Soviet Atomic Energy - Vol. 38, No. 6

Body: 1-i)?Zel jit l'ilttr" Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 ? ? ? '-17 # e -17) Russian OriginalVol. 3, No. 6, June, 1975 q?) December, 1975 244 ,L V 462-c-3-147 (73.'V1 ATEAZ 38(6) 469-570 (1975) SOVIET ATOMIC ENERGY ATOMHAFI 3HEPri1fl (ATOMNAYA iNERGIYA) TRANSLATED FROM RUSSIAN CONSULTANTS BUREAU, NEW YORK Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 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 translation of Atomnaya thergiya, 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. EditoriarBoard of Atomnaya Energiya: ? Editor: M. D. Millionshchikov Deputy Director I. V. Kurchatov Institute of Atomic Energy Academy of Sciences of the USSR Moscow, USSR Associate Editor: N. A. Vlasov A. A. Bochvar, N. A. Dollezhar ( V. S. Fursov I. N. Golovin V. F. Kalinin ' A. K. Krasin A. P. Zefirov ? V. V. Matveev M. G. Meshcheryakov P. N. Palei ' ' V. B. Shevchenko V. I. Smirnov, A. P. Vinogradov Copyright C) 1975 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. For bibliographic accuracy, the English issue published by Consultants Bureau carries the same number nd date as the original Faissian from which it was translated: .For example, a Russian issue published in December will appear in a Consultants Bureau English translation about the following June, but the translation issue will carry the December date. When ordering any volume or particu- lar issue of a Consultants Bureau joupal, please specify the date and, where appli- cable, the volume and issue numbers of the original Russian. The material you will receive will be a translation of that Russian volume or issue. Subscription Single Issue: $50 Single Article: $15 $87.50 per volume (6 Issues) ' Prices somewhat higher outside the United States. , CONSULTANTS BUREAU, NEW yoRK AND LONDON 227 West 17th Street New York, New York 10011 Lower John Street London WI R 3.1)13* England Published monthly. Second-class postage paid at Jamaica, New York 11431. Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya December, 1975 Volume 38, Number 6 June, 1975 ARTICLES ? Optimization of the Operating Conditions of Mass-Diffusion Cascade Installations ? V. A. Chuzhinov, N. I. Laguntsov, B. I. Nikolaev, and G. A. Sulaberidze ? Mathematical Simulation of the Extractive Reprocessing of Nuclear Fuel. 3. Redox Reextraction Using Iron Salts ? A. M. Rozen, M. Ya. Zeltvenskii, and I. V. Shilin Two-Dimensional Diffusion Program HEXAGA II for Many-Group Calculations of CONTENTS Engl./Russ. 469 363 473'. 367 Hexagonal Lattices ? T. Apostolov and Z. Woznicki 479 372 Investigation of the Adiabatic Outflow of Water through Cylindrical Channels ? V. S. Aleshin, Yu. A. Kalaida, and V. V. Fisenko 483 375 ? Oxidation of Tritium in Air under the Action of Intrinsic Radiation ? L. F. Belovodskii, V. K. Gaevoi, V. I. Grishmanovskii, and N. V. Nefedov 488 379 Experiments on the Synthesis of Neutron-Deficient Isotopes of Kurchatovium in Reactions with Accelerated "Ti Ions ? Yu. Ts. Oganesyan, A. G. Demin, A. S. Illinov, S. P. Trettyakova, A. A. Pleve, Yu. E. Penionzhkevich, M. P.Ivanov, and Yu. P. Trettyakov 492 382 REVIEWS ? Problem of Environmental Protection in the Operation of Nuclear Power Stations ? N. G. Gusev 502 391 Contemporary Trends in Experimental Shielding Physics Research ? V. P. Mashkovich and S. G. Tsypin 510 398 Numerical Solutions of the Kinetic Equation for Reactor Shielding Problems ? T. A. Germogenova 513 401 BOOK REVIEWS Yu. A. Egorova (editor). Problems of the Physics of Reactor Shielding ? Reviewedby B. R. Bergeltson 518 405 ARTICLES Problems of Secondary Gamma Radiation in Reactor Shields ? A. A. Abagyan, T. A. Germogenova, A. A. Dubinin, V.1. Zhuravlev, V. A. Klimanov, E. I. Kostin, V. P. Mashkovich, V. K. Sakharov, and V. A. Utkin 520 406 ABSTRACTS Effect of Proton Irradiation on the Operation of a Scintillation Counter ? B. V. Gubinskii, E. M. Iovenko, V. A. Kuztmin, V. G. Mikutskii, and V. N. Nikolaev 524 411 ? Experimental Determination of the Temperature Dependence of the Thermal Conductivity of Uranium Dioxide under Conditions of Reactor Irradiation ? B. V. Samsonov, Yu. G. Spiridonov, N. A. Fomin, and V. A. Tsykanov 525 412 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 CONTENTS (continued) Engl./Russ. Mechanical Strength of Uranium Field-Emitters ? A. L. Suvorov, G. M. Kukavadze, D. M. Skorov, B. A. Kahn, A. F. Bobkov, V. A. Fedorchenko, B. V. Sharov, and G. N. Shishkin 526 412 Dose Distribution in a Tissue-Equivalent Medium from a Plane Thin Isotropic Alpha- Particle Source ? D. P. Osanov, V. P. Panova, Yu. N. Podsevalov, and E. B. Ershov 527 413 Recovery of the Integrated Spectrum of Neutrons in the Energy Range 0.1-3 MeV by the Extrapolation Method ? R. D. Vasil'ev, E. I. Grigoraev, G. B. Tarnovskii, and V. P. Yaryna 528 414 Determination of Traces of Nitrogen in Pure Metals by Gamma Activation ? A. F. Gorenko, A. S. Zadvornyi, A. P. Klyucharev, and N. A. Skakun 529 415 Microscopic Distribution of Ionization Events in an Irradiated Medium as a Characteristic of the Quality of Ionizing Radiation ? I. B. Keirim-Markus, A. K. Savinskii, and I. V. Filyushkin 530 415 Allowance for Fluctuations in theIrradiationDose of Lungs by Highly Active Particles ? 0. M. Zaraev and B. N. Rakhmanov 531 416 LETTERS TO THE EDITOR Calculation of Heterogeneous Nuclear Reactors by the Method of Inserted Elements ? I. S. Slesarev and A. M. Sirotkin 533 419 Transient Changes of the Thermoelectric Characteristics of Thermocouples by the Action of Reactor Radiation ? V. P. Kornilov and E. . B. Pereslavtsev 536 420 Oxidation of Solid Solutions of Uranium and Niobium Monocarbides ? V. G. Vlasov, V. A. Alabushev, and A. R. Beketov 539 422 Spectral Characteristics of the Background Noise in the Primary Loop of an Atomic Generating Plant ? K. A. Adamenkov, V. I. Gorbachev, Yu. V. Zakharov, V. P. Kruglov, S. A. Paraev, Yu. A. Reznikov, and R. F. Khasyanov 543 425 Pores in Helium-Saturated Nickel under Irradiation by Nickel Ions ? S. Ya. Lebedev, S. D. Panin, and S. I. Rudnev 545 426 Scintillating Plastics with Improved Radiation Resistance ? V. M. Gorbachev, V. V. Kuzyanov, Z. I. Peshkova, E. A. Rostovtseva, and N. A. Uvarov 547 427 Selecting the Operating Point in Resource Tests of Thermionic Converters ? V. I. Berzhatyi, A. S. Karnaukhov, and V. V. Sinyavskii 550 429 Neutron Resonances in 18lTa at 2-70 eV ? T. S. Belanova, A. G. Kolesov, V. A. Poruchikov, S. M. Kalebin, and V. S. Artamonov 553 430 COMECON NEWS Collaboration Daybook 555 432 BOOK REVIEWS E. M. Filippov. Nuclear Geophysics ? Reviewed by J. A. Czubek 557 433 INFORMATION: CONFERENCES AND MEETINGS Symposium of the International Agency on Atomic Energy Regarding the Choice of Areas for Nuclear Installations ? N. P. Dergachev a All-Union Scientific Conference on Shielding against Ionizing Radiations from Nuclear- 559 434 Industrial Installations ? V. P. Mashkovich 562 436 4 Fourth All-Union Conference on the Physics and Technology of High Vacuum ? G. L. Saksaganskii 565 437 BOOK REVIEWS B. A. Ushakov, V. D. Nikitin, and I. Ya. Emelayanov. Foundations of Thermionic Energy Conversion ? Reviewed by Yu. I. Skorik 568 439 The Russian press date (podpisano k pechati) of this issue was 5/28/1975. Publication therefore did not occur prior to this date, but must be assumed to have taken place reasonably soon thereafter. Declassified and Approved For Release 2013/09/25 : CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 ARTICLE S OPTIMIZATION OF THE OPERATING CONDITIONS OF MASS-DIFFUSION CASCADE INSTALLATIONS V. A. Chuzhinov, N. I. Laguntsov, UDC 621.039.31 B. I. Nikolaev, and G. A. Sulaberidze In the design and execution of practical cascade installations a calculation of the optimum operating conditions of the separating devices is extremely vital. A correct choice of conditions enables the number of such devices in the cascade and the energy capacity of production to be reduced for special external working conditions. When solving problems of the type encountered in separation practice, it is desirable in a number of cases to use the method of mass diffusion; this has the advantages of servicing simplicity, fairly short times of installation, compactness of apparatus, universality, and speed of operation, i.e., it enables iso- topes of a variety of elements to be produced in the same apparatus in a very short time [1, 2]. How- ever, the absence of published data as to methods of optimizing mass-diffusion apparatus, as distinct from such classical methods of separation as thermal diffusion and distillation [3-8], makes it difficult to create effective mass-diffusion separators. This paper is devoted to questions of optimizing mass-diffusion units (columns or pumps) in cas- cades. Since the creation and operation of practical separating installations involve the expenditure of considerable material resources and electrical power, optimization of the cascade is best carried out by reference to a criterion as fully as possible reflecting the economy of the method and the separating pro- cess. One such criterion is the net cost of the resultant product. An analytical expression for estimating the net cost of an isotope mixture enriched with the valuable component may be obtained if we take account of the main expenses incurred in the production of an isotope in the cascade (for example, one consisting of mass-diffusion columns). For simplicity, let us assume that all the columns in the cascade work in the same mode. The form efficiency of the cascade 77, defined as the ratio of the separating power AUid of an ideal cascade having specified values of individual flows and concentrations [F and cp in the feed, P and cp in the outgoing ma- terials (product), and W and cw in the spent materials (waste)1, to the separating power AU of a real cascade with the same flows and concentrations at the input and outputs, may be expressed in the following form [4]: 1= AU id 1 (PD (CA+ WO (cw)?Fa)(cp)1, AU (//2/4K) lz (1) where H2/4K = SU is the specific separating power of the column, H and K are the coefficients of the trans- fer equation for the column, / z is the total length of all the columns in the cascade, .t (c) = (2c-1) In [c /(1?c)] is the separating potential. From Eq. (1) we obtain an expression for the total length of the columns in a real cascade: /x? (H/4K) [PO (cp)d- (cw) ? Fa) (CAI. 2 (2) Since mass diffusion relates to irreversible methods of separation, we may consider that the energy required by the cascade and the amount of cooling liquid employed are proportional to the length of the Translated from Atomnaya Energiya, Vol. 38, No. 6, pp. 363-366, June, 1975. Original article submitted July 1, 1974. ? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 469 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 2 7 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 0 2 4 6 6 6 Fig. 1. Net cost and cost compo- nents as functions of the vapor flow. cascade. Using aK to denote the cost of unit length of the column, we may write the capital expenditure 37K required for creating the cascade in the form YK = aK/I (3) If To is the period of service of the cascade, determined in ac- cordance with established norms, while t is the time of operation required to obtain a specified amount of product, the part of the cost of the cascade which is transferable to the cost of the iso- tope (amortization deductions) will be Ya = aazt/To. Since the power consumed by the cascade installation is proportional to the total flow of vapor, its magnitude may be defined as Q (4) (5) where q is the flow of vapor associated with unit length of the cascade, b is the energy consumed in creat- ing unit vapor flow. The specific flow of the vapor q may be expressed in terms of the working parameters of the column: q = 22-cr2u = 2lEnDi0cr? (6) Here r2 is the radius of the inner condenser of the column, u is the density of the vapor flow at the con- densing surface, n is the molar density, Dio is the diffusion coefficient of the separated gas in the vapor of the working liquid, a = ur2/nD to is a dimensionless parameter of the column, an analog of the Peclet dif- fusion number. Hence the energy needed in order to obtain a specified quantity of the isotope is E 2abnD10oth, and its cost may be written in the following form YE 2atztEDioat/E, where aE = baoE (ace is the distribution cost of the unit of energy). (7) ( 8) In an analogous manner, for estimating the expenditure on cooling liquid we obtain the expression yb= 2na6nD10ut/E, (9) where ab is a coefficient allowing for the coolant expenses associated with unit flow of the vapor q. Apart from these components, the net cost of the isotope must include the expenditure on the raw ma- terial required to produce the specified amount of isotope yF = aF Ft, the expenditure of the wages of the servicing personnel ywg awgt, and also the expenditure on repairing the cascade installation. The coeffi- cients aF and awg represent the distribution cost of one unit of raw material and the wages per unit time respectively. The cost of the capital and preventive repairs of the separating apparatus we define as a part of the capital outlay; we allow for these in the expression for the net cost of the product by means of a coefficient ki. The value of k1 depends on the complexity and specific characteristics of the apparatus employed and is chosen on the basis of experience in the use of analogous installations. Thus the expression for estimating the total cost of the product being manufactured (the isotope) assumes the form Y? [PO (C p) W (C ? (CF)1 [ (1?ki) (H2/4K) 2anD jog abdaEFt-:- awg t. To (10) Let us consider various possible cases of the use of this function. If in a natural isotope mixture only one isotope is valuable, then, allowing for the balance equation relating to the whole cascade 470 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 P+W=P; PCp+WCW=FCF> we obtain the following expression for the net cost of the isotope from Eq. (10): 1 To SPr (H2/4K) L F k Pim Cp?Cp oic wi Cp?Cwoic ir aR(1-1-ki) cp?cw k CI??CW +231nDioa (aE-1-abd Pt _1 L aF CP?CW' - Cp?Cw P For a specified geometry of the column the specific separating capacity 6U = H2/4K is determined by the dimensionless flow of vapor y and by the circulation, the maximum value of which is in turn pro- portional to u. Since the separating capacity (for a fixed flow of vapor) increases monotonically with in- creasing circulation, tending toward a limiting value 61Jiirn, it is desirable to establish the value of the circulation for which 6U is close to SUlim. Then the choice of operating conditions for the columns in the cascade will be determined simply by the vapor flow. Since the capital outlay is proportional to 1/SU and the power required to the quantity cr/oU, it follows from Eq. (11) that the energy and apparatus optima for a cascade of mass-diffusion columns do not coin- cide. Hence the operating conditions of the columns in the cascade have to be chosen in such a way that the net cost of the resultant produce may be a minimum. It is also useful to use the net-cost function (11) in order to optimize the cascade with respect to the dimensions of the stripping section. For this purpose it is sufficient to find the minimum of Spr with respect to the value of cw. In the particular case in which the net cost of the raw material is small, or when the product obtained in the spent materials may be realized at the original (supply) price, there is no need to consider the spent-materials section, and the expression for the net cost may be written in the form: where s(I)(cp, cp) I_ aK (1 ?ki) +271nDIOG (aE-Hab) _I+ Pr (H2/4K) To cp) ?2cP) (r)(c. cp)=-- (2c ? 