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SOVIET ATOMIC ENERGY VOL. 54, NO. 2

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CIA-RDP10-02196R000300020002-8
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RIFPUB
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K
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86
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January 4, 2017
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February 6, 2013
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2
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Publication Date: 
February 1, 1983
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Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Ib N uust$-bJ1X Russian Original Vol. 54, No. 2, February, 1983 August, 1983 SOVIET ATOMIC ENERGY ATOMHAR 3HEPn4H (ATOMNAYA ENERGIYA) TRANSLATED FROM RUSSIAN CONSULTANTS BUREAU, NEW YORK Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Soviet Atomic Energy is a translation of Atomnaya Energiya, a SOVIET I publication of the Academy of Sciences,of the USSR. ATOMIC ENERGY Soviet Atomic Energy is abstracted or in- dexed in Chemical Abstracts, Chemical Titles, Pollution Abstracts, Science Re- search' Abstracts, Parts A and B, Safety Science Abstracts Journal, Current Con- tents, Energy Research Abstracts, and 4 Engineering Index. 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 %st issue of the Russian journal. - Editorial Board of Atomnaya tEnergiya: Editor: 0. D. Kazachkovskii Associate Editors:, N. A. Vlasov and N. N. Ponomarev-Stepnoi Secretary: A. 1. A4 temov Mailed in the USA by Publications , Expediting, Inc., 200 Meacham Ave- nue, Elmont,-NY 11003.. POSTMASTER: Send address changes to Soviet Atomic Energy, Plenum Publish- ing Corporation, 233 Spring Street, New York, NY 10013. I. N. Golovin , V. V. Matveev V. I. 1l'ichev' I. D. Morokhov V. F. Kalinin A. A. Naumov -P. L. Kirillov A. S. Nikiforov Yu. I. Koryakin A. S. Shtan'- E. V. Ku- lov B. A. Sidorenko B. N. Laskorin M. F. Troyanov Copyright ? 1983, Planum Publishing Corporation. Soviet Atomic Energy partici- pates in the program of Copyright Clearance Center, Inc. The appearance of--a code line at the bottom of the first page of an article in this journal indicates the copyright owner's consent that copies of the article may be made for personal or internal use. However, this consent is given on the condition that the copier pay the stated per-copy fee through the Copyright Clearance Center, Inc. for all copying not. explicitly permitted by Sections 107 or 108 of the U.S. Copyright Law. It does not extend to other kinds of copying, such as copying for general distribution; for advertising or promotional purposes, for creating new collective works, or for resale, nor to the reprinting of figures, tables, and text excerpts. Consultants Bureau journals appear about six months after the publication of the original Russian issue. For bibliographic accuracy, the English issue published by Consultants Bureau carries the same number and date as the original Russian from which it was 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 journal, 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 (2 volumes per year) Vols. 52 & 53: $440 (domestic); $489 (foreign) Single Issue: $100 Vols. 54 & 55: $500 (domestic); $555 (foreign) Single Article: $7.50 CONSULTANTS BUREAU, NEW YORK AND LONDON 233 Spring Street; - New York, New York 10013 Published monthly. Second-class postage paid at Jamaica, New York 11431. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya August, 1983 Volume 54, Number 2 February, 1983 CONTENTS Engl./Russ. ARTICLES International Tokamak-Reactor "INTOR." Phase I - B. B. Kadomtsev and V. I. Pistunovich ......................... 83 83 Cooling Conditions of the Reactor of the Beloyarsk Nuclear Power Station First Unit during Shutdown of the Station and Depressurization of the Evaporative Circuit - P. A. Gavrilov, V. G. Zakharov, G. A. Zvereva, A. I. Ionov, V. D. Kozyrev, V. I. Mikhan, Yu. I. Mityaev, T. D. Ogina, L. N. Podlazov, N. Z. Rybakov, A. G. Sheikman, and S. V. Shirokov ............... . ................... 103 98 A Test of Correlation Measurements oftheFlow Rate of Sodium in the BN-600 Assembly - L. A. Adamovskii, V. G. Vysotskii, V. V. Golovanov, V. V. Golushko, V. N. Efimov, B. V. Kebadze, E. P. Kozubov, and V. I. Kupnyi .............................................. .... 106 100 Role of Fast Reactor Physical Characteristics in Limiting the Consequences of Hypothetical Accidents - I. A. Kuznetsov, Yu. E. Bagdasarov, and Yu. M. Ashurko .............. 110 103 Measurements of the Cross Section of the 237Np(n, 2n) Reaction at Neutron Energies of 14.8 MeV - E. A. Gromova, S. S. Kovalenko, Yu. A. Nemilov, L. D. Preobrazhenskaya, Yu. A. Selitskii, B. I. Tarler, Yu. N. Trofimov, V. B. Funshtein, and S. V. Khlebnikov.................................. . 116 108, Effects of Previous Heat Treatment and Mechanical Working on the Swelling of OKh16N15M3B Stainless Steel - V. A. Krasnoselov, V. I. Prokhorov, A. N. Kolesnikov, and Z. A. Ostrovskii .................................................. 121 111 Corrosion of Zirconium Alloys in the Superheated Steam of a Power Reactor - B. V. Samsonov, S. V. Seredkin, V. N. Shulimov, V. K. Shamardin, and G. I. Maershina........ .......... 124 114 Effect of Bombardment with Deuterium Ions on the Structure of Polycrystalline Niobium - A. A. Pisarev, A. I. Evstyukhin, Yu. A. Perlovich, and V. G. Tel'kovskii .............. 127 116 Behavior of Titanium Diboride under Irradiation and Post-Radiation Annealing - L. M. Murzin, V. V. Ogorodnikov, and V. D. Kelim.......... 130 118 LETTERS TO THE EDITOR Vacuum Fission Chamber with Compensation of y-Induced?Current - Yu. P. Bakulin, E. K. Malyshev, S. V. Chuklyaev, and 0. I. Shchetinin .............. ................................. 136 123 - Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 CONTENTS (continued) Engl./Russ. Neutron Fields for Research in the IR-100 Reactor - S. A. Barabanov, G. A. Borisov, E. I. Grigor'ev, I. N. Martem'yanov, V. D. Sevast'yanov, G. B. Tarnovskii, and V. P. Yaryna ...................................................... 139. 124. Kinetics of Two Strongly Coupled Pulsed Reactors - A. V. Lukin ......................................................... 141 125 Radiation Damage to Tungsten Single Crystals by an Argon Ion Beam - V. N. Bugrov and S. A. Karamyan ............................ 144 127 Comparative Neutron-Physics Calculations of Fast Reactors - I. D. Iordanov and N. A. Antonov .................................... 147 129 A Universal Neutron Irradiator with a 252Cf Source - M. A. Bak, A. S. Krivokhat-skii, V. A. Nikolaev, B. D. Stsiborskii, and B. M. Shiryaev ................................. 152 132 Effects of Hydrogen-Ion Bombardment on the Structure and Composition of a Nickel-Rich Alloy - M. I. Guseva, A. N. Mansurova, 0. S. Naftulin, Yu. V. Nikol'skii, P. A. Fefelov, and 0. I. Chelnokov .................................... 154 134 Removal of Mono-2-ethylhexylphosphoric Acid from Solid Extraction Agents with Di-2-ethylhexylphosphoric Acid - V. B. Dedov, P. S. Trukhlyaev, B. S. Kalinichenko, and I. K. Shvetsov .................................................... 157 135 Neutron Activity of the Earth and the Cl-Ar Neutrino Experiment - I. R. Barabanov, V. N. Gavrin, G. T. Zatsepin, I. V. Orekhov, and L. P. Prokop'eva .................... 158 136 Ion Synchrotron Complex of V. G. Khlopin Radium Institute - N. A. Perfilov, V. P. Shilov, V. P. Eismont, V. L. Auslender, V. N. Lazarev, and B. L. Faktorovich ................. 161 137 The Russian press date (podpisano k pechati) of this issue was 1/27/1983. 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 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 At the end of 1978 on the initiative of the Soviet government, work began on the project of an international tokamak-reactor, dubbed INTOR. An International Working Group (IWG) con- sisting of representatives of scientists and engineers from four parties - Euratom members, the USSR, the USA, and Japan - was organized at the IAEA in Vienna. The International Council on Thermonuclear Fusion (ICTF)?, which is a consultative organ of the IAEA General Directorate, formulated the following task for the Group: 1. To determine the goals and characteristics of an experimental facility in the world thermonuclear tokamak program which would follow the large JET, JT-60, T-15, and TFTR tokamaks under consideration at present and which can be built on the basis of international coopera- tion. 2. To evaluate the preparedness of industry for the construction of such a facility in the early 1990's. 3. To specify experiments necessary for producing a project for the reactor. 