work will be done here on all mass chains, except A=21-44. That chain has been traditionally the domain of the State University of Utrecht (Holland). The United States plans to make a revi- sion of the A-chains once every four years and invites other countries to participate in this so as to make the Translated from Atomnaya Energiya, Vol. 41, No. 6, p. 441, December, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1101 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 analysis easier for themselves, on the one hand, and, on the other hand, to make the data file international. The US delegation reaffirmed the principle of free exchange of data of all forms. Before the Vienna meeting, representatives of some West European countries got together to discuss some questions of the compilation and evaluation of nuclear data (April 27, 1976, in Belgium). At their meeting they pointed out the importance of international cooperation in the evaluation of nuclear data. The Vienna meeting adopted as a recommendation the "Recent References" system of keywords as an international system for the exchange of bibliography on data on nuclear structure and decay. The format of the Oak Ridge laboratory was recommended as a provisional one for the exchange of numerical and evaluated data. The next meeting is planned for September or October, 1977. The site was not fixed definitely. The proceedings of the meeting are in the Center of Data on Nuclear Structure and Nuclear Reactions (Moscow). IAEA SYMPOSIUM ON THE DESIGN AND EQUIPMENT OF "HOT" LABORATORIES B. I. Ryabov Delegations from 32 countries and three international organizations participated in the Symposium which was held Aug. 2-6, 1976, in Finland. A total of 46 papers were presented in four main sections: safety in planning and design; systems of air transfer and purification; critically monitoring, fire protection, and waste handling; radiation protection and administrative measures; operating experience. Primary attention was paid to technical, structural, and organizational measures for increasing radiation safety during work with kilogram quantities of plutonium and with gram quantities of transuranium elements. Interest in these problems has been aroused by extension of fast reactor programs. Increased radiation safety during work with irradiated fuel with a high plutonium content required addi- tional safety measures. Comparatively few special laboratories are equipped for this purpose; mention was made of only two new facilities being built in Japan and projects for another two in the USA. In other countries, existing facilities are being reconstructed and modernized. Most laboratories for work with irradiated fuel in- corporate large "hot" caves (up to 20 m long, 4-6 m wide, 7-10 m high) with a high degree of containment, biological shielding for activities up to 106 Ci, dismountable cover, bridge crane inside, and electromechanical manipulators. Besides large hot caves, ordinary-sized ones are used, but with dismountable shielding (or re- tractable wall) and boxes. Caves and boxes are equipped with critical-mass safety devices, a system for handling the product without affecting the air-tightness, an efficient system for decontaminating the exhaust air, and automatic fire-fighting installations employing liquid carbon dioxide or tetrafluorodibromomethane. Par- ticular attention is paid to fire safety because the products are pyrophoric, which means that the possibility of a fire breaking out cannot be ruled out, and in such an event the filters go out of service and the ejection of a large quantity of activity is unavoidable. In the United States the mean radiation dose allowed for personnel in "hot" laboratories is 1 rem/yr (0.5 mrem/h), the buildings and equipment are designed to withstand hurricane winds (tornadoes) and seismic activity of up to 9 points, which makes the construction about 10% more expensive but guarantees that dangerous contaminants are not dispersed in the event of natural disasters. When new laboratories are designed and old ones rebuilt, measures are planned for the well-timed shutdown of plants when breakdowns occur, and for the deactivation, dismantling, and evacuation of the equipment. Much consideration is given to the standardization of the basic equipment and its subassemblies; this is a means not only of cutting the cost of radiation safety but also of increasing it since this makes it possible to Translated from Atomnaya Energiya, Vol. 41, No. 6, pp. 442-443, December, 1976. This material is protected by copyright registered in the name of 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 $7.50. 