1) ln c?c ?42c') = (12) (13) Also of interest is the case in which both isotopes of the separated mixtures are valuable. Then in addition to the value of Cp we must specify the required concentration of the second isotope 1?cw, which is enriched at the other end of the cascade. We may consider that the expenses incurred in producing each isotope are proportional to the total length of the columns of the particular section in which the isotope under consideration is enriched: = yp YIV, Yp (cp, Yw Wcp(cw, co (14) (15) where Yp and Yw are the expenses associated with the creation and exploitation of the enriched and spent (waste) sections of the cascade respectively. The net cost of each isotope will in this case be determined by an expression analogous to Eq. (12): P4) (cp, Cp) Cp?Cw SP P Pr pt (11.214K)1 (cp, co +MD cr) {[(13 (c2')+ cc cc ?Fir ?1)(c"') cp?CW (1)(cF)..1 X [ To io E b --L23-EnD 0 (a -} a d + cr?Civ 1-- -- 2-iga (16) CF?CW P S"? Pr ? Cp?CW (Dm WCD(Cw, CF) f r c,? (D(c.o+ W (CIV) Cp?CF YW Wt (H214K)ii PO (cp, +WO (cw, co I Cp?Cp ab) + F cp?cp w [ a (aE? a cr?cw+ a wg al (1+ ki) 2anDio _ T (17) By repeating the discussions outlined above we may obtain a function for the net cost of the isotopes concentrated in a cascade of mass-diffusion (Hertz) pumps. In this case, in order to estimate the total net cost of the product, we must replace the separating power H2/4K in Eq. (10) by the separating power of the pump 61JH expressed in terms of its working and geometrical parameters [9,10]. The term allow- ing for the energy expenses also changes slightly. For a cascade of mase-diffusion pumps with specified geometry and working conditions Eq. (10) takes the form 471 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 I [ )] ;, +ki) z Y? t [Pao (cp) 4- O (civ) ? (4 a )] nD au 01 W T I e In q. 1 ?On ? a FFt awgt. (18) Here Z is the area of the porous diaphragm, /e is the effective _diffusion length [10], lnq = u/e/nDio is the Peclet diffusion number, en is the partition coefficient of the vapor flow, a'E k ak, an' are coefficients allowing for the corresponding expenses. From Eq. (18) we may easily obtain the net-cost function for various cases of transferring the total expenses to the isotopes produced. Figure 1 shows the net cost of 99% I3CH4 calculated from Eq. (12) and also the components allowing for the capital (SE) and energy (SE) expenditure on the creation and exploitation of the installation as func- tions of the quantity a. The data required for the calculation were obtained during laboratory research into mass-diffusion columns carried out earlier [1]. The values of SE, SE, and Spr are given in relative units. The terms of Eq. (12) allowing for the wages of the servicing personnel and the cost of the original ma- terials are omitted, since these have no effect on the appearance of the curves. We see from Fig. 1 that, as a result of the noncoincidence of the optimal energy and capital outlays, the value of a corresponding to the minimum net cost is smaller than the value of a corresponding to the maximum separating power of the column. By using the method proposed we may generalize the resultant net-cost function to other methods of separation (thermal diffusion, gas diffusion, and so on). It is clear that in every specific case the optimiza- tion parameters are to be chosen with due regard for the special features of the separating method and the separating devices. The results may be used in designing mass-diffusion separating installations and also when comparing the efficiency of methods used for separating various isotopes. The authors wish to thank G. A. Tevzadze for discussing the work and also for valuable comments. LITERATURE. CITED 1. B. I. Nikolaev et al., At. Energ., 23, No. 1, 62 (1967). 2. B. I. Nikolaev et al., Isotopenpraxis, 6, No. 11, 417 (1970). 3. K. Jones and W. Ferry, Separation of Isotopes by Thermal Diffusion [Russian translation], IL, Mos- cow (1947). 4. A. M. Rozen, Theory of Isotope Separation in Columns [in Russian], Atomizdat, Moscow (1960). 5. G. D. Rabinovich, R. Ya. Gurevich, and G. I. Bobrova, Thermal Diffusion Separation of Liquid Mix- tures [in Russian], Nauka i Tekhnika, Minsk (1971). 6. W. Rutherford et al., Rev. Sci. Instrum., 39, No. 1, 94 (1968). 7. M. P. Malkov et al., in: Transactions of the Second Geneva Conference, Vol. 10 [in Russian], Atom- izdat Moscow (1969), p. 54. 8. Ya. D. Zel'venskii, A. A. Raitman, and A. P. Timashev, Khim. Prom., No. 7, 538 (1973). 9. G. F. Barvikh and R. Ya. Kucherov, in: Transactions of the All-Union Scientific-Technical Con- ference on the Use of Radioactive and Stable Isotopes in the National Economy [in Russian], Izd. AN SSSR, Moscow (1958), p. 120. 10. I. G. Gverdtsiteli, R. Ya. Kucherov, and V. K. Tskhakaya, in: Transactions of the Second Geneva Conference, Vol. 10 [in Russian], Atomizdat, Moscow (1969), p. 69. 472 Declassified and Approved For Release 2013/09/25 : CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 MATHEMATICAL SIMULATION OF THE EXTRACTIVE REPROCESSING OF NUCLEAR FUEL 3. REDOX REEXTRACTION USING IRON SALTS* A. M. R9zen, M. Ya. Zel'venskii, UDC 621.039.59.001.57 and I. V. Shilin Processes of separative reextraction play an important part in radiochemical technology; they are used to separate plutonium and neptunium from each other and from uranium. However, the mathematical extraction model developed in earlier treatments [1-4] is insufficient for describing these processes, since it takes no account of the kinetics of redox reactions. The aim of the present investigation was to develop a mathematical model and a computing algorithm for the redox reextraction of neptunium and plutonium with due allowance for the kinetics of the chemical reactions in the aqueous phase, and also to analyze the influence of the basic parameters of the process on its efficiency, leading to a rational choice of the optimum conditions of operation. As redox reagent we took the Fe2??Fe3+ system, for which a relatively large number of kinetic data have been published; however, the algorithm developed in this connection is of a general character, and may be used for pro- cesses involving other nonextracted redox reagents. Kinetics of the Reduction of Plutonium and Oxidation of Neptunium by Iron Salts The reduction of plutonium (IV) by iron (II) proceeds in accordance with the reaction Pu4+ + Fe2+ Pu3+ + Fe3+, while the oxidation of neptunium (IV) by iron (III) obeys NO+ + Fe3+ + 2H20 Np04- + Fe2+ + 4H+. The kinetic equations of these reactions are *Parts land 2, see At. Energ., 37, No. 3 (1974). 2 iilles Reducing agent Up MO." Pa Np ? NO, 4 2175-EET?zisg gent Fig. 1. Arrangement of reextraction unit; 1) Flow of original organic solution; 2) flow of extraction agent; 3) aqueous flow from the first extractor; 4) flow of reextraction agent in the first and second extractors; 5) extract from the first extractor (also forming the original solution for the second) and from the second extractor; 6) reextract from the second extractor. Translated from Atomnaya Energiya, Vol. 38, No. 6, pp. 367-371, June, 1975. Original article sub- mitted July 18, 1974; revision submitted January 3, 1975. 0 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 473 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Losses of Np, Pu, %; U, g/liter 3 4 6 7 10-9 I t I I 45 qit 0,3 41 1,0 49 48 47 46 nnomin0 131 45 42 45 44 XHN Fig. 2 Fig. 3 Fig. 2. Loss of valuable components (Pu and Np,% of the original content) as a function of the ratio of the flows nnom/ n = L/Lnom for XHNO3 = 0.1M: 1)-4) losses of plutonium from the organic phase of the first reextractor; 5)-8) losses of neptunium from the or- ganic phases of the second reextractor; 9) losses of neptunium from the aqueous phase of the first reextractor; 10) losses of plutonium from the organic phase of the first_reextrac- tor for the ideal-displacement situation; 11)-14) losses of uranium from the aqueous phase of the second reextractor. (The figures oncurves 1-10 give the contact time in the mixing and settling chambers of the stage in min; those of 11-14 give the number of stages in the uranium preextraction section.) Fig. 3. Dependence of the neptunium and plutonium losses on the acidity of the reextrac- tion agent: 1)-6) losses of neptunium from the aqueous phase of the first reextractor; 7)- 11) losses of plutonium from the organic phase of the first reextractor; 12-16) losses of neptunium from the organic phase of the second reextractor. (Numbers on the curves indicate the values of L/Lnom = nnomAL) (Pu (III)] = I42 [Pu (IV)] ?k [Pu at d [Np (V)] --= ki?IP [Np (IV)] ? kn.[Np (V)]. at By approximating the experimental data of [5-9] we obtain equations for the constants icpu _ 1620 [Fe (II)] kpu kr' [Fe MIA (1+2.0 [Nail) [H+1 ' 2 600 [11+1 ' = exp ( ? 0.36811 + 0.345) [Fe (III)]/[1113; 14113-= exp (0.506R+ 3.1) [Fe (II)] [11], (3) where ? = 3 [UO2 (NO3)2] + [HNO3] is the ionic strength; [NOn = 2 [UO2 (NO3)2] + [HNO3] is the concentration of nitrate ions. 474 Declassified and Approved For Release 2013/09/25 : CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Losses of Pu ? 102,% 477- 4,6- 2 1 2 3 Tset, min 0 2 4 6 8 /0 /2 14 Nre Fig. 4 Fig. 5 Fig. 4. Influence of the time of contact in the settling chamber on the losses of plutonium from the organic phase of the first reextractor (L/Lnom = 0.8; XHNO = 0.2 M): 1) for ordinarymixer?settlers, Tmix = 1.5 min; 2) for centrifugal extractors Tmix = 0.2 min. Fig. 5. Dependence of the plutonium and neptunium losses on the number of stages in the reextraction section of the first extractor (L/Lnom = 0.8; XHNO3 = 0.2M): 1),2) neptunium and plutonium losses for a constant total time of contact in the extractor Nre Tmix = 15 min and Tmix: reet = 1:2; 3) plutonium losses on adding stages with a standard contact time (Tmix = 1.5 min, Tset = 3.0 min). Mathematical Description of the Process in Each Stage, and Algorithm for Calculating the Distribution of the Microcomponents with respect to the Various Stages of the Extraction Apparatus under Steady-State Conditions We shall assume that the reextraction process does not amount to a slow chemical reaction, and that its velocity, determined by mass-transfer processes, is fairly high; owing to the smallness of their dis- tribution coefficients, the extractability of Pu(III) and Np(V) may be neglected. We shall consider that the distribution coefficients of Pu(IV) and Np(IV) as microcomponents do not depend on their own concentration but are determined by the concentration of uranium and nitric acid in the aqueous phase. The mathematical description of the process at the i-th stage comprises the equations of material balance (allowing for the structure of the flows) (a), the equations of extractive equilibrium (b), and the kinetic equations of the redox reactions in the aqueous phase (c). The equations describing the process are exactly the same for both plutonium and neptunium (the only difference lies in the numerical values of the kinetic constants and the distribution coefficients) and take the following form: for the mixing chamber of the stage a) x5, i-i+ x5, i-t+ nYi+1 b) c) for the settling chamber of the stage Xs, i- ?ns, i- =0; ? - ? 1/4, i = CCiX4, i; dxs, jX4, - I? k2, 1X5, dt a) xy, f + X- 5, f X4, f X5, =0; b) ;4, i==U4,i; dx5, c) dt ?k1, x4, 1?k2, (4) (5) 475 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 0,4 -NO3 a Np 0 o Pc' 1 i 2 4 6 8 10 12 2 4 6 8 10 12 Number of stage Number of stage Fig. 6 2 4 6 8 10 12 Number of stage Fig. 7 2 4 6 8 10 12 14 Number of stage Fig. 6. Distribution of the components with respect to the stages of the first extractor in the aque- ous and organic phases (a and b respectively). Fig. 7. Distribution of the components with respect to the stages of the second extractor in the aqueous and organic phases (a and b respectively). where n is the ratio of the flows of the organic and aqueous phases, a is the distribution coefficient of Pu(IV) or Np(IV), x4 is the concentration of Pu(IV) or Np(IV), x5 is that of Pu(III) or Np(V) in the aqueous phase, y4 is the concentration of Pu(IV) or Np(IV) in the organic phase. The index i means that the quantity relates to the yield in the i-th stage; intermediate values (at the outlet of the mixing chamber) are denoted by a stroke over the symbol. The system (4), (5) reduces to a matrix equation describing the step as a whole: 44, 1+1 X5, i 1 A2/nct1A9 ? A4 ? 1)/nA9 (Aso + (1?A8)A6A2/A 9)/ai 0 44 (A8 ? 1)/(A6 ? 