4. To develop a detailed project for the reactor and a detailed program of experiments on it. 5. To build the reactor and ensure its operation on an international basis. It was agreed that the work on INTOR would be carried out in separate phases. The zero phase of the project was successfully implemented in 1979 [1]. The main work was done by ex- perts of the participating parties under the leadership of the Group. The contribution of each party is estimated at 15-20 man-years with the participation of the leading scientists and engineers working in the field of thermonuclear fusion. Each party presented a national contribution on the zero phase of the project to the Group for consideration [2-5]. On the basis of these contributions after discussions at meetings of the Group an international re- port on the zero phase was prepared [1]. The goals set for INTOR characterize this facility as an engineering-technological test tokamak-reactor with a long burn time of the deuterium-tritium fusion reaction. INTOR should represent the stage of engineering demonstration of controlled fusion as a new source of energy. This step is a natural stage after the physical demonstration which is proposed for the facilities now under construction (JET, JT-60, T-15, and TFTR). The main results of the work-of Soviet scientists and engineers in the zero phase were presented in [6]. On the basis of the results of the zero phase the Group decided to continue this activ- ity in the next phase (phase I) in order to develop the conceptual project of the reactor. The ICTF considered the report of the Working Group and recommended that the IAEA continue the work in this phase. Phase I was completed in June, 1981. The INTOR parameters agreed upon by the Working Group in the zero phase served as the initial conditions for designing. The contribution of each party in phase I was presented in a national report on the conceptual INTOR design. The four national designs [7-10] formed the basis for the international con- ceptual design [11] which the Working, Group prepared in June, 1981. The ICTF discussed the Working Group report on phase I and recommended that the work be continued in the next phase (phase IIA) within the organizational framework employed in the previous work. The phase-IIA workshop envisages optimization of the INTOR conceptual design and is to be conducted from July, 1981 to June, 1983. The objectives of INTOR follow from the role that it should play in the development of the world fusion program and by the engineering capabilities which may arise in the next sev- Translated from Atomnaya Energiya, Vol. 54, No. 2, pp. 83-98, February, 1983. Original article submitted July 15, 1982. 0038-531X/83/5402-0083$07.50 ? 1983 Plenum Publishing Corporation 83 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 TABLE 1. Schedule of INTOR Operation O O Cd z Principal investigations Operation with hy- drogen plasma, veri- fication of function- ing of units Operation with D -T plasma Engineering tests Engineering tests at maximum param- eterst 0,16 0,31 0,62 3,6 6,9 13,8 *In test modules set up on the outside of the torus. tRoughly 5 MW?yr?m-2 for 10 yr after the end of the second stage. veral years for the construction of the reactor project. The engineering foundations were assessed and the required additional investigations were determined in the zero phase [1]. The program objectives of INTOR are: INTOR should attain the plasma parameters necessary for the operation of a demonstration reactor; study and develop the reactor technology required for the construction of a demonstra- tion power reactor; serve as a facility for blanket testing, production of tritium and materials, and de- velopment of an engineering technology; demonstrate the reliability of fusion reactor components; study the system for operating a thermonuclear reactor; make it possible to investigate factors that affect the reliability, operating safety, and protection of the environment during operation ofrthe thermonuclear reactor. The technical objectives of INTOR have been formulated so that this program can bejac- complished. The technical objectives of INTOR are: A. Reactor operating regime: 1. Ignition of a fusion reaction in D-T plasma. 2. Burning of D-T reaction controlled for more than 100 sec. 3. Reactor level of heat and particle fluxes ( > 1 MW/m2). 4. Optimal plasma parameters. 5. Load factor> 70%. 6. Utilization factor 25-50%. B. Reactor technology: 1. Superconducting toroidal and poloidal coils. 2. Control of the composition of the plasma (divertor). 3. Monitoring of the fusion power level. 4. Plasma heating and fuel makeup. 5. Heat removal from blanket and production of tritium. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 6. Tritium fuel cycle. 7. Remote servicing. 8. Vacuum engineering. 9. Fusion power cycle. C. Engineering tests: 1. Breeding and extraction of tritium. 2. Advanced blanket concepts. 3. Materials testing. .4. Plasma engineering. 5. Production of electricity [5-10 MW(elec.)]. 6. Neutron fluence "5 MW?yr?m-2 during phase III for testing the reliability and radia- tion resistance. These requirements will be met in several stages of INTOR operation (Table 1). In stage I optimum regimes of operation with D-T plasma will be investigated. Most of the problems of groups A and B will be solved in the first stage of reactor operation. Stage II will be de- voted to engineering tests, the bulk of which will pertain to group C. During operation in stage III a reactor regime is to be attained with a high utilization factor and a high neu- tron fluence for testing materials and the operating reliability of the reactor units. General Remarks. The conceptual design executed in phase I is based on the realisti- cally expected laws of similarity of the behavior of.plasma with fusion parameters and on the expected development of technology in the next several years. This consistent design, which is fairly detailed as to the individual units, gives an idea of emergent problems and possible ways of solving them. The principal parameters of the reactor (Fig. 1) are given: Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 major radius of chamber, m . . . . . . . . . . . . 5.2 chamber volume, m3 - . . . . . . . . . . . . . 320 area of chamber surface, m2 . . . . . . . . 380 Plasma: minor radius of plasma, m . . . . . . . . . . . 1.2 elongation of plasma cross section . . . . . . . . 1.6 aspect ratio of plasma . . . . . . . . . . . . . . 4.4 mean value of beta , % . . . . . . . . . . 5.6 poloidal beta SI . . . . . . . . . . . . . . . . 2.6 average ion temperature, keV . . . . . . . . . . 10 average ion density, 1020 M-3 ... . . . . . . . . 1.4 energy-confinement time, sec . . . . . . . . . . . 1.4 plasma current, MA . . . . . . . . . . . . . . . . 6.4 field on axis, T . . . . . . . . . . . . . . . 5.5 stability safety factor (on separatrix) . . . . . 2.1 fusion power, MW . . . . 620 neutron load on wall, MW m 2 . . . . . . . . . . . 1.3 Operating: burn time (stages I/II and III), sec . . . . . . . 100/200 load factor (stages I/II and III), % . . . . . . . 70/80 No. of pulses during operation . . . . . . . . . . 7.10? maximum utilization factor, % . . . . . . . . 50 Neutral-beam heating system: No. of injectors (operating/reserve) . . . . . 4/1 beam power, MW . . . . . . . . . . . . . . . . 75 beam energy, keV . . . . . . . . . . . . . . . . 175 pulse duration, sec . . . . . . . . . . . . . . . 10 Fuel makeup system . . . . . . . . . . . . . . . . . Pellet injection and gas filling Impurity control: method . . . . . . . . . . . . . . . . . . . . . . Single-zero poloidal divertor collector . . . . . . . . . . . . . . . . . . . . Tungsten, riveted or welded to copper or stainless steel cooling backing power reaching divertor, MW . . . . . . . . . . . 80 First wall: power at first wall (without neutrons), MW . . . 44 at outer contour of torus: material . . . . . . . . . . . . . . . . . . . . SS-316 stainless steel, cooled with water D20 thickness, mm . . . . . . . . . . . . . . . . . 11.7 at inner contour of torus: material . . . . . . . . . . . . . . . . . . . . SS-316 stainless steel, cooled with water H2O thickness, mm . . . . . . . . . . . . . . . . . 13.5 lifetime, yrs . . . . . . . . . . . . . . . . 15 Blanket: material . . . . . . . . . . . . . . . . . . . . H20, SS-316, Li2SiO31 Pb, _ C temperature, ?C . . . . . . . . . . . . . . . . 400-600 _ thickness, m . . . . . . . . . . . . . . . . . . 0.5 location . . . . . . . . . . . . . . . . . . . . . On outside and top breeding ratio . . . . . . . . . . . . . . . . . . 0.65 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 tritium removal . . . . . . . . . . . . . . . . . Continuous, with helium tritium charged, kg , , , , , , , , , , , , , , , 1 Tritium makeup: tritium circulation rate, g?h-1 . . . . . . . . . 64 annual tritium requirement at 25% utilization factor, kg?yr 1 . . . . . . . . . . . . . . . 7 isotopic enrichment . . . . . . . . . . . . . Cryogenic distillation Tritium charge, kg: blanket . . . . . . . . . . ... . . . . . . 0.5-1 reserves . . . . . . . . . . . . . . . : . . . . 2.3 plasma system . . . . . . . . . . . . . . . . . . 0.2 makeup system, pumps, etc . . . . . . . . . . . . . . 0.4 Shielding, m: internal (blanket + shield) , . , ... , , , , 0.85 external (blanket + shield) . . , , , , . . . . . 1.55 in central injection pipe . . . . , , , . , , , 1.0 around injectors . . . . . . . . . . . . . . . . 0.5-0.75 Toroidal vacuum system: initial base pressure, Pa . . . . . . . . . . . 1.33.10-$ base pressure between pulses, Pa , , , . . , . . 4.10-3 pumps . . . . . . . . . . . . . . . . . . . . . . Mixed cryosorption. pumping out . . . . . . . . . . . . . . . . . Through divertor chamber Toroidal field coils: number . . . . ... . . . . . . . . . . . . . . . . 12 internal size, m . . , , , . . , , , . . 7.75.10.7 conductor Nb3Sn, NbTi stabilizer Cu maximum field, T . . . . . . . . . . . . . . . . 