1102 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 introduce into practice equipment that is well designed, tried and tested, and therefore reliable. This purpose is also served by the use of high-quality materials. Standardized equipment and subassemblies is manufactured on an industrial scale in France and is sold not only on the home market. Another aspect of increased safety is that of automation of frequently recurring operations, such as analysis of fuel composition, 7 spectrometry, and photometry. The operations are carried out using a computer. To automate auxiliary operations, France has developed a new manipulator, the MM-8, with programmed control and feedback, the program being re- corded in the computer automatically when the operator makes the required movements manually. A short- coming of the manipulator is that all the slave motors are inside the cave and this cannot ensure a long period of operation with a high level of 7-ray activity. The air from the a- and y-contaminated drive mechanisms, shielded by a sheath with forced air, is expelled through a filter. Stringent requirements are put on the ventilation systems of hot laboratories. Caves and boxes are kept at a vacuum of 25 to 40 mm H20 (or even more) and the pressure drop between a cave and the repair zone is 6-25 mm H20. Most caves and boxes have a vent system for inflow ventilation which maintains the specified parameters of air interchange if the vacuum drops because of a leak (rupture of a glove, broken glass, etc.); in this case an alarm system is actuated and a sound signal is sounded. The inflowing air is purified by filters, and the exhaust from hot caves and boxes is, as a rule, in many stages. 131I is trapped with high-quality carbon filters, designed with a frame compressed by a screw mechanism which ensures uniform density of the pow- dered carbon. A sophisticated air purification system is used when fuel elements are taken apart. Behind the carbon filters are scrubbing columns which are irrigated with sulfate compounds, mercury nitrates, and other solutions. After purification, the air is treated on filters of diodide or silver nitrate and on a monitoring filter. The coefficient of air purification with such a system is 105. It was particularly noted that there are as yet no filters resistant to high temperature (more than 200?C); consequently, in a fire the purification systems cease to function. Major accidents can be caused by violation of the critical-mass safety. Data were presented at the sym- posium about the force of the bang in a box during a spontaneous reaction in terms of exploding TNT: with 2.1017 fissions per second the force of the bang is that of 25 g TNT, and at 7. 1018 fissions/sec, 1000 g TNT. The most likely minimum accident is with 2 1017 fissions/sec, but even in this case boxes are destroyed. A possible accident is prevented by critical-mass-safety devices as well as the limitation of the quantity of plu- tonium in the apparatuses. Particular difficulty is presented by the monitoring of critical-mass safety. Under development are new two-stage scintillation detectors which begin operating at the beginning of the reaction and give an alarm signal at a dose of 0.6 R/sec. Organizational and technical measures enabling safe operating conditions to be maintained in "hot" labo- ratories found an important place in the discussion. This includes the development of new sanitary regulations, detailed reports by existing enterprises on the safety measures in plants, on a concrete radiation facility, sta- tistics, etc. After considering and approving these reports, the competent state organs give permission to con- tinue operations. Reports are resubmitted if the technology at the plant is altered or if the basic equipment is changed (France). In many countries, permission for work with plutonium is renewed each year with an indica- tion of the maximum quantity allowed the given laboratory. The personnel undergoes the requisite training and should have high qualifications, and the basic equipment is subjected to a prophylactic examination and main- tenance at least every 18 months. Considerable interest was aroused among foreign delegations by the USSR delegation report which pre- sented the main directions taken in solving the safety problems in designing "hot" laboratories and experimen- tal facilities in our country. The symposium proceedings will be published by the IAEA. 1103 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 Declassified and Approved For Release 2013/09/23: CIA-RDP10-02196R000700080005-5 SECOND SEMINAR ON COMPUTER SIMULATION OF RADIATION AND OTHER DEFECTS Yu. V. Trushin The seminar, which was held in the M. I. Kalinin Polytechnic Institute in Leningrad, June 22-24, 1976, was devoted primarily to radiation defects. The seminar was opened with a review by A. N. Orlov (A. F. Ioffe Physicoteclmical Institute, Academy of Sciences of the USSR) who analyzed publications on "Radiation effects and the nuclear power industry" in the light of requirements put upon the material of thermonuclear reactor walls. Computer simulation can be used to solve some problems and in individual cases is as yet the only ac- cessible method of investigation. A paper by V. Y