1) (A7 -I- A8A6A 2/A9)/Cti 0 ,48A4/A9 X4, X5, SI y4, i r,,t X5, i-i (6) _ Here A7 =k7, i +k21; A2 =1 + ain; A3 = ki,i/A2 +k2,1; A4 = e-A3Tmix and A5 = e -A1Tset for the ideal dis- placement situation; A4 = (1 + A3Tmix)-1 and A5 = (1 4- AlTsetr I for the ideal mixing situation; A6 = kl,t (1 ?A4)/(k1, t + k2, jA2); A7 = ki, i (1?A5)/Ai; Ag = (k2, jA5 + ki,i)/Ai; A9 = 1?A6; A10 = 1?A7, where Tmix and Tset are the times spent by the aqueous phase in the mixing and settling chambers of the stage. For the whole extractor the expression analogous to (6) is 0 24, Out X5, out SNSN-i ? ? ? Yinno ne 0 0 ?SN. o o ne 0 1 0 0 0 1 Sly, ? . . Si Y4, Mt 0 0 (7) where ne and nre are the ratios of the flows in the extraction and reextraction sections of the apparatus, no is the ratio of the flow of the reprocessed organic solution to the flow of the aqueous phase, NI is the number of stages in the whole extractor, with due allowance for the fictitious stage of feeding in the solu- tion to be reprocessed, which has the number NI + 1. In shortened form Eq. (7) appears thus: whence 476 0 x4, out X5, out ?Q-FR qt 42 43 Tit r21 r23 out r2irsi Y out= r11x4, out =42? qi x5, out = q3 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 (8) (9) Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Thus the algorithm for calculating the distribution of neptunium and plutonium over the various stages of the apparatus consists of the following; 1) calculation of the distribution of the macrocomponents, namely, uranium and nitric acid (for the methods of calculation see [4], which also gives the equilibrium equations); 2) calculation of the kinetic constants of the reactions by means of Eq. (5) and of the distribution coeffi- cients (see [4]) at each stage; 3) construction of the matrices Si in accordance with Eq. (6); 4) determination of the matrix R and the column Q in accordance with Eqs. (7) and (8); 5) calculation of the neptunium and plutonium concentrations in the flows of the aqueous and organic phases emerging from the extractor by means of Eq. (9); 6) successive calculation of the neptunium and plutonium concentrations in the two phases at each stage by means of Eq. (6), using the already prepared matrices Si. Calculations of the Redox Reextraction Process Using the Minsk-32 computer we calculated a reextraction system consisting of two extractors (Fig. 1); the reduction and reextraction of plutonium took place in the first extractor, and the oxidation and reextraction of neptunium in the second. The aim of the calculations was to study the influence of the ratio of the flows, the acidity of the aqueous phase, the number of stages, and the time spent in the mixing and settling chambers on the extraction of the valuable components. We also determined the range of condi- tions for both extractors in which the losses of uranium, plutonium, and neptunium were lowest. For the first extractor we took no = 8.0; re 1.5; "re = no + ne = 9.5 as nominal conditions. The original organic solution contained 0.38 M uranium and 0.17 M HNO3. The concentration of plutonium and neptunium in this solution was taken as unity. Into the preextraction section we fed a 30% solution of tri- butylphosphate (TBP) in synthine; reextraction was effected with 0.1 M HNO3 containing 0.04 g-ion/liter Fe(H) with an admixture of 10% Fe(III) (produced by self-oxidation). Into the second reextractor we fed the organic flow from the first reextractor (at 80% of the nominal rating in the latter according to the ratio of the flows), containing 0.32 M uranium, 0.0333 M HNO3, 0.83444 arbitrary units of Np, and 1.7 .10-4 arbitrary units of Pu. The nominal ratios of the flows were no = 11.875; ne = 1.125; nre = 13.0 The re- extracting 0.1 M HNO3 contained 0.04 g-ion/liter Fe(III). The ratio of the flows in the two reextractors was varied by varying the flow of the aqueous phase L[n nnomAL/Lnom)]. We see from Fig. 2 that an increase in the ratio of the flow increases the losses of the components from the organic phase (plutonium in the first extractor, neptunium in the second), but reduces the losses from the aqueous phase (neptunium in the first and uranium in the second). Whereas in the first reextrac- tor the uranium concentration in the reextract is no greater than 1.5 mg/liter for any of the operating conditions calculated, in the second reextractor a three-stage preextraction section fails to provide an acceptable (10 mg/liter) degree of purity of the reextract with respect to uranium (Fig. 2, curve 11), and the number of stages must be increased from four to six (Fig. 2, curves 12-14). Increasing the acidity of the reextraction agent increases the losses of the components from the aqueous phase (Fig. 3); hence it is desirable to conduct the whole process at a low acidity. The time spent in the stage has a substantial influence on the losses of neptunium and plutonium (Fig. 2), and due allowance for the kinetics of the redox reactions must therefore never be neglected. This may at first glance appear paradoxical, since the velocity constants are high and the relaxation time of the process (to = 1/k) should be short (for example, in the case of XHNO3 = 1 M and [Fe(II)] = 0.04 g-ion/liter; = 16.2 min-I and to = 1/kPu = 0.06 min, while Tmix = 1.5 min,i.e., 25 times greater than to). How- ever, one characteristic of the process in the two-phase system is a transition of the plutonium (neptunium) to a state of reduction (oxidation) on passing into the aqueous phase from the organic phase, in accordance with its distribution coefficient; the velocity constant then falls by a factor of (1 + an), while the relaxation time correspondingly increases. This leads to a considerable retardation of the redox processes in the extractive stage by comparison with the same process in the aqueous phase. The calculations also showed that, despite the occurrence of redox reactions in the mixing and settling chambers, the role of the latter was insignificant owing to the absence of mass transfer in these chambers even for small values of Tmix (Fig. 4). Influence of the Number of Stages. We found (Fig. 5, curves 1 and 2) that for a constant total time of contact in the apparatus (NreTmix = const) it was advantageous to have more stages with a shorter con- tact time in each. Increasing the number of reextraction stages while preserving their standard dimensions naturally reduces the losses of the components (curve 3). 477 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Structure of the Flows. A comparison between the results of calculations relating to the operating conditions of the process for the case of ideal displacement and ideal mixing (Fig. 2, curves 2 and 10) once again [2] indicates the considerable advantage of using piston-type flow rather than ideal mixing. The range of acceptable operating conditions is determined by the completeness of the plutonium, neptunium, and uranium separation required. Since many parameters of the process act in contradictory ways on the losses of these components, this range is comparatively narrow. For the first reextractor the best arrangement is that corresponding to 80% nominal (flow ratios no = 10.0; ne = 1.875; "re = 11.875) with a reextraction-agent acidity of no greater than 0.2 M. Under these conditions the plutonium losses are no greater than 0.05%, and the neptunium losses no greater than 1%. A shift in the direction of lower values of n leads to an increase in the losses of neptunium with the aqueous phase, while higher values of n lead to an increase in the losses of plutonium with the organic flow passing into the second reextractor. (The distribution of the components with respect to the stages in this condition is indicated in 'Fi6.) In order to widen the range of acceptable conditions it is essential to enlarge the reextraction section of the apparatus. The second reextractor should have at least four stages in the uranium preextraction section, the best working condition being that corresponding to 60% nominal (ratios of the flows no = 19.8; ne = 1.87; nre = 21.7). For an XHNO3 no greater than 0.5 M the neptunium losses then fall below 1.10-3% and the amount of uranium in the reextract is no greater than 10 mg/liter. (The distribution of the components with respect to the stages in this situation is indicated in Fig. 7.) Increasing the preextraction section to six stages leads to a considerable expansion of the range of acceptable conditions, which is then only restricted by upper limits of no = 30.0 and ne = 2.8; the losses of neptunium are no greater than 5.104%. We note that the ratios of the flows corresponding to the nominal operating mode of the first re- extractor (as well as values lower than these), together with a reextraction-agent acidity of no greater than 0.2 M, ensure an almost complete transition of the neptunium and plutonium into the aqueous phase in one extractor (the amount of uranium in this section will then lie below -1 mg/liter). Thus depending on the technological requirements either individual or combined separation of Pu and-Np from U may be achieved. Influence of Temperature. Since the activation energies for the reduction of plutonium and oxidation of neptunium (19.7 and 35.2 kcal/mole respectively [7]) are comparatively high, the velocity constants increase sharply with temperature, i.e., increasing the temperature reduces the losses of plutonium and neptunium, or alternatively reduces the contact-time (calculation shows, for example, that on conducting the process in centrifugal extractors, in which Tmix = 5-10 sec, to should be raised to - 40?C). LITERATURE CITED 1. A. M. Rozen et al., Third Geneva Conference, Paper No. 346 (1964). 2. A. M. Rozen et al., in: Liquid Extraction [in Russian], Khimiya, Leningrad (1969), p. 5. 3. A. M. Rozen, Yu. V. Reshet'ko, and M. Ya. Zelivenskii, in: Transactions of the Comecon Symposium on Research in the Reprocessing of Irradiated Fuel, Vol. 1 [in Russian], Czechoslovakian Atomic Energy Commission, Prague (1972), p. 118. 4. A. M. Rozen, Yu. V. Reshettko, and M. Ya. Zellvenskii, At. Energ., 37, No. 3, 187 (1974). 5. T. Newton et al., J. Phys. Chem., 64, 244 (1960). 6. J. Huizenga and L. Magnusson, J. Amer. Chem. Soc., 73, 3902 (1951). 7. V. S. Koltunov, Kinetics of Redox Reactions of Uranium, Neptunium, and Plutonium in Aqueous Solution [in Russian], Atomizdat, Moscow (1965). 8. G. Best, J. Inorg. and NucL Chem., 12, 136 (1959). 9. M. Germain, ibid., 32, 245 (1970). 478 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 TWO-DIMENSIONAL DIFFUSION PROGRAM HEXAGA II FOR MANY-GROUP CALCULATIONS OF HEXAGONAL LATTICES T. Apostolov and Z. Woznicki UDC 621.039.51 The HEXAGA II program uses a uniform triangular difference mesh employed for the calcula- tions for various reactors of basically the water-moderated water-cooled type. The programiswritten in FORTRAN IV for the systems SYBER-70 and EC-1040. HEXAGA II enables one to solve diffusion equa- tions in the approximation from two to ten groups with allowance for diffusion of neutrons from high to low energy groups. Neutron scattering accompanied by an increase of energy is taken into account when the model is usedwith two or more thermal groups. The first variant of the program calculated the neutron fluxes at 5000 points of the reactor lattice, and for the subsequent variants this number increases to 10,000 and more. The many-grouped model of neutron diffusion is a search for a solution of the system of adjoint elliptic second-order partial differential equations ? div [Dg grad Og] Egavg ? 2 zg'-gog' Fe-gog', g=1, 2, .. g'=1 gr*g (all the notation in (1) is standard and given in [1]). The extraction cross section for the given group, Eg, is represented by the equation Eg = 15 ? 2 g'*g Equation (1) is augmented by the boundary conditions on the surface of the reactor: Dg acag czg0g =0 On (the derivative is along the outer normal to the boundary of the region). Equations (1) and (2) are solved by the source-iteration method. For the numerical solution of the problem in the group, one uses a finite-difference approximation in the triangular lattice. Thus, finite- difference equations are obtained and these can be represented for each group by a system of linear equa- tions: (1) (2) /10=c, (3) where A is a nondegenerate matrix of coefficients; 4. is the required vector of the solution for the given group; c is the known source vector, whose components take into account the processes of fission and scattering for the given group. For the solution of such linear systems, factorization iterative methods have been developed, for example, the method of Buleev [2], OLIPHANT [3, 41, STONE [5,6], DUPONT [7], and others. In [8], factorization iterative methods called two-pass iteration methods were developed. The essence of such a method, used in the HEXAGA II program to solve a system of linear equa- tions [2], is as follows. Institute of Nuclear Research and Nuclear Energy, Sofia, Bulgaria. Institute of Nuclear Research, Swierk, Poland. Translated from Atomnaya gnergiya, Vol. 38, No. 6, pp. 372-374, June, 1975. Original article submitted February 28, 1974; revision submitted November 12, 1974. ? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 479 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 AVAM &VA\ IVA EWA AV1YAYAYAWA\ I I #AVAVAVAWAA YAYAWAYMMAY V VA I IA WAY/ //AWN A kVA IlrAv vox/ I FAVA r V WA I AV L'M YA1 A A WA1 MVA VA 1 I MVAI * Avean NA %VA I I IN V vNAWA IAT VIVA I MVO! WAVAV/ V Fig. 1. Dtfferent arrangements of the hexagonal lattice relative to the central axis of the reactor. We represent the matrix A in the form A=K?L?U, where K = diag(kii) 0 is a diagonal matrix in which kii > 0 for all 1 s i s n; L = (1ii) 0 is the lower triangular matrix in which /ii = 0 for all i s j; U = ?/0 is the upper triangular matrix in which uii = 0 for all i j and (4) Mil + Uji ) for all 1 i n, and for at least one i there is a rigorous equation which guarantees nonnegativity of the matrix A and the existence of A-1 0. The matrix A can be represented as A=K?PA?(L+H)?(U+Q)-1-PA+11?Q, (5) where P A = diag(pii) 0, in which pH.? 0 for all 1 < i s n; H = (hij) 0 is the lower triangular matrix in which h?? = 0 for all i 5 j; Q = (qii) ? 0 is the upper triangular matrix in which ? = 0 for all j. ? ij Further, one assumes K >PA >0, ? where kit > pii > 0 for all 1 s i s n. Then DA=K?PA?>.0 is a nonnegative diagonal matrix in which DA-1 0 for all 1 i 5 n. After transformations, we obtain A=DA.?(L+H)?(U+Q)+PA+11-1-Q- In what follows, we use the equation DA? (L H)?(U-FQ) = [I?(L + H)I:4] DAV ? (U-I-Q)] ?(L- H) (U Q). The nonnegative matrix (L + H)D7A!(U + Q) can be expressed by the sum + H>rrAl + Ri +TA+Hid-Qi, where R1 = diag [(L + H) D (U + Q)]. If we choose matrices PA, H, and Q such that PA = R1,H =111, and Q = Q1, then A= M NA, [I?D-A1(U + Q)] and NA = TA, and where MA = 1-1?(L + H)D-1 480 Nil >0 andNA >0. Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 (6) (7) (8) (9) (10) Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 It is known from Varga's theorem [9] that the iteration matrix = r A has spectral radius p(MNA) < 1 and the iterative method Oh' ? M.74INA4:Di />0 (where j is the number of the iteration) converges for the solution of a system of linear equations for all vectors In the practical use of the two-pass iteration methods one uses recursive formulas for some auxiliary vector p, after whose calculation the vector ,t(J +1) is determined. In our case the vector 413(J +1) is expressed in the form = [/?D; (U ())1"--, x [I ?(L ?H) T AcDo +11 (U + Q)1-' D7.1' [I? (L +11) D2]-, c. where j 0. After multiplication of (12) by the matrix [I?D-11: (U + Q)], we obtain =DA1 ((ti +0 Oh' ? II) D;i1-1 (T AO(' c)}. We introduce the vector p(i +1) : 10-o= [I? (L + H) DAT, (TAD" c). After multiplication of (14) by the matrix [I?