11.6 Poloidal field coils: total flux, W?sec . . . . . , , . , , , . , , , 110 location , , , , , , , , , , , , , , , , On outside of toroidal coils conductor., ,. ., -.NbTi maximum field, T , , , , , , , , , , , , , , , , 8 Windings for gas breakdown: breakdown voltage, V , , . , , , . , . . , 70 location , , , , , , , , , , , , , , , , , , Inside solenoid conductor , , , , , , , , , , ; , , , , , , , , , Cryoresistive, Cu Electrical stored energy: No. of energy storage devices . . . . . . . . . . 6 total stored energy in one generator, GJ 4 peak load on generator, MVA , . , , . . , , 750 Mechanical configuration: The sectors of the blanket are assembled by rectilinear motion in the horizontal direc- tion through windows between the toroidal coils. The semifixed shield inside the torus, on top and at the bottom, forms the primary vacuum boundary at the inner surface. The end flange of the primary vacuum boundary is at the outer boundary of the blanket sec- tor. The test modules are inserted horizontally in the median plane. All of the super- conducting coils are in a common cryostat. Of the 12 sectors, 5 are used for the injec- tionof fast atoms, 2 are for fuel makeup, 3 are for-tests, and 2 are for diagnostics and control. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Switching off Pause ,-n)= 1, 4.1020 m-3 0,05 0,14 5 11 211 217 -_`225 Time, sec Fig. 2. INTOR operating cycle: 1) voltage per circuit of plasma filament; 2) plasma density; 3) plasma current; 4) electron temperature. As one of the technical objectives for INTOR it is proposed to bring the utlization fac- tor up to 50% in the last stage of operation. Analysis of the reliability of the reactor opera- tion, based on an estimate of the reliability of its separate components, shows that reactor reliability requires intensified efforts to ensure reliable operation of its components. When the present operating reliability of the various units is taken into account, the utiliza- tion factor may be 30-40%. Physical Substantiation. The physical foundations of INTOR, described in the zero-phase report (1], were developed and reviewed in phase I from the point of view of consistent solu- tions of the physical and engineering aspects. The main problems are those of plasma confine- ment, the maximum value of 0, the characteristics of the tearing instability, energy trans- port caused by the corrugation of the magnetic field, production and maintenance of plasma equilibrium including the requirements of the system of the poloidal magnetic field, the physics of wall and divertor plasma, the thermal load and the particle flux at the first wall and in the divertor chamber, and problems associated with the operating cycle (raising the current, heating the plasma to ignition of the reaction, control of the burn, suppression of the reaction and turning off the discharge and the pause between pulses). The use of alcator scaling leads to attainment of the ignition regime. A considerable indeterminacy still exists, however, in the extrapolation of the experimental results obtained for INTOR. Undesirable tendencies cannot be ruled out as yet, in particular as 13 is increased. Moreover, the estimates of the possibility of confining energy in plasma do not take suffici- ently accurate account of the energy losses for the radiation of heavy impurities. This means that the poloidal divertor should serve as an effective means of preventing the plasma from accumulating heavy impurities in it. The theoretically predicated limiting value of 0 is roughly 3%. Such a value, however, has already been attained in a tokamak with a circular plasma cross section. If the elonga- tion and triangular shape of the cross section of the plasma do give the expected positive result, then 13 = 5.6%, adopted in the INTOR parameters, will prove possible. These data must. be substantiated experimentally and this will be done in the near future. A description of tearing instability and its characteristics has been elaborated for INTOR. Although additional investigations of tearing have been carried out, there is still not sufficient knowledge of the details of this effect in existing devices, and extrapolation to the INTOR operating regimes adds a large indeterminacy. In better regimes with ohmic heat- ing of the plasma the tearing frequency increases. For INTOR, a tearing frequency of 5.10-3 and 10-3, respectively, has been assumed for the initial and later periods of operation with a characteristic energy-liberation time of 20 msec. It has been shown experimentally that the limiting value of the density in the tokamak grows as the plasma is purified of impurities. The use of additional heating leads to high values of density. The energy losses because of the corrugation of the magnetic field depends weakly on the temperature. An estimate of the loss of particles injected for heating the plasma indicates that the losses remain within allowable limits. Considerable efforts have been undertaken to develop a divertor design for INTOR that would be consistent as far as the physics and engineering are concerned. It has been shown Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 /0 11 VI/i M 1-1 ~ ~ A T- I V u u,~~u u 0 1 2 3 4 5 6 m6 r 4/l `J Fig. 3. Cross section of reactor: 1) plasma; 2) blanket; 3) injector of fast atoms;' 4) torodial field coils; 5) poloidal field coils; 6) cryoresistive part of inductor; 7) superconducting inductor; 8) divertor chamber; 9) removable shield; 10) semifixed shield; 11) cryostat. that the plasma can be confined in the configuration of a poloidal divertor with one zero at SI = 2.6 by means of a system of poloidal field coils located outside the toroidal windings. The modulus of the sum of the currents in the poloidal coils is about 100 MA x No. of turns. Preliminary analysis indicates that this configuration can be established during heating of the plasma from the ohmic regime to ignition. The horizontal position of the plasma can be easily controlled. As for the control of the vertical position and the shape of the plasma filament, it is proposed that a system of passive turns on the first wall be used. New re- sults have been obtained on the modeling of plasma behavior in the region of the wall and in the divertor chamber. New experimental results have been obtained on tokamaks with a diver- tor. It has been established that INTOR will operate at high density (ti5.1019 m 3) and at low temperature (ti100 eV) of plasma in the wall region if it is possible to have a regime with a strong recycling of gas in the space of the divertor. Under these conditions, the pumping rate that ensures removal of helium and other impurities reaches a value of about 2.10' liters-sec-1. The depth of penetration of impurity atoms in the plasma is small, being a few centimeters. The self-sputtering of the divertor plates is also minimal. An open poloidal-divertor geometry, therefore, is applicable and it is expected that a cold protective plasma layer will exist near the first wall if the thickness of the divertor layer is 10 cm or more. The main features of the operating cycle, which were presented in the zero phase, have been confirmed in phase I (Fig. 2). New results on the initiation of current in plasma en- able the initial voltage per circuit of torus to be reduced to 70 V. It is desirable to intro- duce rf power of 5-10 MW in order to facilitate gas breakdown and an increase in the current. A 5-sec ohmic phase of plasma heating seems to be a reasonable duration. Heating the plasma to ignition by 175-keV, 75-MW atomic beams continues to be preferable. Radio frequency methods of heating (ion-cyclotron, lower-hybrid, and electron-cyclotron) can be used if they can be demonstrated to be physically and technically feasible. The burn time has been in- creased to 200 sec but remains substantially shorter than the INTOR total skin time. Con- trolling the reaction burn regime by means of corrugation losses which grow with the tempera- ture proves to be less definite. On the other hand, the latest experiments revealed the ex- istence of a limiting !3 such that losses increase sharply as it is approached. It cannot be ruled out that it is necessary to have a control system incorporating a combination of dif- Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 n ~ Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Fig. 4. Component parts of an INTOR sector: 1) torus sup- ports; 2, 3) semifixed torus sector and inner shield; 4) vacuum seal; 5, 6) interchangeable torus sector and outer shield; 7) test modules; 8) divertor module; 9, 10) outer and inner collector plates; 11) blanket; 12) interchange- able outer shield; 13) pump-out piping; 14) inner support column; 15, 16) beam and lower support of. torus; 17) sealed surface of supports in sector. ferent methods. Further work is required to find a satisfactory solution. The duration of the quenching phase has been increased to 15 sec in conformity with new results of calcula- tions. Preliminary analysis of gas pumping from the operating volume during the pause be- tween working pulses revealed that a duration of 20 sec is sufficient. Mechanical Configuration. Servicing. Activation of the machine by thermonuclear neu- trons, the presence of tritium, and the complicated electromagnetic system of the tokamak give rise to potentially serious problems for servicing and preparing the reactor for opera- tion. From the very beginning of work on the design, therefore, consideration of the prob- lems of servicing has been fundamental for the development of the design. A philosophy of the main reactor servicing that satisfies the requirements of the configuration under consid- eration was elaborated during the conceptual design stage. The existence of such a philos- ophy has led to a modular concept of the design and has affected the design of all systems of the tokamak from the point of view of fulfilling the requirement that they be accessible. The main features of the configuration of the INTOR design are considered. below. A cross section of the reactor is shown in Fig. 3. The Accessibility Requirement in the Design of the Magnetic Field Coils. The most significant feature of the design configuration is the possibility of access for servicing the torus system. The 12 toroidal field (TF) coils are of a size such that a toroidal sec- tor, constituting 1/12-th of the torus, can be removed completely simply by displacement in one direction between the outer supports of the TF coils. These 12 sectors are installed in- side a semifixed frame that protects the torus from the top, inside, and underneath. In order to simplify the reactor servicing, all of the poloidal field (PF) coils are in- stalled on the outside of the TF coils, above and below, so as to provide free windows for the assembly of the sectors. All of the PF coils can be superconducting since their assembly does not require mechanical coupling. A small solenoidal resistance coil is placed inside the solenoid for ohmic plasma heating so as to deliver a voltage pulse for gas breakdown. Cryogenic Vacuum Topology. Since the PF coils are on the outside of the TF coils and are superconducting, it prove possible to design a single cryostat linking all the coils. The cryostat incorporates all the individual enclosures for the outer supports of the TF coils as a part of the overall cryostat. It is thus possible to ensure access to the torus with- out disrupting the cryogenic vacuum boundary. Another significant difference of the design is the complete separation of the cold and hot subassemblies, which simplifies the require- ments as to the thermal deformation of the large structures. The Vacuum Topology of the Plasma Chamber. The toroidal system, consisting of the first wall, blanket, shield, and divertor plates, has been divided into two large parts: a semi- fixed shield and interchangeable sectors (Fig. 4). The parts subjected to the strongest Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 4(0,1) +5(0,15) b'(0,0's) 1 2 3(0,27) \ I I'll I i I I I 7001 plasma 12 1 / U107 III 1,64 16 II 0,83 ;& 12 I ~ ,97 S, 2 Axis of chamber 8,25 Fig. 5. Radial dimensions of INTOR: 1) starting winding; 2) solenoid; 3) support cylinder; 4) jacket of toroidal coil; 5, 13) sylphon bellows; 6) inner clearance; 7, 9) clearances; 8) blanket; 10) outer shield; 11) outer clearance; 12) toroidal coil; 14, 15) outer and inner divertor layers (all the dimen- sions in parentheses and without parentheses are given in meters). effect of the stream of particles and heat (the first wall and the region of the blanket) have been combined into a sector which could be moved separately away from the toroidal shield. It is most important that the vacuum seal for this sector be entirely on the outer part of the torus. The welded seal, on a rectangular flange, between the outer supports of the TF coils is readily accessible for servicing. The other part of the torus, comprising the struc- tural framework and the semifixed shield, forms the primary vacuum boundary and is not moved away during planned normal servicing procedures. Single-Zero Poloidal Divertor. The divertor collector devices are the most subject to damage and, therefore, the design provides for their frequent repair. The modular divertor sector has been designed so that it could be moved away just as the main toroidal sector is moved away. A two-zero divertor offers only limited access to the upper part of the torus. For this reason the concept of a single-zero divertor has been chosen, regardless of the prob- lems involved in producing an asymmetric poloidal field configuration and caused by the large particle and heat loads on the divertor plates. In a divertor system with a single zero the divertor chamber is located in the lower part of the toroidal chamber. This has proved possible as a result of the displacement of the plasma upward by 0.6 m relative to the horizontal axis of the TF coils. In this case centered assembly of the toroidal and divertor chambers is possible inside the system of toroidal coils and servicing is facilitated. The configuration of the divertor region con- sists of 12 divertor modules (Fig. 4), arranged inside the removeable toroidal sector. Support Design. Great efforts went into the design of the support configuration. The TF coils along with the PF coils generate very strong pulsed magnetic forces which act on the TF system. Thermal calculations show that the support structures should operate at cryogenic temperature. The structural form of the support structures, consisting of mechanically con- nected reinforced parts, was designed with allowance for the condition of maintaining the space necessary for removing-the toroidal sector. The design was checked by means of three-. dimensional calculations using a finite-element model. The model took account. of the pre- sence of all 12 TF coils and their support structure. In order to study the details of the designa local stress analysis was made with allowance for the local planar bending. Another feature of the structural configuration is the support system bearing the mass of the reac- tor. The supports of the magnet system are located on the outer side of the reactor so as to ensure access to the lower part of the reactor (see Fig. 3). Radial Dimensions of the Tokamak. The spatial arrangement of all the reactor components and the required clearances for assembly of the separate units are shown in Fig. 5. Assembly and Servicing. The INTOR reactor was designed with allowance for the require- ments of assembly and remote servicing. An important factor in establishing these require- ments is the frequency of equipment breakdown. The assumed numbers of repairs per year for 1,9 >4 1,2 12 0,50 $ 05 2,05 15(0,3) I 94(0,1) 11OS 13(0 2) Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 TABLE 2. Cooling System Parameters Injector + rf system Cooling system 42.5 (helium) 53 power, kW 1100 (nitrogen) 1100 Liquid-helium flow 1800 2000 rate, liters-h-1 some of the main components of the tokamak are given below: Divertor, ion sources of injector . . . . . . . . . . 2-10 (minor repair) First wall, blanket (removable toroidal sector) . . . 1 (medium repair) Magnets, shield (semifixed toroidal sector) . . . . . 0.1 (major repair) The requirements associated with servicing of the equipment have been studied from the point of view of general transporting as well as in order to determine the role of manipula- tors and special equipment and inspection of the systems. - The magnetic and electrical systems are designed for the production of the necessary magnetic field configuration by the single-zero poloidal divertor and for providing the power supply to the main reactor systems. System of TF Coils. The TF coils have been designed so as to permit use of one of the three main superconducting-cooling concepts under study in the world. The superconductor con- cepts considered were: hybrid NbTi/Nb3Sn, cooled by immersion at 4.2?K; NbTiTa, cooled by superfluid helium at 1.8?K by the immersion method; and Nb3Sn with forced cooling with liquid helium as an alternative. The first variant is based on cooling technology that has already been tried and tested in functioning devices. A drawback is the use of Nb3Sn, which is a relatively brittle material and is difficult to fabricate. The second variant does not have this drawback. At a lowered cryogenic temperature it is possible to use NbTi, but the techno- logy of producing superfluid helium has not been sufficiently tested. The third variant has substantial technical advantages over the other two: better heat removal from the conductor and a simpler method of insulation, but this variant is based on a technology which has not had sufficient practical testing. The choice from these variants will be made on the basis of experimental results in the development of the appropriate programs. Losses due to Fou- cault (or eddy) currents have been taken into account in designing the cryogenic'system. The system of PF coils was developed in connection with the requirement that the coils be positioned outside the TF coils. The PF coils are divided into three main groups: a superconducting solenoid that is placed inside the supporting cylinder and performs the function of maintaining the current in the plasma and effecting the ohmic heating of the plasma; a cryoresistive solenoid that is positioned coaxially inside the superconducting sole- noid and serves to cause breakdown of gas in the tokamak as well as initiation and initial growth of the current in the plasma; eight annular superconducting coils that are positioned above and below the TF coils and ensure that the plasma is maintained in equilibrium. The system of PF coils is asymmetric relative to the median plane