(L + H)1)-11, we obtain 10-0= (L li) D0+1) T A4:1)01) c; _D [(u +0 (Do -I) (12) (13) (14) (15) This approach is none other than the elimination method of Gauss and is equivalent in this case to the sweep method for three-diagonal matrices. To accelerate the convergence, the method of upper relaxation can be used. In [8] this method is compared with the Gauss?Seidel method and it is shown that where L1 is the iteration matrix in the Gauss?Seidel method: Li? (I ? (16) (17) In the determination of the region of the solution in the program it is borne in mind that the majority of reactors with hexagonal lattice of the core have angular symmetry (every 120?). Such an approach is used in reactors of the water-moderated water-cooled type. Figure 1 shows three arrangements of the hexagonal lattice relative to the central axis of the reactor. The neutron fluxes at arbitrary but symmetric points in the regions I, II, and III also have the same values. Thus, to describe the lattice one need consider only one of the regions and it is not necessary to determine the logarithmic boundary conditions on the axes X and V, whose values are obtained from the symmetry condition. It is only necessary to determine the logarithmic conditions on the boundaries parallel to the X and V axes lying within the core, on its boundary or outside it. One also considers a variant using logarithmic conditions on four boundaries of the region (Figs. 2 and 3). In one of the program variants, the regions may have the form of a parallelogram (see Fig. 2; such a form of the region is used in the program PDQ-7); in others, they can have the form of a hexagon (see Fig. 3), which simplifies the problem and makes the use of programs convenient. Besides the described lattice form, one also considers a combined configuration of hexagonal and oblique-angled polygons, and also a combined configuration of hexagonal lattices and lattices with (r, 0) geometry. The combination of such lattices enables one to represent more accurately the actual geometry and the material properties of the reactor lattice. However, the programming of the problem is then much more complicated. 481 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 0 411411= 410111 S.A=1157# AI 11116/41/AvirMAI 41FMAvilIrainvIrIr irAmmirilioritioril IMMINVIIMAIM IrffiririONFIRIIMir IFAIMMAMIDVAII V Fig. 2. Regions of the reactor having the the form of a parallelogram. ININIAVAIPLWIIVW MA I FA IMINI IM V MOMAVIM 011121/4.1FAINNWAVAVIN NA I IA III !AV/ kVA MAI 0411110 VAVA I MA I VA 1141 I PIM I IP IMAISMI I NW PA 0110/41 121 VA 112 V Fig. 3. Hexagonal form of reactor regions. Two-pass iterative methods were used in the two-dimensional programs EWA II [4,6] and AGA II to make a calculation for reactors with rectangular lattice on the GIER computer. The results show that the rate of convergence of these methods is 5-10 times faster than methods based on the Gauss?Seidel com- putational model. It is to be expected that the use of two-pass iteration methods of solution of the diffusion equations for a hexagonal lattice will increase the rate of convergence compared with the case of a rectangular lattice. However, to confirm this conjecture concrete results are necessary; these are expected very shortly. It should be pointed out that the number of arithmetic operations for one point in one group and one iteration when one uses the two-pass method and the method of upper relaxation (of the HEXAGA II pro- gram) exceeds ten operations of multiplication and ten operations of addition. When the Gauss?Seidel method is used, there are eight operations of multiplication and eight of addition. It is assumed that if a two-dimensional program is used for cylindrical and flat (y, z) geometry (30 groups and 4000 points), it will be possible to calculate the group geometric parameters B2z needed for HEXAGA II. The algorithm EWA II is based on the modified two-pass iteration method. The program is written in FORTRAN IV for the IBM-360/40 computer. LITERATURE CITED 1. G. Habettler and M. Martino, in; Proc. Symp. on Appl. Math., Amer. Math. Soc., Providence, Rhode Island, 11, 127 (1961). 2. N. Buleev, Matem. Sb., 51, 227 (1960). 3. T. Oliphant, Quart. Appl. Math., 20, 257 (1962). 4. T. Oliphant, ibid., 19, 221 (1961). 5. H. Stone, Trans. Amer. Nucl. Soc., 13, 180 (1970). 6. H. Stone, SIAM J. Number. Anal., 5, 530 (1968). 7. T. Dupont, ibid., 753. 8. Z. Woznicki, Doctoral Dissertation, Computing Center CYFRONET, Institute of Nuclear Research, Swierk (1973). 9. R. Varga, Matrix Iterative Analysis, Prentice Hall (1962). 482 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 INVESTIGATION OF THE ADIABATIC OUTFLOW OF WATER THROUGH CYLINDRICAL CHANNELS V. S. Aleshin, Yu. A. Kalaida, UDC 621.039.001.5 and V. V. Fisenko Partial evaporation of the fluid along the channel length occurs during the outflow of saturated and underheated water with high initial parameters through a cylindrical channel with pointed entrance edges, and the stream is two-phasal at the exit from the channel. A large number of papers, including the surveys [1,2], is devoted to the investigation of the adiabatic flow of saturated water. However, up to now there have been no reliable methods of estimating the weight discharge for a broad range of initial parameters and the total picture of the flow of evaporating water in cylindrical channels has been studied inadequately. This is explained by the complexity of the heat and mass exchange processes between the liquid and vapor phases flowing in the channel and by the influence of various factors (the initial parameters, the //d ratio, where / is the channel length and d is its diameter, the counter-pressure at the exit, etc.) on the outflow processes. The results of investigating the outflow of hot water through cylindrical channels of 5 and 9.53 mm diameter with pointed entrance edges and an //d ratio between 0.5 and 18 are examined in this paper as the pressure PI changes ahead of the outflow channel between 25 and 150 kg/cm2 and the water underheating Ats is 0-100?C up to saturation. 2 3 4 5 50 100 150 200 P1. kg/cm2 Fig. 1 Fi g. 2 Fig. 1. Schematic diagram of the apparatus: 1) disposal section; 2) valves to produce counterpressure; 3) probe; 4) outflow valve; 5) quick-shutoff valve; 6) hot-water chamber; 7) high-pressure air system. Fig. 2. Dependence of the stream discharge on the initial pressure for different Ats (d = 5 mm, //d = 0.5): 1) discharge of cold water at t = 18?C; 2) rated dis- charge of dry saturated vapor; 3) rated discharge of saturated water (x = 0). Atomnaya Energiya, Vol. 38, No. 6, pp. 375-378, June, 1975. Original article submitted April 30, 1974. ? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 483 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 pl;t, /////////y/////7///0_ 200 197 15 t151 4 ,55 -20 0 20 40 60 80 1,mm 3 9 lid Fig. 3 Fig. 4 Fig. 3. Change in medium parameters (p and t) along the channel length for d = 9.53 mm, //d = 9.55, 6ts = 20?C for different initial pressures pi, kg/cm2: 1) 100; 2) 75; 3) 50; 4) 25. Fig. 4. Change in parameters of the medium (p and t) along the channel length for pi = 75 kg/cm2,Lt = 20?C, d = 9.53 mm for different //d: 1) //d = 9.55; 2) //d = 6; 3) lid =3; 4) //d = 0.5. 270 260 231 195 \ 4,t, \ v////////// /////////// ////% The experimental apparatus (Fig. 1) consists of a 350 liter (hot water) chamber to obtain hot water at pressures to 150 kg/cm2 and a temperature to 350?C; an outflow section of telescope type, which permits altering the working channel without disassembly of the apparatus; a tank for disposal of-the-outflowing medium through a throttling unit under a layer of water; a high-pressure air system to maintain a given pressure ahead of the outflow section and to assure its constancy during the outflow process. The hot water chamber was filled with water purified in special filters, and the water was deaerated before each test. The cold water hydraulic characteristics were recorded for each of the channels before the tests. The error in measuring the temperature and pressure in the hot water chamber, at the entrance to the outflow channel, and at its exit was no higher than ?5%. The stream static pressure and temperature along the outflow channel length were measured by using a probe. The impulse hole to determine the pressure and the microthermocbuple imbedded in the probe permitted a record of both the continuous diagram along the channel length, and the parameters at fixed points. The discharge of the outflowing medium was measured with a flowmeter with a continuous recorder tape, which is important in investigations of the critical flow modes. Weight Discharge of Water. To determine the influence of the initial pressure pi, the degree of underheating of the water below boiling Ms, and the //d ratio on the weight discharge characteristics, a series of tests was conducted in which only one of the factors listed above was altered while the other conditions remained equal. Presented in Fig. 2 is a graph of the dependence of the ratio between the weight discharge and the channel cross-sectional area on the pressure ahead of the entrance to the outflow sec- tion for a channel with //d = 0.5 (the dashes show the assumed weight discharge at an initial pressure above 150 kg/cm2). It follows from Fig. 2 that the experimental discharge characteristics almost agree with the hydraulic characteristics up to a 70-75 kg/cm2 pressure for any degree of underheating. The vapor for- mation in the stream becomes so intense at pressures above 75 kg/cm2 and a 0-50?C degree of underheat- ing that the presence of the vapor phase noticeably diminishes the stream density and reduces the weight discharge. 484 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 80 70 60 50 449 30 20 10 20 40 60 80 100 Pep. kg/cm2 Fig. 5 Fig. 5. Dependence of grel and p2 on the = 7.26, Ate = 20?C and different pi, kg/cm2; 1) 100; 2) 75; 3) 50; 4) 25. Fig. 6. Dependence of e = p2/pi on //d, Ate, and pi. I 1t s0 s 451 444 1 8 42 90 100 41 46 At 0 45 ??, b 60 70 4 180 90 4 qv It iiti t 'III 2 4 6 8 150 125 100 80 60 Lid p, kg/cm2 Fig. 6 counterpressure pep for d = 6.4 mm, //d 20 Analogous graphs have been obtained for weight discharges for other //dratios(0.5; 2.0; 3.0; 4.0; 5.0; 6.0; 9.0; 18.0). It follows from the graphs that a deviation of the experimental from the hydraulic curves sets in at a lower pressure as //d increases. Thus, for //d >6-8 the diminution in discharge starts at pr-=,' 25 kg/cm2 as compared with the hydraulic value. Pressure and Temperature Distribution. The temperature and pressure along the channel axis were measured by displacement of a probe. The investigations were carried out in 9.53 mm channels for an //d between 0.5 and 9.55. More than 100 combined diagrams were obtained. Some are presented in Figs. 3 and 4. Analysis of the test curves shows that three characteristic sections are formed along the channel length in channels with //d > 6-8. An abrupt pressure drop below the saturation pressure corresponding to the initial temperature is observed in the first section in the area of the entrance edge, and intensive vapor formation occurs, whereupon the temperature of the medium drops. A certain lag in the temperature drop as compared with the pressure reduction as the water moves (particularly for saturated water) indi- cates its possible overheating and a metastability of the process in progress. Then the velocity of the stream motion diminishes as the stream broadens beyong the leading edge, the pressure rises and the process of vapor condensation is observed, as the rise in stream temperature indicates. The nature of the pressure drop depends on the initial temperature. The higher the stream tem- perature, the smaller the fraction of the pressure drop at the entrance section and the greater the fraction at the exit of the channel. Let us note that the phenomenon of a sharp pressure drop and its subsequent rise at the leading edge has not been observed in researches performed earlier either because of the low initial parameters (a pressure less than 25 kg/cm2) or of the absence of a continuous record of the pressure and temperature along the channel length. Depending on the initial values of pi and ti, either a recovered hydraulic stream of saturated water at a temperature corresponding to the initial temperature (for Ate > 20?C) or a two-phase stabilized stream with a temperature somewhat lower than the initial value (for Ats < 20?C) flows in the second section, the section of stabilized parameters. As the underheating is reduced, the vapor content increases in the stream and reaches a maximum at the outflow of the saturated water. The lower the initial water tem- perature, the closer will the stabilized parameters section be to the exit section. The vapor formation 485 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 process starts again in the third section of the channel as the stream approaches the exit edge; the stream temperature also drops as the pressure drops. The pressure drops to p2 in the exit section, a value greater than atmospheric and dependent on the initial parameters. Thus, for example, as pi rises from 25 to 100 kg/cm2 at Ats = 20?C the pressure p2 grows from 14 to 57 kg/cm2, and the pressure ratio a = p2/pi remains practically constant. A steady, equilibrium two-phase stream for which a r-z1 0.55 flows out of the channel. As the underheating increases because of the diminution in the vapor phase content in the stream, the quantity a drops. Thus, for Ats = 59?C, a is reduced to 0.334. Shown in Fig. 4 is the nature of the change in the stream p and t along the channel length as a function of //d for saturated water flowing out at an initial pressure of 75 kg/cm2. The nature of the process in long channels with //d > 6-8 is described above. Diminution in the channel length is accompanied by the growth in metastability of the stream, by a rise in its density, and by an increase in the weight discharge. Thus, for //d = 0.5 the stream turns out to be overheated 45?C in the exit section. The pressure at the exit from the channel drops more sharply in short channels. It is characteristic that there may be a sec- tion with stabilized parameters even in short channels with //d = 3.0. It also follows from Fig. 4 that the pressure drop along the channel length is accompanied by an abrupt reduction in temperature, and therefore, by a continuous vapor formation process. Investigation of the Critical Modes. A series of tests was conducted in channels of 6.4 mm diameter with //d = 0.5-7.26 and Ats = 0-100?C to investigate the critical flow modes of boiling water. The initial pressure varied between 25 and 150 kg/cm2. Processing the test results permitted construction of depen- dences of the relative discharge grel = gep/go (go is the discharge for a stream outflow into the atmosphere, and gcp is the discharge at a variable counterpressure pep) and of the pressure p2 at the exit section on the counterpressure. One of the graphs is presented in Fig. 5. It follows therefrom that the passage over to the critical mode occurs more smoothly as boiling water flows out, the pressure at the exit edge p2 tends to some constant value for a continuously diminishing counterpressure pep. The time of the onset of the flow crisis agrees with the time of build-up of a constant pressure in the exit section of the channel (the beginning of the critical flow mode is shown by dashes). This indicates that the critical channel section coincides with the exit, and the steady pressure ratio p2/pi is critical. The dependences 132 = f (pcp) and grel =f (pep) obtained permit establishment of three characteristic boiling water flow domains in the channel (the domain boundaries are shown for an initial pressure of pi = 100 kg/cm2). 