of the toroidal coils, as a consequence of the use of a single-zero poloidal divertor. The cross section of the an- nular superconducting coils is 1.5 x 1.5 in. The.total current at the end of the burn pulse is 38 MA x No. of turns in the superconducting solenoid, 41 MA x No. of turns in the cryore- sistive solenoid, and 60 MA x No. of turns in the eight annular superconducting coils. The fields and forces acting on the winding were calculated, and the distribution of the forces on the support structure was determined. Two concepts of superconducting coils have been developed. One concept is based on the forced method of cooling the conductor, while the other is based on immersion cooling with liquid helium at 4.2?K. The choice of variant will be made after results of experiments on the appropriate test stands are obtained. The ac power supply system provides the power necessary for the operation of INTOR and has a sufficient reserve for further development. The high-power line ensures a supply for the following reactor systems (in MW): Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 iii Fig. 6. Configuration of the first wall: a) poloidal section of the reactor; b) cross section of the first wall; 1) area of contact with plasma during tearing; 2) region of entry of neutral beam; 3) region of diaphram; 4) surface of emergence of particles due to corrugation of magnetic field; 5) first wall; 6) site of roller joint; 7) corrugated surface; 8) coolant. Motor-generators . . . . . . . . . 120 All cooling pumps . . . . . . . 20 Cryogenic system . . . . . . . . . . . 41 Vacuum system . . . . . . . . . . 5 Auxiliary heating (in foundation) 20 Tritium system . . . . . . ... . . . . 5 Burn control system . . . . . . . . . 15 Auxiliary equipment . . . . . . . . . 15 Total 241 An independent low-power line, at a lower voltage, supplies the respective systems even when the high-power line fails. Provision has been made for diesel generators to provide energy for the critical loads in cases when both lines fail. The system for storing electrical energy reduces the peak power required in the trans- mission line. It uses power from the line during.the burn phase of the operating cycle and from the inductive load of the energy converter of the poloidal magnetic field during the discharge phase and stores this energy for the pulsed loads of the next cycle. It is pro- posed to use the surge generator for large pulsed loads isolated from the power lines. Five surge generators with a stored energy of 4.5 GJ each are necessary for supplying the poloidal field system and the fast-atom injector. Each generator has a peak power of 800 MVA and the duration of the starting pulse is 5 sec. The cryostat of the magnetic system is a vacuum volume in which all of the cooled struc- tures are enclosed. Each coil has its own helium volume. Helium vapor cools. baffles placed inside the vacuum volume to reduce the thermal fluxes from the surfaces which are at liquid- helium temperature. It is proposed, if necessary, to use compounds with a high electrical Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 resistance so as to reduce the induced toroidal currents. The cryogenic system of INTOR is based on the forced-cooling variant in which the TF and PF coils are cooled with liquid helium and includes two cooling circuits: a liquid-helium system, ensuring a temperature of 4-4.5?K for maintaining the superconducting coils in the working state and a liquid-nitrogen system at 80-95?K for cooling the cryoresistive solenoid and the baffles separating surfaces at 4 and 300?K. The liquid-helium system also serves to cool cryogenic panels in the neutral-beam injectors. The cooled units have the following parameters. The mass of the structures cooled to liquid-helium temperature is 9000 tons; the mass of the structures cooled to liquid-nitrogen temperature is 150 tons; the area of the surface carrying off heat by radiation is 3000 m2; and the length of the iron supporting structures is 1 m and their cross-sectional area is 2 2 m . The helium system should have liquefiers and refrigerators. The requirements placed on the cooling system are given in Table 2. The heating and fuel makeup system is designed to heat the plasma to the ignition tem- perature of the D-T reaction and to provide a makeup supply of D-T fuel for it. The Neutral-Beam Injector. Four injectors provide an injected power of 75 MW. An auxil- iary injector is backup. Each injector contains six independent ion sources, the injection angle being from 16 to 22?. The width and height of the cross section that is formed by the intersection of each injector with the tokamak chamber were 1.0 and 1.2 m. The injectors are connected to the tokamak chamber by means of an absolute gate. A shield 1 m thick around the drift tube of the injector increases the area of the intersection of the injector with the tokamak to 9.6 m2 for each injector. The width, height, and depth of the main injector cas- ing are 6, 8, and 4 m, respectively. The mass of each injector is about 100 tons, while the mass of the shield surrounding the injector is about 350 tons. The shield is made so that the injector could be assembled and disassembled. The power supplies and auxiliary equip- ment (cryogenic refrigerators, control systems, cooling systems, etc.) are positioned outside the reactor hall. The gas efficiency of the ion sources is assumed to be 50%. Roughly 104 Pa'liter?sec-1 of cold gas is pumped from the six ion sources by differential pumping of the chamber, carried out with cryogenic pumps with an area of 30 m2 at a pressure of 1.33.10-2 Pa and 64 m2 at 5.10-4 Pa. These pumps, of the cryogenic-condensation type, are covered with liquid-nitrogen cooled stripes. The atom guide ends with an absolute vacuum gate which is opened only during injection, thus reducing the flow of tritium and helium from the tokamak to the injector. An iron magnetic shield 15 cm thick surrounds the gas target chamber. This reduces the scattered magnetic fields of the tokamak to u8.10-4 T in the region of the ion source and the gas target. The influence of neutron irradiation has been reduced by the pre- sence of the shield and an appropriate choice of materials. The insulators in the source, in the system of the direct converter, and in the magnet are either covered or made of inorganic materials. The fuel makeup system includes a system for admitting gas and injecting pellets, and ensures controlled admission of gas into the plasma of INTOR. Gas can be admitted by using all 12 special pipes, on which devices for admitting the gas have been installed, or by using some of them. The backup devices enable the system to be made reliable and flexible for a controlled supply of gas into the plasma chamber after appropriate purification. Two pellet injectors, which can inject deuterium and tritium pellets simultaneously, are installed on opposite sides of the torus. In principle, one injector can ensure satisfactory operation of the tokamak, although a more uniform density distribution in the toroidal direction can be attained with two injectors. The design value of the velocity of pellets from the injector into the tokamak plasma is 2.103 m?sec-1. Both the pneumatic and centrifugal methods of ac- celeration of pellets have been developed in fair detail and can be used in pellet injectors in both channels simultaneously. The First Wall. The concept of the construction of a first wall (Fig. 6) that will func- tion for the entire service life of the reactor has been developed in the INTOR design. The structure of the first wall includes an outer surface which is a substantial part of the sur- face of the plasma chamber and receives particle and heat fluxes from the plasma and radia- tion from the divertor; an internal region subject to the action of particles and radiation while the thermonuclear reaction burns in the plasma and receives the main part of the plasma energy during tearing instability; a diaphragm region on the outer part of the wall, conf in- ing the plasma filament in the initial stage of the discharge; part of the inner surface of the chamber, receiving atoms with the injection energy which pass through the plasma filament Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 TABLE 3. Operating Conditions of First Wall in Steady-State Regime of Reactor* ion Re '2 F ~ E 6 c E ' Ea No, of pulses (with allow - ance for ma= terial fatigue g 0 b . 'd E o N ub ,..,p N c r N 3 oa~i Outer wall 11,7 8 , 7 'f 200 360 3 . 106 107 Corrugated i 1 , 7 8, 7 - 297 400 1. 101% 107 device Diaphragm 12,8 0,81' 280 41o 8-1015 107 Inner w'd11 13,5 10,6$ 275 408 9.10' 107 Region ofbeam 13,5 10,5$ 332 495 2.10-' 107 reception *The maximum temperature is assumed to be 350?C; the maximum allowable stress is 650 MPa from the plasma side and 765 MPa from the coolant side. tPhysical sputtering. 1 Physical sputtering + vaporization. in the initial period of injection; and a region of the outer surface on which substantial particle fluxes impinge in the final stage of neutral injection as a consequence of the cor- rugation of the magnetic field. The main variant of the first-wall design uses water-cooled stainless steel panels. The wall thickness in special regions, i.e., of the diaphragm and the inner part, is increased because of the great erosion of these portions. The structural material chosen is type 316 stainless steel, which ensures radiation resistance throughout the operation of the reactor. It is proposed to avoid corrosion-induced low-temperature cracking of the stainless steel by controlling the chemical composition of the water at low pressure. The thin-walled corrugated cooling channels of the panel structure of the first wall were, chosen in order to reduce the flexural stresses and to increase the service life in comparison with tubes. The outer wall is part of the blanket and is capable of neutron multiplication. The design presents calculations of the erosion rate and the required panel thickness for the various regions. The calculated rate of vaporization due to tearing instability is 8.10-` mm per tearing for an energy density of 289 J?cm 2, released in 20 msec. The extent of the erosion of the inner part of the wall (1.8 mm) during 15 years of reactor lifetime is obtained with an indeterminancy factor of 2. It is assumed that a thin molten layer ('l40 pm) is'formed on the inner part of the wall during a, tearing and does not erode in 'L10 msec (Table 3). Near cuts in the inner part of the wall, an erosion of up to 10% of the thickness of the molten layer (''0.14 mm/tearing) is assumed. As an alternative material for the first wall consideration was given to aluminum, which is attractive because of its high thermal conduc- tivity and, therefore, lower thermal stresses. Preference was given to stainless steel in the design, however, mainly because of the large body of data about its properties, its ex- tensive use in the worldwide fusion program, and its great capability for withstanding tearings. An alternative variant of the first-wall concept is that of a radiation-cooled graphite liner, placed on the inner wall, with water-cooled. stainless steel panels as the outer wall. The graphite liner consists of plates with an area of 30 cm2 and a thickness of 4 cm, mounted on rails. In order to replace them a sector of the blanket with the shield must be moved away. The graphite liner functions at \,1500?K and absorbs an additional power of 40 MW which is radiated to the outer and inner walls. Divertor Collector Plates. The system chosen for eliminating impurities in INTOR is that of a single-zero poloidal divertor positioned in the lower part of the plasma chamber (see Fig. 3). Of the total of 80 MW of power arriving at the divertor, 70 MW impinges di- rectly on the divertor plates. This means that the divertor collector plates operate under the most stressed conditions of all the subassemblies of the tokamak and they must be replaced several times during the lifetime of the reactor. s Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 The collector plates are subjected to large thermal stresses with considerable tempera- ture gradients, physical sputtering, and radiation damage in the form of swelling, embrittle- ment, and creep of materials. To reduce the effect of plasma flows, the inner plate is in- clined at 30? and the outer plate at 14.5? to the separatrix. With this positioning of the plates the maximum density of the heat flux is reduced to 2 MW'm2 and the maximum particle flux, to 1.5 _1022 m 2?sec 1. The two concepts of collector plates considered envisage shielding in the form of tung- sten or molybdenum plates,,which have the lowest sputtering coefficients and can be joined to a cooled stainless steel or copper backing. The shielding plates are sputtered under the effect of particles and, therefore, they must be replaced. The main difference between the concepts considered rests in the method of attaching the protective plates. In the first case, soldering is employed, ensuring high thermal conductivity between the plate and the cooled surface. The other concept is based on a mechanical method of attachment, leading to an uneven removal of heat from the protective plate. When soldering is used, the temperature of the shielding plates is low, but their thermal expansion is limited by the existence of the joint. Compression of the soldered joints leads to large stresses and can cause cracking of the plates, seams, or the cooled backing. On the other hand, at the present time it is difficult to ensure the integrity of the soldered joints during the entire operation of INTOR. Moreover, the copper backing is subjected to radiation fracture, which may limit its service life to 1 yr. In the case of mechanical fastening, the shielding plates can rotate freely and expand when the temperature changes during a pulse. The plate thickness in this variant can be in- creased so as to extend their serivce life as far as sputtering is concerned. During an operating pulse the temperature of the shielding plates rises from 1600 to 2300-2500?C, at which 40-50% of the incident energy is reemitted to the divertor and plasma chambers. This reduces the temperature gradient in the shielding plates and the heat flux to the stainless steel backings, which are designed for the entire operating lifetime of the reactor. The main drawback of this variant is that of operation at a high temperature, which may lead to re- crystallization and embrittlement. At high temperature the chemical sputtering by oxygen im- purities may be large and may reach roughly 0.75 of the rate of physical sputtering. Both concepts are feasible but entail considerable uncertainties. Futher investigations are re- quired in order to choose the final variant. Tritium-Breeding Blanket. The concept of the INTOR blanket envisages the possibility of obtaining a considerable quantity of tritium. From the engineering point of view it seems advisable to place the tritium blanket on the outer and upper parts of the INTOR chamber. In designing the blanket the minimum value of the tritium-breeding ratio was assumed to be 0.6. Solid and liquid materials were consid- ered for the breeding of tritium. Solid materials have an advantage from the point of view of engineering simplicity and a relatively low stored chemical energy. The main problem per- taining to the use of solid materials lies in how the tritium is to be extracted. The great- est uncertainty is associated with the possible influence of radiation on the mechanism of tritium liberation. Ceramics are the solid materials that have the greatest potentialities for acceptable liberation of tritium. The main variant of blanket design uses lithium metasilicate Li2SiO3, while Li4SiO4 is an alternative material.. An alternative concept-of a blanket with Li20 and Li17Pb83 has also been developed. A lead neutron multiplier is placed in the blanket in order to reach the required tritium-breeding ratio. Structurally, the first wall is an inseparable part of the blanket. Water at low temperature is used to cool all parts of the blanket. Two blanket concepts have been chosen. The main difference between them is in the choice of mo- derator and in the positioning of the materials breeding tritium relative to the coolant. In the first concept (design 1) graphite is the moderator and the cooling tubes are surrounded by a lithium-containing material. In the second concept (design 2) water (H20) is the modera- tor while lithium-containing materials are placed concentrically on the outside of the cool- ing tubes. In both concepts the lead neutron multiplier is positioned between the first wall and the tritium-breeding region. Low-temperature water H2O is used to cool the tritium- breeding region, while D20 is employed to cool the first wall and the neutron multiplier. Below we give the main parameters of the tritium-breeding blanket: Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Design 1 Design 2 Neutron multiplier: material. . . . . . . . . . . . . . . . Pb Pb thickness, cm . . . . . . . . . . . . . . . 5 5 maximum temperature,?C . . . . . . . . . 290 290 melting point, ?C . . . . . . . . . . . . 327 327 coolant . . . . . . . . . . . . . . . . . . H2O D20 Tritium-breeding region: breeding material . . . . . . . . . . . . . . L12SiO9 L12SiO9 temperature (max/min), ?C . . . . . . . . 600/400 600/400 effective filling density . . . . . . . . . 0.7 0.7 fuel-element diameter, cm . . . . . . . . . . 4-6 6 6Li enrichment, 7 . . . . . . . . . . . . . . 30 30 tritium-extracting gas . . . . . . . . . . . He He structural material . Type-316 stainless steel maximum temperature of structure, ?C . . . 150 150 coolant . . . . . . . . . . . . . . . . . . . H2O H2O coolant temperature (in/out), ?C . . . . . 50/100 50/100 coolant pressure, MPa . . . . . . . . . . . . 0.7 0.7 neturon multiplier . . . . . . . . . . . . . C H2O thickness of tritium-breeding region, cm . . 43 26 effective covering with blanket . . . . . . . 0.6 0.6 tritium-breeding ratio . . . . . . . . . . . 