1. The critical mode domain (A). The weight discharge is a maximum in this domain and is in- dependent of the counterpressure. 2. The near-critical domain (B). As the counterpressure increases in this domain the pressure in the exit section rises monotonely, remaining greater than pep. The discharge decreases in- significantly. 3. The third domain (C) is provisionally called the domain of the pseudohydraulic flow mode. It sets in at the time the pressure in the exit section equals the counterpressure and is retained down to total equalization of the pressure (pep = pi). The discharge decreases sharply in this domain and is zero for pep = Tests conducted with unchanged pi = 75 kg/cm2 and Ats = 0-100?C showed that the crisis in the dis- charge sets in atlowercounterpressures as Ats increases, i.e., e =P2/PI diminishes. Processing the numerous experimental results permits construction of a nomogram (Fig. 6) of the dependence of a on the initial pressure, the degree of underheating of the water to the saturation state, and the relative channel length. The dashes show the sequence in determining a for a given //d ratio and initial Ats and pi. The following dependences can be recommended to determine the specific weight discharges by generalizing the test results. 1) For //d > 0, Ats 3. 494 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 000 o) 8 0 8 coo 0 0 0 00 0 0 10 0? 1 0 C60 oPo 6b00 0 I 80 ? 0?Q0 1 0 20 401 a 0,02 t, sec Fi g. 6 /2 t, sec 240 Z60 280 Energy of ions, MeV Fig. 7 Fig. 6. Distribution of tracks of fragments of spontaneous fission in the reaction of 207pb 50Ti: a) time cycle of measurements from 0.003 to 0.03 sec; b) from 1 to 13 sec. Fig. 7. Integral functions of excitation (a) for the isotopes 131Pm (e) and 2?4111Pb (C), formed in the reaction 2081T1Pb + --- and ?) results of a calculation with parameter of diffuseness d = 0.44 -10-13 and d = 0.34 .10-13 cm, respectively (the value of the remaining parameters are the same as in [12]). Integral functions of excitation (b) for the spontaneously fissioning emitter with TT 2 "-=1 5 msec: ?) calculated values of the excitation for the reactions ("Ti, 2n) an ("Ti, 4n) at d = 0.34 -10-13 Cm. It was shown earlier [12] that reactions with the emission of two neutrons are very sensitive to the value of the minimum energy of excitation of the compound nucleus. The transition from 48Ti ions to "Ti ions, according to our estimates, increases the cross section of 2n-reactions approximately 100-fold. In view of this, a beam of accelerated "Ti ions was used in the experiments. Production of Accelerated Titanium Ions The calculated value of the barrier to interaction in the reaction 238Pb + "Ti is equal to 230-235 MeV in the laboratory system of coordinates. The maximum energy of the beams of accelerated ions for the 300 cm cyclotron of the Laboratory of Nuclear Research, United Institute of Nuclear Research, is deter- mined by the ratio Emax = 250 Z1 /AI, and for 50Ti+7 and 50Ti+4 is 245 and 320 MeV, respectively. For the acceleration of "Ti ions in such a high charge state, we created a special ion source, in which the enriched isotope "Ti in the form of the metal was used as the starting material. The intensity of the isolated beam of 50Ti8 ions was 5 -1013-1011 particles/sec, while the intensity of the internal beam of 50Ti+8 ions was 2 -1011 particles/sec. The energy of Ti+7 ions is only 10-15 MeV greater than the barrier to interaction; therefore, the isolated beam was used to study the reactions pro- ceeding with the emission of two neutrons, and the bulk of the experiments was conducted on the internal beam of the cyclotron. Experimental Procedure In experiments with "Ar ions [12] it was shown that when isotopes of lead are used as the target, there is practically no background associated with the spontaneous fission of side products of the reaction. 495 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 6 - 18 144 148 152 156 Fig. 8 Fig. 8. Systematics of the periods of spontaneous fission the isotopes 250102, 255104, and 256104. Fig. 9. Illustrative representation of the barriers to fission for the isotopes Fm, 254102, and 256KU. (The periods of spontaneous fission are given along the y axis.) 1 year 1 day 1 h 1 sec 1 msec 1 ?sec 1 nsec 160 N /44 152 Fig. 9 considering new data for 150 Therefore, for the synthesis of heavy elements by this method, a highly sensitive rapid method for the detection of nuclei according to spontaneous fission can be successfully used. In the experiments we used the distribution of 205Pb isotopes (content 97.8%, impurities 207Pb 1.6% and 206Pb 0.6%); 207pb (content 83.4%, impurities 208Pb 14.3% and 206Pb 2.3%); 206Pb (content 90.4%, im- purities 208Pb 6.7% and 207Pb 2.9%)? The experimental setup for the recording of short-lived spontaneously fissioning nuclei is schemati- cally presented in Fig. 4. The beam of 50Ti ions drops along a tangent to the surface of a hollow cylinder, oriented vertically and rotating at a maximum angular velocity of 3000 rpm. By the method of atomiza- tion, a layer of lead about 2 mg/cm2 thick, simultaneously serving as the target and the collector of recoil nuclei, was applied on the side surface of the cylinder. As a result of the fact that the beam is incident at a small angle to the surface of the cylinder, the layer of lead is an "infinitely" thick target, in which there is an integration of the function of excitation from Bint to Emax. The maximum energy of the internal beam of 50Ti ions was selected equal to 260 MeV (see the excita- tion function in Fig. 3). The intensity and position of the beam of ions were monitored during the experi- ment with a special device, situated outside the target. The integral flux of ions that passed through the target was determined by an activation method according to the yield of the isotope 117In (T112 = 2.8 days), formed in the bombardment of copper with titanium ions; for these purposes, a copper target ? 15? thick with an area of 1 cm2 (-1% of the total area of the lead target) was situated on the side surface of the cylinder. Track detectors of spontaneous fission fragments of mica with a content of uranium and thorium impurities < 10-7 g/g were positioned around the rotating target at a distance of 2 mm. Despite the fact that the energy of the recoil is comparatively great (? 50 MeV) and corresponds to a range in lead of ? 6 mg/cm2, in view of the small angle of entrance into the target these nuclei are situated close to the sur- face of the layer, and the efficiency of the recording of spontaneous fission fragments is about 50%. Shields protecting the detectors from scattered ions and forced fission fragments, as well as the special method of treatment of the detectors, entirely eliminated the background at a distance of only ?2 cm from the site of incidence of the beam. 496 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 This method was tested in experiments [12] on the production of the isotopes 244Fm in the reaction 206, 207, 208pb (40 ? -, dA xn)2?Fm, where it was shown that it can be successfully used for recording short-lived spontaneously fissioning emitters, if their lifetime is greater than 3 msec and the cross section of their formation exceeds 10-35 cm2. Synthesis of the Isotopes 256Ku and255Ku The isotope 258Ku can be obtained in the reaction 208Pb(50Ti, 2n). From the systematics of the proper- ties (see Fig. 1) it follows that the lifetime of this isotope with respect to a decay ranges from 50 to 200 msec, whereas the partial period of relatively spontaneous fission is substantially greater on account of ?the stabilizing action of the subshell N = 152. Therefore, in the reaction of 206Pb + 56Ti, we can count on the recording of spontaneous fission of the isotope 252102 (T112 c=.: 2 sec, 30% spontaneous fission), which is formed in the a decay of the 256KU nuclei. In the first experiment the rate of rotation of the cylinder was selected equal to 8 rpm, which per- mitted recording of the spontaneous fission with T112 > 0.5 sec. In the case of irradiation of 20813b with an integral flux of 8 -1015 56Ti ions, 12 tracks of spontaneous fission were observed, which is substantially less than the expected value. This may be evidence that either the mechanism of the fusion reaction changes substantially from 40Ar ions to 50Ti ions, or the properties of the isotope 256KU differ substantially from those predicted. To test the second hypothesis, the experiments were repeated at a rate of rotation of the cylinder of 1500 rpm. At an integral flux of 50Ti ions ?1015 particles, 70 tracks of spontaneous fission fragments, with the time distribution presented in Fig. 5, were recorded. Subsequent experiments were conducted with a target of 207Pb at a rate of rotation of the cylinder 1500 rpm. At an integral ion flux of ?1.2 -1015 particles, 53 spontaneous fission fragments were recorded (Fig. 6a). Evidently the half-life is significantly greater than the selected time value, and therefore the experiment was repeated at a lower rate of rotation (54 rpm). Here also an analogous picture of the dis- tribution of tracks was obtained; therefore, in the third series of experiments the rate was lowered to 3.6 rpm (Fig. 6b). Thus, it follows from the experiments that in the formation of 207Pb by 50Ti ions, the formation of a spontaneously fissioning emitter with a half-life of several seconds is observed. Finally, in the last series of experiments we irradiated a target of 20613b. At an integral flux of ?0.4 -1016 ions, only four tracks were recorded, half of which, taking into account the data obtained earlier, may be due to reactions on impurity isotopes 207Pb and 206Pb. In other words, in the combination of 2061,b + 50Ti, only the upper limit of formation of spontaneously fissioning nuclei can be determined. All the experimental results are presented in Table 1. The error in the determination of the values of the cross sections, according to our estimates, do not exceed a factor of two, whereas the relative error is determined chiefly by the statistical accuracy and is ?30%. Thus, from the aggregate of data obtained it follows that in the case of irradiation of targets of separated lead isotopes with 50Ti ions, the formation of two spontaneously fissioning emitters with greatly differing half-life: about 5 msec and several seconds, is observed. The emitter with half-life 5 msec, observed in the reaction of 206Pb + 50Ti, cannot be assigned to nuclei with atomic number Z < 104, since all the isotopes of element 102 are known up to N = 146 and do not possess such properties [16], while for the odd isotopes of element 103 there is a strong prohibition of spontaneous fission [17]. The maximum yield of this emitter is observed in the reaction with 208Pb; the effect decreases by more than 10-fold with a target of 207Pb and is practically entirely absent for 206Pb. Thus, analyzing the experimental cross sections of the reactions and the properties of the known isotopes of kurchatovium and lighter elements, it can be assumed that the observed effect is due to decay of the isotope 256Ku, which is formed in the reaction 208Pb(50Ti,2n)256Ku. The question of the properties of the isotope of kurchatovium with N = 152, as will be evident from the following, is of theoretical significance; therefore it is important to accurately identify the mass num- ber of the emitter with half-life ?5 msec. Since there is a great prohibition of spontaneous fission for odd isotopes, possible candidates are the even nuclei 256Ku and 258Ku, which are formed in reactions with the emission of two and four neutrons, respectively. 497 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Despite the fact that the calculated values of the cross sections of reactions with the emission of two and four neutrons differ substantially from one another (see Fig. 3), separate experiments were conducted on the measurement of the integral function of excitation for the emitter with T1/2 ?=-,. 5 msec. Figure 7 presents the experimental values and the calculated dependences of the integral cross sections of forma- tion of the isotope with half-life ?5 msec and the nuclei 151Pm (the fission fragments of the compound nucleus) and 234mPb (products of a transfer reaction) in the irradiation of 238Pb by 53Ti ions. The depen- dences presented in Fig. 7 additionally confirm the fact that the short-lived emitter is formed in a reaction with the emission of two neutrons and is the isotope 253Ku. Comparing the yield of short-lived and long-lived activity in the reaction of 238Pb + 53Ti, we can con- clude that the isotope 256KU in the majority of the cases experiences spontaneous fission. It should be noted that the experimental cross section of the reaction Pb(53Ti, 2n)Ku proved to be 10 times lower than the theoretical values (see Fig. 3). However, in a calculation of o-(x,n), the empirical dependence of Sikkeland et al. [18], based on the assumption of an influence of the subshell N = 152 on the value of I'll/I' f, was used for the ratio of the partial width rn/rf. And yet, as will be shown later, this assumption is not substantiated for nuclei with Z 104. The exclusion of the influence of the subshell N = 152 on the value of rn/rf leads to a significant decrease in the calculated cross sections, which virtually eliminates the discrepancy noted above. The long-lived emitter with half-life about 4 sec, in all probability, is the isotope 255KU, which is formed with a maximum cross section in the reaction 237Pb(53Ti, 2n), and with lower probability in the reaction 238Pb(53Ti, 3n), and is absent in the reaction 236Pb(53Ti, 1n). However, we should note that the lifetime of this emitter proved close to the lifetime of the known isotope 252102, which in this combination might be obtained in the reaction of the type of 237Pb(53Ti, aln)- 252102. Let us consider separately the question of the probability of side processes in reactions with ions of the type of "lir or 53Ti. At a maximum energy of the bombarding 53Ti ions equal to 260 MeV, the maximum energy of excitation of the compound nucleus 257KU is ^'40 MeV. Estimates show that the threshold of the reaction with evaporation of an a particle and one neutron is equal to ?260 MeV. Moreover, in this region of nuclei the ratio of the partial width ra/ rn, like rp/r n, is small 10-2). Therefore, the contribution of the processes occurring with the evaporation of charged particles from the compound nucleus is negli- gible. And yet, the process of direct emission of a particles, followed by evaporation of a neutron, as occurred in reactions with light ions of carbon, nitrogen, and oxygen, is energetically possible. However, as was shown in [19-21], the mechanism of such reactions depends substantially on the structure of the bombarding ion. In the sequence from light ions to particles of the type of 43Ar and 53Ti, the probability of the escape of direct a particles decreases hundreds and thousands of times. Model experiments were conducted in which spontaneously fissioning nuclei of 244FM were obtained in reactions in the irradiation of the isotopes 235T1 and 233T1 with 45Sc ions and of the isotopes 266Fb with 43Ar ions. As can be seen from Table 1, the ratio of the cross sections ?