0.65 0.65 In both designs tritium is extracted from ceramic materials by means of a stream of purified helium at low pressure (,u0.01 MPa). Design 2 is distinguished by several advantages: sim- plicity of assembly; a minimal charge of solid tritium-breeding material; and a thinner blan- ket because of the use of the most efficient moderator. The total quantity of tritium in the blanket (200 g) is determined on the basis of data for the unirradiated material. As shown by theoretical estimates, this quantity can be in- creased to roughly 1 kg. The INTOR design, therefore, envisages the use of 0.5-1.0 kg tri- tium in the blanket. Radiation Shielding. The radiation shielding of INTOR has been designed for the purpose of protecting the reactor components from extreme radiation damage and nuclear heating, re- ducing the level of activity induced in the reactor components, and shielding the operating personnel and the population from the radiation. The system of shielding by means of a local shield makes it.possible for personnel to enter the reactor building 24 h after reactor shut- down. This system consists of a toroidal main shield, an auxilliary shield for the open parts of the torus not protected by the main shield, and the walls of the reactor building, which also serve as biological shielding. The main shield is divided into an inner shield, positioned inside the torus, and an outer shield. The inner shield serves primarily to protect the central part of the supercon- ducting magnet system. Its total thickness in the median plane from the plasma side of the' first wall to the inner surface of the region of the toroidal magnetic coil is 1.1 in. Incor- porated here are a region 40 cm thick that consists of stainless steel (90%) and water (10%) as well as a 34-cm region of boronized steel (60%) and water (40%). The real thickness of the shield, including the first wall and a 3-cm gap, is 0.85 m. Below we give the radiation conditions on the inner part of the toroidal magnetic coils under a neutron load of 1.3 MW* m 2 on the first wall and a total neutron fluence of 6'MW?yr?m-2: Maximum neutron fluence in superconductors, neutrons-cm- 2 . . . 4.1017 Maximum resistance of copper stabilizer, Q?cm . . . . . . . . 3.10'8 Maximum No. of displacements per atom in copper stabilizer, displacements?atom 1 . . . . . . . . . . . . . .2.5.10-` Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Maximum dose in insulators, Gy: thermal insulator . . . . . . . . . . . . . . . . . . . . . 2.5.10' electrical insulator . . . . . . . . . . . . . . . . . . . 7.106 Maximum nuclear heat release in the superconductor, W? Cm 9 . . . . . . . . . . . . . . . . . . . . . . . . . . 9.10-5 Total power of nuclear heating, kW: 12 TF coils . . . . . . . . . . . . . . . . . . . . . . . . 7 magnetic vacuum volume . . . . . . . . . . . . . . . . . . 9 The outer shield, 105 cm thick, has been designed to protect the individual components of the reactor and to provide access by the personnel when the reactor has been turned off. This shield and the 50-cm blanket are separated by a 10-cm gap. This makes it possible to reduce the biological dose rate to 2.5 mR/h (1 R = 2.58.10-4 C/kg) on the outside of the main shield for 24 h after reactor shutdown. The main part of the outer shield (90% steel and 10% water), 70 cm thick, is placed beyond the blanket. Then comes a 28-cm-thick region of boronized steel and water and, finally, a 4-cm layer of lead. Steel with a low nickel con- tent is used in order to reduce the induced activity. Since the details of the design of the auxiliary shield have a considerable effect on. the cost and operation of the reactor as well as on the requirements concerning personnel access, detailed calculations of neutron and photon transport during reactor operation and y-ray transport after shutdown have been carried out. The calculations established that 3 kW of nuclear power is accounted for by a fraction of the vacuum pumps of the atom-injection system with the gate opened. In order to reach a biological dose rate of 2.5 mR/h in 24 h after reactor shutdown it is necessary to have an auxiliary 1-m shield for the atom guide connecting the injector to the tokamak; 0.75 m for the surfaces of the injector chamber and 0.5 m for the remaining parts of the injector and the divertor section are also necessary. The Tritium and Vacuum Systems. The INTOR design provides for a complete deuterium- tritium-lithium cycle. The tritium system ensures execution of the following functions with a-minimum tritium content in the system and a minimum effect on the environment and the operating personnel: regeneration of tritium coming from the plasma chamber, storage of this tritium, and feeding into the plasma supply; tritium breeding in the blanket; extraction of the tritium from the blanket and feeding into the plasma supply system; purification of wastes and coolant to re- move tritium; and monitoring of the quantity of tritium in all buildings. The tritium sys- tem should be shielded from y rays and neutrons and should be designed for maximum reliability coupled with the best utilization factor. The tritium circulation rate in the system during continuous operation is 1035 g/day in stage I and 1597 g/day in stages II and III, including the tritium bred and the external makeup. The quantity of tritium circulating in the plasma system is ti200 g and to this we must add the 390 g of tritium present in the vacuum pumps, at the pelletizing sites, and in the blanket regeneration system. Upon assuming the amount of stored tritium to be 2300 g from calculation of 30 days of continuous operation and taking the amount stored in the blanket to be 0.5-1 kg, we find that the total quantity of tritium in the systems of INTOR is 3.4-3.9 kg. The gases expelled from the plasma chamber are subjected to low-temperature rectifica- tion. At cryogenic temperature, the gas impurities, except helium, are deposited on the cooled surface and are separated. The purified fuel is redirected into the plasma supply system. Tritium is removed from the blanket with a purified stream of helium in the form of T20. At first, in the tritium-breeding system all of the gas functions are converted into water on passage through a catalytic reactor. Then at cryogenic temperature the water is separated from the helium stream and is subjected to electrolysis. The liberated tritium is fed back into the plasma supply system. An important problem is that of the penetration of tritium through the first wall and the divertor plates. The implantation of a large number of tritium atoms in the surface of the first wall in the absence of appreciable recombination may substantially increase the amount of tritium penetrating through the first wall into the coolant. Further investigations in this area are necessary. 98 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 In order to limit the total liberation of tritium to a level of 10-20 Ci?day-1 (1 Ci = 3.700.1010 Bq), the coolant should be processed and should have a fairly low tritium concen- tration. The system is being designed so that the tritium content in the air in all the pre- mises would be less than 5 jCi?m 3. If necessary, triple confinement of the tritium in the system is employed. Double confinement is envisaged everywhere. A special ventilation sys- tem protects the surrounding space from tritium fallout. The total quantity of tritium dis- charged into the environment should not exceed 10 g?yr 1. The Vacuum System of the Torus.- The initial pressure -in, the toroidal vacuum chamber should be 1.33.10-5 Pa. It is proposed to use heating and cleansing of the chamber. The vacuum chamber in assembled form is heated to 200-300?C. The maximum heating temperature is limited by the lead zone of the blanket. The inner surface of the vacuum chamber is cleaned of oxygen and carbon by a hydrogen discharge and is in a hot state. The effective pumping rate necessary for ensuring the initial pressure is 5.10? liters sec-1. During the pause the pressure in the vacuum chamber reaches 4.10-9 Pa in 20 sec at an effective pumping rate of 1.5.105 liters-sec-1. During the burning of the reaction the helium concentration in the plasma is kept below 5%. In order to accomplish this, it is necessary to eliminate from the chamber about 1% of the helium atoms that emerge from the plasma. Estimates show that the effective helium pump- ing rate should be roughly 2.105 liters-sec-1. Thus, the requirements on the pumping system during burning of the reaction are decisive. The pumping system of the toroidal chamber consits of 12 evacuating cells, symmetrically arranged about the reactor axis. Each cell is connected to the divertor chamber by means of piping with a 1 x 1.2-m cross section and a length of 9 m. In order to ensure an effective helium pumping rate of 2.105 liters-sec-1, each cell should have a pumping rate of 5.10? liters?sec-1. The vacuum system of the fast-atom injector includes a vacuum chamber which contains a gas neutralizer and sorption pumps, high-vacuum and forepumps, valves, and measuring instru- ments. A source of positive ions is connected to the vacuum chamber of the injector. The vacuum system of the injector should ensure: an initial pre"s`sure of 1.33.10-5 Pa; vacuum conditions in the ion sources, neutralizer, and the injection channel; as well as a minimum tritium flow from the discharge chamber into the injector; a possibility of replacing the ion source without disruption of the vacuum; measuring the pressure; and effecting vacuum blocking. It is proposed that condensation pumps with liquid helium be used as the principal means of high-vacuum pumping. The rate of deuterium pumping by helium cryogenic panels is roughly 8 liters CM 2-sec-1. The injector is provided with a differential pumping system, helium cryo- genic panels installed in chambers separated by diaphragms. This system makes it possible to reduce the area of the cryogenic panels to 45-50 m2. The pumping rate in each injector is (3 - 4)?106 liters-sec'. In order to obtain the initial pressure