(a, 2n)/o- (2n) 0.03. From this it can be concluded that the probability of formation of the isotope 252102 in the irradiation of 237Pb with 53Ti ions is negligible. From a comparison of the cross sections of formation of the isotopes 255Ku and 256KU it follows that the odd isotope 255KU has comparable values of the half-life with respect to a decay and spontaneous fission. Finally, the absence of an effect in the reactions of 23813b + "Ti may mean that the life-time of the even isotope 254Ku is less than 3 msec. DISCUSSION OF RESULTS Let us consider the question of how the properties of the synthesized neutron-deficient isotopes of kurchatovium agree with the known data on the stability of heavy nuclei with respect to spontaneous fis- sion. For the isotope 256Ku with N = 152, instead of the expected half-life, comprising tens and hundreds of second, the experimental value was 103-104 times smaller. 498 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 The position of this point in the systematics of spontaneous fission substantially changes the concept of the stabilizing influence of the subshell N = 152 in the transition from Z = 102 to Z = 104. The depen- dence of TSF for kurchatovium isotopes on the number of neutrons is evidence of a monotonic increase in the lifetime of the even?even nuclei with increasing mass, without any significant variations in the region of N = 152 (Fig. 8). The lifetime of the odd nucleus 255KU is determined by the prohibition of spontaneous fission, which in this case is ?103. Now it is not surprising that the isotope 260Ku has TSF 0.1 sec, and there is no 1/2 need to assume such high prohibitions (-107-1012) for the other odd isotopes of kurchatovium with mass numbers 257 and 259, as was done in the form of Ghiorso et al. [6, 11]. From the systematics (see Fig. 8) it is evident that for practically all the odd isotopes of kurchatovium, the prohibition of oddness leads to an inhibition of spontaneous fission by 103-104-fold. Since a substantial change in the half-life with respect to spontaneous fission (more than 10-fold) is observed in the sequence from the isotope 254102 to the isotope 258Ku, it is natural to attempt to give an at least qualitative explanation for the data obtained. According to the modern concepts, in the region of transfermium elements the barrier to fission has a complex structure, and the stability of the nuclei with respect to spontaneous fission is determined chiefly by the contribution of the shell correction to the total energy of deformation of the nucleus. Let us consider from this standpoint the isotopes of fermium, for which a strong influence on the half-life TSF,/2 by the subshell N = 152 has been established experimentally. When N = 152, the barrier to 1 fission, calculated according to the method of Strutinskii [22], seems to consist of two barriers and takes the form of a two-hump curve with two minima, corresponding to the ground state and isomeric state of the nucleus. The lifetime of this nucleus with respect to spontaneous fission is determined by the integral of motion on the entire path from the ground state to a point lying beyond the second barrier (Fig. 9). With decreasing number of neutrons (transition to lighter isotopes of fermium), the value of the second barrier will be lowered as a result of a decrease in the liquid drop energy of deformation of the nucleus, which should lead to a decrease in the periods of spontaneous fission. In the limiting case, when the ground state is higher than the second maximum, the value of T772 will be determined by the perme- ability only of the first barrier. It can be assumed that as a result of this circumstance the transition from 252Fm to 244Fm will lead to a change of approximately 1012-fold in TSF. 1/2 And yet, as was shown in the work of Randrup et al. [23], the possibility remains that such a situa- tion might arise in the movement from N = 152 toward heavier isotopes of fermium. In this case the super- position of the shell correction and liquid drop energy of deformation may also lead to a substantial de- crease in the second maximum and, as a result, sharply reduce the lifetime of the heavy isotopes with respect to spontaneous fission. Actually, in the sequence from 252Fm to 258Fm, the value of TS' changes 1/2 approximately 1013-fold. Now let us fix N = 152, and we shall measure the number of protons in the nucleus. With increasing Z, the liquid drop portion of the energy of deformation will decrease, which leads to the same consequences that were observed in the case of neutron-deficient isotopes of fermium discussed above. It can be as- sumed that the greatest change in TSF, will occur at the moment when the lifetime of the nucleus is de- 1/2 termined chiefly by the permeability of the first barrier. The possibility remains that this situation exists in the transition from 252Fm to 258Ku, since the half-life with respect to spontaneous fission changes ap- proximately 1012-fold in this case. If this assumption is correct, then further movement in the direction of large Z or variation of N at set Z should not lead to such strong changes in TSF since the value of the first barrier, as it follows 1/2 from the calculations of [22], is comparatively insensitive to the nucleon composition of the nucleus. It seems to us that this viewpoint is qualitatively confirmed by the latest calculations of Pauli and Leder- gerber [24] and Moller and Nix [25,26]; however, the quantitative conclusions require a detailed theoretical analysis, considering new data with respect to the properties of isotopes of kurchatovium and heavier elements. Such an interpretation of the experimental data means that for even?even isotopes of kurchatovium the lifetime with respect to spontaneous fission is already determined practically entirely by shell effects. This is analogous to what is expected for the region of ultraheavy elements, and it must be hoped that in 499 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 the case of movement in the direction of the twice-magnetic nucleus with Z = 114, N = 182, a sharp in- crease in TSF,/2 will actually be observed, as it follows from the theoretical predictions. 1 CONCLUSIONS A number of conclusions can be drawn from the aggregate of experimental results. The method of synthesis of transfermium elements in the irradiation of lead isotopes with ions with mass AI 40 atomic units, investigated earlier [12] for the example of the reaction Pb("Ar, xn)Fm, is also extremely effective using "Ti ions. The reactions Pb("Ar, 2n)Fm and Pb(50Ti, 2n)Ku have approxi- mately equal cross sections, which might have been expected on the basis of theoretical estimates. This permits us to hope that this method can be used successfully for the synthesis of heavier elements in the reactions induced by 54Cr, 55Mn, and "Fe ions. In all probability, the substantial changes in the systematics of the half-lives for even?even isotopes of kurchatovium are associated with the structure of the barriers to fission of these nuclei. From this standpoint, it seems important to investigate the properties of more neutron-deficient isotopes of kurchato- vium, using, in particular, the reactions with2?4Pb, and to attempt to advance into the region of elements with Z 106. On the other hand, a detailed theoretical analysis must be made of the data obtained on the basis of the modern theory of nuclear fission, which, in our opinion, will aid in a more reliable prediction of the properties of heavy and ultraheavy elements. The authors are deeply grateful to Academician G. N. Flerov for his great support, constant aid, and valuable suggestions at all stages of this work; we should like to thank V. M. Plotko and N. A. Danilova for their great contribution to the development of the experimental method and their active participation in the experiments, as well as K. I. Merkin and T. I. Rybakov for their painstaking work on the examination of a large number of detectors of fission fragments. The authors are grateful to the operating group of the U-300 accelerator, supervised by A. N. Filipson, for producing intense and stable beams of accelerated ions. The enriched isotope "Ti was kindly provided by the State Isotope Reserve of the USSR, and the authors are grateful to V. P. Bochin, V. S. Romanov, and S. A. Sel'yanov for aid in the preparation of samples of this isotope for the ion source. LITERATURE CITED 1. G. N. Flerov and I. Zvara, Report of the United Institute of Nuclear Research D7-6013 [in Russian], Dubna (1971). 2. G. N. Flerov et al., At. inerg., 17, No. 4, 310(1964); Phys.Letters, 13, 73 (1964). 3. I. Zvara et al., Report of the United Institute of Nuclear Research P7-3783 [in Russian], Dubna (1968); Radiokhimiya, 11, 163 (1969). 4. I. Zvara et al., Report of the United Institute of Nuclear Research D7-4542 [in Russian], Dubna (1969); Radiokhimiya, 12, 565 (1970); J. Inorg. and Nucl. Chem., 32, 1885 (1970). 5. I. Zvara et al, Report of the United Institute of Nuclear Research D12-5845 [in Russian], Dubna (1971); J. Inorg. and Nucl. Chem. Letters, 7, 1109 (1971). 6. A. Ghiorso, in: Proc. R. A. Welch Found. Conf. on Chemical Research, XIII. The Transuranium Elements. The Mendeleev Centennial, Houston, Texas, 17-19 November (1949), p. 107. 7. A. Ghiorso et al., Phys. Rev. Letters, 22, 1317 (1949). 8. A. Ghiorso et al., Phys. Letters, 32B 95 (1970). 9. Yu. Ts. Oganesyan et al., At. Energ., 28, No. 5, 393 (1970). 10. M. J. Nurmia, Nuclear Chemistry Annual Report LBP-666, Berkeley (1971), p. 42. 11. A. Ghiorso and T. Sikkeland, Phys. Today, 20, No. 9, 25 (1967). 12. Yu. Ts. Oganesyan et al., Preprint of the United Institute of Nuclear Research D7-8194 [in Russian], Dubna (1974). 13. W. Myers and W. Swiatecki, in: Proc. Intern. Symp.,"Why and How Should We Investigate Nuclides for the Stability Line," Lysekil, Sweden (1966); Almquist and Wiksell, Stockholm (1968). 14. Yu. Ts. Oganesyan et al., Preprint of the United Institute of Nuclear Research P7-7863 [in Russian], Dubna (1974). 15. A. S. Il'inov, Report of the United Institute of Nuclear Research P7-7108 [in Russian], Dubna (1973). 16. G. N. Flerov, At. Energ., 24, No. 1, 5 (1968); Ann. Phys., 2, 311 (1967). 17. G. N. Flerov et al., Report of the United Institute of Nuclear Research P7-4932 [in Russian], Dubna (1970). 500 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 18. T. Sikkeland, A. Ghiorso, and M. Nurmia, Phys. Rev., 172, 1232 (1968). 19. T. Galin et al., Preprint IPNO -RC-73-03, Orsay (1973). 20. L. Moretto et al., Preprint LBP-1966, Berkeley (1973). 21. A. G. Artyukh et al., Report of the United Institute of Nuclear Research P7-7189 [in Russian], Dubna (1973). 22. M. Brack et al., Rev. Mod. Phys., 44, 320 (1972). 23. T. Randrup et al., Nucl. Phys., A217, 221 (1973). 24. H. Pauli and T. Ledergerber, in: Proc. 3rd. Symp. on Physics and Chemistry of Fission, Rochester, Paper IAEA-SM-174/202 (1973). 25. P. Willer and J. Nix, ibid., Paper IAEA-SM-174/202 (1973). 26. P. Moller and J. Nix, Preprint LA-UR-74-417, Los Alamos (1974). 501 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 REVIEWS PROBLEM OF ENVIRONMENTAL PROTECTION IN THE OPERATION OF NUCLEAR POWER STATIONS* N. G. Gusev UDC 551.510.72 Protection of population and environment in the design and operation of nuclear industrial units is becoming an important contemporary social problem. This review considers the need for the distribution of established maximum doses over radiation sources and the need for estimates of collective and popula- tion doses from all forms of nuclear energy. Requirement for Distribution of Maximum Population Doses Scientists of all countries have concentrated their attention on the problem of protection against ioniz- ing radiation from the very first steps in the peaceful use of atomic energy. As a result, the safest branch of industry for personnel and population ? nuclear power ? was created in a short time. The international Commission on Radiological Protection (ICRP) established maximum permissible doses (MPD) which are considered safe for the occupational worker and the general population. National regulations were based on these recommendations. In planning units for the nuclear industry, however, the application of these regulations is complicated by a number of factors arising out of the extensive in- troduction of nuclear energy into many fields of human activity. If one assumes that by the year 2000 the *Reviews published in this issue are revised texts of papers read at the plenary session of the All-Union Scientific Conference on Protection against Ionizing Radiation from Nuclear Installations (Moscow, 17-19 December, 1974). TABLE 1. Distribution of Maximum Population Dose over Radiation Sources, % of Total Maximum Dose Radiation sourceSum Gaseous and aerosol wastes Liquid and solid wastes .- Water-cooled, water-moderated reactors (VVER) 4 2 6 Single-loop reactors (EtBMK) 5 3 8 Gas-cooled reactors 2 1 3 Thermal nuclear power stations 3 1 4 Fast reactors 2 1 3 Experimental reactors 3 2 5 Other types of reactors 4 2 6 Fuel reprocessing plants 10 8 18 Ores, hydrometallurgy, fuel elements 2 3 5 Radiochemistry, irradiators 1 1 2 Burial grounds for liquid and solid waste 1 3 4 Structural materials ? ? 5 Agricultural fertilizers ? ? 4 Accelerators, electron-beam tubes ? 1 Television sets ? ? 1 Reserve for other sources ? ? 25 Translated from Atomnaya nergiya, Vol. 38, No. 6, pp. 391-397, June, 1975. Original article sub- mitted January 8, 1975. ? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 502 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 100 10 41 2 6 1950 1970 1980 1990 2000 Year Fig. 1. Mean annual whole-body dose (USA); 1) total dose; 2) natural radiation; 3) medical pro- cedures; 4) global fallout; 5) various sources; 6) occupational irradiation; 7) nuclear power stations and other sources of environmental contamination. TABLE 2. Maximum Annual Doses, rem electrical power from nuclear power stations over the entire world will amount to 3000 GW, then from a rough estimate there will be generated in that period from nuclear power alone (in billions of curies) a mixture of fission products amounting to ?40,000 (no holding time) or 600 (after one year's holding time) including the long- lived fission products (no holding time); 85Kr, 3; 89Sr, 400; 995r, 30; 13.1Cs, 35; 1311, 250; 1291, 10-5, etc. A large *For thyroid gland in children, MPD = 1.5 rem/yr. amount of long-lived isotopes (3H, 14C, 54mn, Co,60 mos) and transuranic elements will be formed. There will be an increase in the number of other sources of artificialradiation and an ever increasing number of people will be subject to their effects. Thus, of the two main principles for protection against radiation promul- gated by the ICRP ? maintaining radiation doses at the lowest possible level (allowing for technological and economic considerations) and a reduction in the number of irradiated individuals ? only the first is actually achievable. In this review, however, we raise the question not of further reduction in dose but of the need for distribution of presently established MPD for the population with consideration of the partial contribution from the separate kinds of radiation sources so that the total dose does not exceed the established maxi- mum. Category Critical-organ group Tr in iv A . Staff personnel 5 15 30 75 B . Indiv iduals in the general population 0,5 1,5 3* 7,5 C. General population 0,17 0,5 1 An analysis was made of a large number of reports on actual and predicted discharges, on levels of environmental contamination, on exposure of the general population to various sources of ionizing radia- tions, and on methods and approaches to the evaluation of population doses. Part of this material appears in [1-171. Based on the analysis, an attempt is made in this review to pick out the most important radia- tion sources (Table 1). The list of radiation sources given in Table 1 includes only the portion susceptible to quantitative analysis. Particularly complicated is the solution of the problem of population exposure for medical applica- tion of x rays, mainly in x-ray diagnostics and stomatological procedures. According to [1], the mean value of the genetically significant dose (GSD) from diagnostic procedures amounts to 20 mrad/yr. Since the GSD is a weighted mean value including the relative birth rate and relative exposure for individuals up to 30 years of age, the actual doses to the gonads, and even more to the bone marrow or abdominal region, will be considerably higher. [By definition [1], GSD (NF F DF -1-1011 Wm 131"f ) ,hW3, h h k .1, j k 503 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 TABLE 3. Maximum Doses Recommended for Individuals, mrem /yr Group Isotope Critical-organ group III Global genetically significant isotopes: 3H, 35S, 85Kr, "3Xe, I34Cs, 137Cs 100 Global somatically significant isotopes: I4C, 1291, 1311, 89Sr, 9?Sr, 'Ce, 238-241Pu, 241Am 400 900 Local isotopes' propagated over a radius > 10 km 170 500 1000 The same for a radius < 10 km 500 1500 3000T 'we arbitrarily call isotopes not in Groups K and L local. tFor the thyroid gland in children, MPD = 1500 mrem/yr. where j and k are the type of irradiation and age; F and M denote female and male; N1 is the number of individuals receiving the dose D1 rad/yr to the gonads; W is the expected number of children per person. All calculations are carried out up to 30 years of age.] The individual radiation doses for the population of the United States from various radiation sources [2] are shown in Fig. 1 as estimated for the period 1960-2000. The comparison was made with respect to whole-body doses except for the dose from medical procedures. In the calculation of the latter, the somatic dose per capita resulting from irradiation of the abdominal region (abdominal dose) was used as the basis. It is clear that this dose is about 90% of the total dose from artificial sources of irradiation and is at least 35% of the dose from all sources, including the dose from natural radiation. Being 2-3 times less than the abdominal dose, the GSD exceeds the dose from any other artificial radiation sources. It is necessary to plan MPD distribution with all radiation sources included and to pay particular attention to the acceptance of technical and organizational solutions for the reduction of population exposure during medical procedures. This form of radiation effect is not included in Table 1 but it must be taken into account (just like exposure from contamination resulting from nuclear explosions) when considering the social aspects of the use of nuclear energy. There are also other sources which do not appear in Table 1. Among them are, for example, dis- charges of natural radioactive materials introduced into the atmosphere by thermal and electric power stations operating with mineral fuel [9,17], nuclear explosion for peaceful purposes, nuclear accidents, exposure in aviation and astronautics when transporting nuclear fuel, etc. However, the reserve set aside for them and also for new sources (25%) does not include the dose from medical procedures. There are standards recommended by the ICRP and accepted in MRB-69 (Table 2). The value MPD = 0.17 rem/yr for the general population (Group 1 critical organs) corresponds to a GDS=5 rem in 30 years. Given below is the recommended distribution of this maximum dose in rem/per 30 years: Staff 1.0 Individuals 0.5 Population 2 Reserve. 1.5 The GSD for the population is distributed in the following manner; 0.5 and 1.5 rem respectively from ex- ternal and internal irradiation. There is no distribution of the maximum somatically significant dose (SSD) with respect to the various population categories. Evidently, the ICRP will make recommendations in time with respect to this problem. New difficulties arise in the development of standards with respect to irradiation. An immediate one is the concept of "global" isotopes. This term arose after it became known that because of nuclear explo- sions, global cpntamination of the biosphere (including the entire animal and vegetable worlds, the 504 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 atmosphere, and the hydrosphere) by such artificial isotopes as 3H, 14C, "Kr, 89Sr, 90Sr, and 137Cs ocurred. However, in nuclear power, 1291 is usually still assigned to the class of global isotopes although the pro- pagation of this isotope over considerable distances by atmospheric flow is unlikely because of the high weight ratio (6 kg/Ci). Nevertheless, a number of other isotopes which are present in atmospheric dis- charges or liquid wastes from nuclear units should be included in the class of global isotopes under con- ditions where there is a high density of nuclear-industry sources. Besides radiological properties (parti- cularly high radiotoxicity), one should consider the total number of isotopes produced in an entire technical cycle, the fraction reaching the environment, the physical and physicochemical characteristics of the iso- topes, the capability of actively participating in biological processes (active metabolism), the possibility of accumulation in soil or other media, the creation of an intensified 7-ray field at a given locality, etc. The most important criterion for assigning one or another isotope (or source) to the global category is a relatively large contribution to the collective or population dose. It is also necessary to take into account the well-known requirement that for presently established MPD occupational workers make up no more than 1.7%, and individuals in the population no more than 3.3%, of the entire population. Table 3 gives MPD corresponding to 100% recommended for individuals in a population and also the isotopes which should be considered global. The MPD for Group K isotopes (100 instead of 117 mrem) is taken from the GSD distribution given above where a genetically significant MPD of 3.5 rem in 30 years is assigned to all categories; the MPD for Group L is taken from Table 2 after subtraction of a dose of 100 mrem/yr resulting from genetically significant isotopes; the MPD for Groups M and N is also taken from Table 2 for categories B and C respectively depending on the distance over which the isotopes may be pro- pagated (usually, the greater the distance the greater the contribution to the collective dose). When using Tables 1 or 3 in practice, one applies the generally accepted rule that the MPD to the entire body or to individual groups of organs from a sum of isotopes (or from several types of ionizing radiation) must not exceed the MPD from a single isotope. Among other controversial questions are the uncertainty of including the contribution to the MPD from radioactive wastes of other nuclear power stations or, generally, other nuclear industrial units, the proper selection of isotopes belonging to the global group particularly if the problem involves liquid wastes, the uncertainty of dose prediction under accident situations, the well-known difficulties in predicting the development of nuclear industry and particularly the appearance of new sources of radiation, etc. We note in conclusion that the USAEC, as a supplement to the basic standards, published a circular in 1971 [18] in which the annual MPD from radioactive wastes of newly planned water-cooled reactors was recommended to be approximately 1/100 of the MPD established for the sum of all sources, namely 5 mrem (instead of 500 mrem) for individuals in the population living at the boundary of the protection zone (roughly 0.5 km) and 1 mrem (instead of 170 mrem) as the mean annual dose for the general population. An extensive discussion of this concept of the USAEC is given in [6-8]. Analysis of actual gaseous, aerosol, and liquid wastes of a nuclear power station and also of actual and predicted doses shows that the proposed distribution of maximum doses does not create any economic or technical problems for nuclear power stations since the actual (calculated) doses are considerably lower than those proposed. Need for Evaluation of Population Dose The population dose is a measure of the general irradiation of the whole body, or of a given organ, for a group (population) as a whole. If the number of people receiving a radiation dose in the range from D to D + dD is N(D)dD, then the population dose Dp is defined by [19] D, = DN (D)dD, (1) where the integration is carried out within the limits of the general distribution of dose for the entire population. If one is talking of a portion of the population, the term "collective dose" is used. The population dose can also be defined by a summation of the individual doses Di,i over all groups of irradiated people Ni, i.e., (2) 505 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 where j is the level of irradiation and the letter i indicates that the specified quantity (Di, Di, Gi) refers to an individual. Sometimes one also uses the term "average individual dose" = P, (3) where p = N. is the number of people in the population. At the present time there are various mathematical models for the quantitative evaluation of indivi- dual and population doses resulting from radioactive wastes of nuclear units. This paper briefly discusses the model developed in [20-23]. Calculation of the individual dose Di from gaseous and aerosol discharges is conveniently performed by means of the expression D, = -97[(rem/sec )/(Ci/m3)] sec/ma] Q [Gil. Here, Q is the discharge of the isotope; Ci is the "meteorological dilution factor" which is numerically equal to the ratio between the atmospheric concentration of the isotope and the amount of isotopic discharge per unit time; Tfi is a dosimetric conversion factor corresponding to the dose rate for unit atmospheric concentration of the isotope. It should be noted that the determination of the dosimetric conversion factor is a subject for serious study since it must take into account all paths for isotopic effects on humans (in- halation, food chain, y -ray fields from radioactive clouds or soil, etc.). Its value may vary very widely for a single given isotope. Numerical data for the factor 07 is given in [6,21, 24]. Unfortunately, there is no uniformity of opinion on the choice of a model for the determination of the factor G. Thus in foreign practice, calculations of Gi are based on the Gaussian model of Pasquill and Gifford [25, 26], which is recommended by IAEA [27]. In our country, other methods are used (for example, see the review [28]). Strictly speaking, Eq. (4) for the individual dose Di is valid for the calculation of doses from gaseous and aerosol discharges for any paths of action except those cases where the contami- nated products travel into other regions. A similar formula for the population dose was produced in the form [20, 21] (4) np= 2 r [persons/m2 I Gp[sec/m ]Q [Gi]. Here r) is the average population density for a given population; Gp is the so-called population dilution factor, which is obtained by integration of the individual dilution factor Gi over the entire area. It can be represented in the following form, (5) G p V- 2 -SI' [ _ exp dt' (6) where F(t) is a dimensionless function which takes into account depletion of the plume because of radio- active decay and deposition of isotopes on the soil during the time t; Hef is the effective height of the dis- charge. In the more complicated case, for example, where the products contaminated at the location of a nuclear power station travel into other regions, the contribution to the population dose can be determined from Dp= DeNplj = D (7) where Dc is a normalization factor numerically equal to the integral dose per unit activity entering the body, rem/Ci/person;* Ij is the intake of a given isotope with food into the bodies of a group of Nj people (here j refers to the amount of the individual intake); Im = is the total intake of the isotope into the bodies of all individuals, Ci. In the simplest case (where one neglects radioactive decay, variation of isotopic content in process- ing of foods, etc.), Im can be replaced by Am ? the total content of an isotope in all contaminated intake products, Ci: *The integral dose or dose commitment is given in general form (for any amount of isotope entering the body) by the expression De = D(t)dt, rad, or De = H(t)dt, rem, where b(t) and H(t) are the absorbed- dose rate and the dose-equivalent rate. 506 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 17? (8) One of the purposes of the population dose is the determination of the public risk in using nuclear energy. In this instance, a linear (and not a threshold)dose?effect dependence is used. Then when calculating the population dose from contaminated products by Eq. (7), or with the approximation (8) included, there is no need to know the distribution of individual doses, the number of individuals, or the locality in which they live. It is only important to know what the total amount Im of an isotope is in contaminated source products during the time of their consumption. In order to include the various metabolic paths of action, Eq. (7) can be rewritten in the following form [22] D p D iK,,K,Q, (9) Here, K1 is the fraction of discharged isotopes deposited on fields or pastures; K2 is the fraction of the isotopes from the discharge deposited on the soil which is transferred into the edible portion of plants; K3 is the ratio between the cumulative content of the isotope in the food intake of the population and its cumula- tive content in the edible portion of cultivated plants or forage at the time of harvesting or at the time cattle are out to pasture. The factor K3 takes into account decay and removal of an isotope during trans- portation, processing and storage of the products as well as its "loss" through migration in the bodies of animals. These dimensionless coefficients are defined by the relations K1 = Gpvgv; (10) K2 =11K4.+112, (11) where K,=_C (Ci/kg) Yr kg 61 1 _1 (12) L ci/m2 zt [ m2 is the fraction of singly deposited activity in the soil which is transported into the edible portion of a given type of plant grown during the time of complete removal of the isotope from the vegetative layer of the soil; Vg is the rate of deposition, m/sec. For dairy cattle out to pasture, for farm animals, for products of plant origin, 3T6 [[ dd aa yy ss yy rr 31 - K3 ? Tef Ways] ie. 0.693 ?k; 11 ?exp ( i?Tf. )1 Kr t.r, 1 ex1)(??.T); f 1 Kf = 2 [i ? exp ( )1. P (13) (14) (15) Here, 1, is the fraction of the area occupied by the corresponding agricultural land; v i and 772 are respec- tively the fractions of all deposited contaminants transported into the edible portions of plants through the roots and leaves; w is the agricultural productivity; C is the total content of an isotope in the edible por- tion of a plant per kg weight for a one-time contamination of 1 m2 of soil surface by an activity of 1 Ci; Fk is the average density of dairy cattle in the contaminated area; S is the area of pasture from which a cow eats grass in a day; Tef is the time for half-removal of an isotope from the edible portionof grass; Kr is the fraction of an isotope absorbed from forage by animals of the r-th type which is transferred into products obtained from these animals (for cows, Kr = Kk); Er,/ is the fraction of all products yielded by the r-th type of animal which is processed into primary food products; Tp and Tf are warehouse storage times for plant products and forage; T1 is the time consumed in processing and transport of the primary form of food products from cattle; TB is the time cattle are out to pasture; A is the decay constant for a given isotope. At the present time, the input parameters for this model are being refined and a computer program is being written. Note that in contrast to Eqs. (1) and (2) where it is necessary to know individual exposure doses, knowledge of population density or of the number of individuals in a population is not needed for the calculation of the population dose from contaminated products when using Eqs. (7)-(15); one only needs to know the amount of isotope present in contaminated food at the time of consumption. 507 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Of the other mathematical models for calculating population and individual doses from radioactive wastes, one should single out the model developed, and already used in practice, by US scientists [10]. Based on this model, the HERMES computer program was written in BASIC; this program encompasses all possible paths of action for radioactive materials in the environment and is intended to solve such problems as the potential development of nuclear power in the US up to the year 2000, optimal selection of sites for nuclear units, escape and propagation of radioactive materials and their resultant concentrations in air, water, soil, and bottom sediments, concentrations in food products, and, finally, radiation doses from radioactive materials entering the body with air, water, and food products and from external radiation _ in air, water, or soil. Results obtained by the use of this and other programs have been presented [2,6,16, 17]. Population or collective doses are a required social criterion in the problem of radiation safety for the population and in protection of the environment since, based on their evaluation, one can solve the problem of the fundamental partial distribution of MPD within a population with respect to radiation sources in order to assure consistency of established maximum doses from all types of nuclear energy used in the activities of society; one can predict population exposure for further development of the nuclear industry and on this basis determine risk factors for the public from the contributions to the population dose of individual radiation sources. Since the risk concept is based on the assumption of a linear dose-effect relation, the public risk factor Rp can be defined by [damage/(rem-person)] N [persons/population] [rem] T, [rem] ---=-/T,Dp. (16) Here, -Ai is the total individual risk factor including somatic and genetic effects (according to ICRP data, 10-4 for chronic irradiation). More detailed discussions of risk factors can be found in [2, 3, 6, 11- 14, 16, 17, 28-30]. The questions touched on in this review require extensive discussion and active participation by specialists in various fields. They must be solved in planning for the development of the nuclear industry of the future in order to ensure protection of all segments of the population against ionizing radiation and protection of the environment against contamination by radioactive materials. LITERATURE CITED 1. United Nations Scientific Committee on the Effects of Atomic Radiation, UN, New York (1972). 2. Estimates of Ionizing Radiation Doses in the United States, 1960-2000. US Environmental Protection Agency, Washington (1972). 3. The Potential Radiological Implications of Nuclear Facilities on the Upper Mississippi River Basin in the Year 2000, USAEC (1973). 4. Natural Radiation Exposure in the United States, US Environmental Protection Agency, Washington (1972). 5. Report on Releases of Radioactivity in Effluents and Solid Waste from Nuclear Power Plants for 1972, USAEC, Washington (1973), 6. USAEC, Final Environmental Statement Concerning Proposed Rule Making Action: Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion:"As Low as Practicable for Radioactive Material on Light-Water-Cooled Nuclear Power Reactor Effluents," Vol. 1-3, WASH-1258 (1973). 7. G. Whipple, Third Intern. Congr. of the Intern. Radiation Protection Association, Washington, 9-14 Sept., 1973, Rep. 156. 8. A. Hull, Nucl. News, 11, 53 (1972). 9. P. Pellerin, in: Proc. IAEA Symp. on Environmental Surveillance around Nuclear Installations, Warsaw, 5-9 Nov., 1973, Rep. SM-180/76. 10. J. Soldat et al., in: Proc. IAEA Symp. on Environmental Behavior of Radionuclides Released in the Nuclear Industry, Aix-en-Provence, France, 14-18 May, 1972, Rep. SM-172/82. 11. The Effect on Populations of Exposure to Low Levels of Ionizing Radiation, HEIR Rep., Washington, D. C., November, 1972. 12. C. Comar, Third Intern. Congr. of the Intern. Radiation Protection Association, Washington, 9-14 Sept., 1973, Rep. 1. 13. J. Crow, ibid., Rep. 2. 14. A. Upton, ibid., Rep. 3. 15. A. Hull, ibid., Rep. 158. 508 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 16. Environmental Radiation Dose Commitment: An Application to the Nuclear Power Industry, USA EPA-520/4-73-002, Washington (1974). 17. Reactor Safety Study. An Assessment of Accident Risks in US Commercial Nuclear Power Plants, USAEC, Washington (1974). 18. Licensing of Protection and Utilization Facilities, US-10-CFR-50, Federal Register, 36, No. 111 (1971). 19. Bo. Lindell, Proc. IAEA Symp. on Environmental Surveillance around Nuclear Installations, Warsaw, 5-9 Nov., 1973, Rep. SM-180/77. 20. N. G. Gusev et al., Third Intern. Congr. of the Intern. Radiation Protection Association, Washington, 9-14 Sept., 1973, Rep. 157. 21. N. G. Gusev and V. A. Belyaev, Proc. IAEA Symp. on Environmental Surveillance around Nuclear Installations, Warsaw, 5-9, Nov., 1973, Rep. 82. 22. V. A. Belyaev, in: Experience in Operation of Nuclear Power Stations and Path for Future Develop- ment of Nuclear Power [in Russian], Vol. II, Izd. FEI, Obninsk (1974), p. 355. 23. V. A. Belyaev, in: International Conference on Physical Aspects of Atmospheric Contamination [in Russian], Vilnius, 18-20 June, 1974, p. 176. 24. P. Bryant, Proc. IAEA Symp. on Environmental Surveillance around Nuclear Installations, Warsaw, 5-9 Nov., 1973, Rep. 12. 25. F. Pasquill, Meteorolog. Mag., 90, 33 (1961). 26. F. Gifford, J. Appl. Meteorolog., 6, 644 (1967). 27. Application of Meteorology to Safety at Nuclear Plants, Safety Series, No. 29, IAEA, Vienna (1968). 28. N. E. Artemova, At. tnerg., 36, No. 1, 32 (1974). 29. Yu. I. Moskalev et al., Concept of Biological Risk for the Effect of Ionizing Radiation [in Russian], Atomizdat, Moscow (1973). 30. Radiosensitivity and Spatial Distribution of Dose (ICRP Publication 14) [Russian translation], Atom- izdat, Moscow (1974). 31. Recent Advances in Nuclear Medicine, New York-London (1974). 509 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 CONTEMPORARY TRENDS IN EXPERIMENTAL SHIELDING PHYSICS RESEARCH V. P. Mashkovich and S. G. Tsypin UDC 621.039-78 The contribution of experimental research to the solution of problems in the physics of shielding against radiations has been decreasing in the last few years in comparison with theoretical and computa- tional research. This change can be traced, for example, in [1,2]. Thus 60% of the papers in the 1969 Symposium were devoted to calculational research, while in 1974 the number was 85%. Moreover, the limitations of theoretical methods require experimental confirmation of the computa- tional methods and the set of constants assumed. At the present time a number of problems cannot be solved by computational means because of limitations of the methods. Contemporary experimental research in shielding physics is characterized by: 1) differentiation of the information obtained; 2) more and more attention to complex geometries (multidimensional and hetero- geneous shields); 3) shift of some of the best understood parts of shielding to engineering practice. All such experimental research can be divided into two groups: research on verifying computational methods and sets of interaction constants, which we call reference point research; mock-up experiments for solving problems which at the present time cannot be solved by computational means because of limitations in the computational methods or difficulties of the model representation. With the development of reference point experiments the statement that "calculation gives the necessary information ? experiment the ref- ence point" is becoming more and more valid. The following are the most interesting reference point experiments on which experimentalists should concentrate their attention. 1) Basic experiments (benchmark type experiments). They should be performed under "clean," standard, very elementary conditions. At the present time it is desirable to perform basic ex- periments in multidimensional geometries. 2) Simulated experiments on actual nuclear-industrial installations such as reactors and accelera- tors to estimate the quality of the computational techniques under actual conditions. These mea- surements enable one to correct the methods and to introduce correction factors into the results of the calculations. 3) Full-scale experiments on actual nuclear-industrial installations. These experiments test the computational methods for solving the most complicated problems of making real shields. Full- scale experiments become much more complicated with increasing dimensionality of the geometry. This generally requires idealizing the problem, which in turn requires careful experimental testing of the admissibility of the assumed idealization and a determination of the errors introduced by such an idealization. The main trends in radiation shielding research at the present time are listed below. 1. The Development of Theoretical Methods and Computer Programs for Calculating Distributions of Ionizing Radiations. In recent years special attention has been paid to the development of the Monte- Carlo method and the method of discrete ordinates. The complex of one-dimensional ROZ programs and the multi-group RADUGA program for two-dimensional reactor shield calculations form a good basis for solving many shielding physics problems. The MD and MDA methods are successful modifications of the Translated from Atomnaya gnergiya, Vol. 38, No. 6, pp. 398-400, June, 1975. Original article sub- mitted January 8, 1975. ? 1975 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 510 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Declassified and Approved For Release 2013/09/25: CIA-RDP10-02196R000400050004-2 Monte-Carlo method for calculating radiation fields. By using these methods radiation fields can be suc- cessfully predicted at large distances from sources, even at earth?air and vacuum?air boundaries. It is gratifying thatprograms have appeared which satisfactorily predict distributions of high-energy particles and electrons. Computer programs have been developed for calculating radiation fields in in- homogeneous media such as shields with multisectional channels and voids, and for nonstationary problems. 2. Accumulation of Experimental and Computational Information on the Shaping of Radiation Fields in Shields. Specialized facilities have been built to study shields. A large amount of work has been done in building various experimental arrangements at reactors and charged particle accelerators. Converters for neutrons of various energies are interesting and seem promising for experimental research on reactor problems. Gamma ray distributions have been investigated. A number of studies have been made of the spatial, angular, and energy distributions of gamma radiation for one-dimensional shield geometry. Interesting work has been done on the time dependence of radiation fields from pulsed gamma sources, and on deep penetration problems, mainly for elementary sources. Studies have been made of neutron and capture gamma distributions. Extensive information has been accumulated on the shaping of distributions of neu- trons of various energies, including intermediate energies, and on capture gamma radiation in various media. Gamma and neutron distributions have been studied in air and at earth?air and vacuum?air bound- aries. These studies are distinguished by the systematic nature of the data and the performance of experi- ments at large distances from sources of various energies, including low-energy gamma sources. Research on shielding high-energy particle accelerators has progressed considerably. The dif- ferential and integral characteristics of electrons beyond thick barriers have been studied intensively. The study of albedos of gamma rays and neutrons from isotopic and reactor sources is largely com- plete. The problems of determining the characteristics of quasialbedo radiation have been studied to a lesser degree. The penetration of radiations throug