and to evacuate the vacuum chamber during regeneration of the cryogenic panels use is made of turbomolecular and mechanical fluid-free pumps. Diagnostics, Data Acquisition, and Control. A preliminary study has-been made of means of plasma diagnostics, technological cycles, and control systems. Two qualitatively differ- ent systems have been distinguished: a system of controls and means of plasma disgnostics and a system of acquisition of data about the parameters of the technological and electrical equipment. Plasma diagnostics, in comparison with that existing in tokamaks, should be developed, for the operating conditions in the field of thermonuclear neutrons of D-T plasma. The opera- tion of the technological and electrical equipment in INTOR must be controlled at a higher level than in previous tokamaks. Some of the means of controlling the technogical systems can be taken from other areas: radiation detection - from nuclear reactors; control of the magnetic system - from high-energy physics; etc. Further efforts are necessary, however, to create combined systems for controlling the reactor operation. An automatic control system is envisaged for controlling the reactor complex by means of feedbacks. Underlying the construction of the control system is the principle of adaptive operation of the individual systems and the entire complex as a whole so as to ensure high operating reliability of the system and to attain optimum regimes of the operation of the complex. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 TABLE 4. Schedule for the Design and Construction of INTOR 1980 11981 11982 1983 1198, 11985 11986 11987 11988 1989 I 1990 I 1991 I 1992 Conceptual design Basic design Technicafdesign Research and development Working design I ( I 1 _ Delivery of equipment I I 1 Construction, installation Assembly Engineering tests Note. The decision as to construction will be made in mid-1985. Criteria for the Selection of the Site and Arrangement of Equipment. The following principal criteria must be taken into account in choosing the reactor construction site: availability of a plot of land sufficiently large for the site of the reactor building and the servicing systems, at some distance from any populated region (the population density in the environs should be low); fulfillment of the pertinent seismic requirements; an appropriate electrical supply system for the complex; existence of water resources for the cooling systems; possibility of building a storage facility for radioactive waste, large units for equip- ment, and weakly tritiated water; fulfillment of the safety requirements during accidental discharge of tritium; availability of manpower for the construction and operation of the reactor and appro- priate living quarters and conveniences; possibility of transporting oversize loads during construction of the reactor, trans- ferring tritium during the reactor operation, and transporting personnel daily and dur- ing trips abroad; proximity of well-developed industry for the assembly of the reactor. The conceptual design of the INTOR complex was elaborated in order to determine the main dimensions of the buildings and the requirements placed on them. The boundaries of the radioactive zone and the layout of the equipment have been established. Equipment is arranged about the toroidal chamber as follows: four operating fast-atom injectors and one backup occupy the space of five torus sectors between the toroidal field windings, while two injec- tors for fuel makeup to the plasma are arranged in two sectors. Three sectors are occupied by test modules and two sectors are earmarked for the diagnostics and control systems. Each of the 12 sectors is furnished with a pumping stem and a pump for removing gas from the di- vertor chamber. The design project presents a program of INTOR tests which determines the program of reactor operation with allowance for the fact that it is being built on an international basis. The plasma physics tests are aimed at problems which cannot be solved on other de- vices - a study of the behavior of plasma during a prolonged reaction. The plasma engineer- ing tests consist in investigating methods of plasma heating, including rf heating. Testing the blanket consists of testing the prototype of a blanket module with an area of 1 m2, in- vestigating the process of tritium extraction, and determining the service life of the stressed elements of the blanket. In the materials testing program it is necessary to ob- tain information about the principal and alternative variants of materials, about materials operating under large thermal loads, and about insulators and tritium-breeding materials. The length of such tests is determined by requirements necessary for the functioning of the first wall and the blanket of the demonstration power reactor. In order to satisfy these requirements, it was decided to change the specimens after bombardment with neutrons at a fluence of 4, 10_, 30, 50, and 100% of the neutron fluence 6.6 MW?yr?m 2 adopted in the INTOR Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 design. After tests on the surface properties of the materials, information will be obtained about the operation of the divertor plates, the limiter, and the first wall. It is proposed to use roughly 5000 specimens of different materials with an area of 1 cm2 each. The neutron tests will be carried out in order to measure the tritium-breeding ratio, the nuclear heating, the reaction rate, and the neutron and y-ray spectra, as well as the induced activity. A demonstration of electricity production on a special blanket sector set up during assembly of the reactor is proposed for the end of stage II and the beginning of stage III. At the end of stage III it will be possible to produce tritium and electricity simultaneously on a blanket sector serving as a prototype of the demonstration power reactor. SAFETY AND THE ENVIRONMENTAL IMPACT The aim of the present section in the INTOR design project is to evaluate the possibility of building a reactor so that its operation would not pose any significant risk to the per- sonnel and the environment. Radioactive Sources. The total tritium content in INTOR is 3.4-3.9 kg. Of this, about 2.3 kg is in store and 0.5-1.0 kg is in the blanket. A tritium discharge into the atmosphere during an accident is most likely from the plasma chamber, the vacuum system, and the system of fuel makeup to the plasma, i.e., an amount of 0.6 kg. The induced activity of the structure and equipment surrounding the plasma chamber, the cooling system, and the atmosphere of the reactor building ranges from 3.2.108 to 1.3.10' Ci after extinction of the reaction. A large part of the activity is accounted for by the first wall: (4 - 10)?10' Ci?m 3. The most radioactive nuclides are 55Fe, 56Mn, 51Cr, SeCo, and 54 Mn, which are contained in stainless steel. The cooling water loop contains a small part (24-70 Ci) of the total activity, which is caused by corrosion or sputtering products of the stainless steel. Moreover, the water coolant will.. contain a large amount of 16N (roughly 106 Ci?m 3). Because of its short lifetime, however, its content does not pose any serious problem. Estimates have been made of the 41Ar concentration in the atmosphere of the reactor hall during operation. After shutdown of the reactor this concentration falls off rapidly as a consequence of the short lifetime. Energy Sources and Possible Accidents. The design considers the most important energy sources which in the event of an accident may lead to a tritium discharge into the atmosphere. The greatest source of stored energy is the superconducting magnet toroidal field system ("40 GJ). Various scenarios of breakdowns of the magnet system and possible consequences have been considered. Analysis of the possible consequences when the coolant is completely cut off in the blanket and in the first wall shows that roughly 300 sec are required, with the reaction pro- ceeding continuously, for melting of the first wall to occur. In this time it is possible to take measures to protect the first wall from fracture. A high probability of melting exists for the zone of the lead neutron multiplier in the blanket. The design should, there- fore, provide for measures against melting of the lead zone. Estimates show that the probability of a hydrogen explosion is low. The concentration of hydrogen isotopes in the plasma in the worst case can be one or two orders of magnitude smaller than the value that leads to danger of an explosion (4%). The INTOR.plasma contains roughly 0.23 GJ of thermal energy. During a tearing insta- bility, in 20 msec energy with a density of 289 J?cm2 is released on the first wall, a layer ,x?0:14 mm thick melts, and vl% of the material of this layer vaporizes. By the end of the operating, time of the reactor the thickness of the first wall is 4 mm, i.e., the probability of of fracture of the first wall because of tearings is low. There is some likelihood of the cryogenic system developing a leak and releasing helium into the reactor building. In this case the pressure in the building may increase by roughly 3104 Pa, which is not dangerous for the designed building. Radioactive Discharges and Their Consequences. In normal operation the discharge of tri- tium owing to diffusion and escape through leaking seals ranges from 10 to 20 Ci?day- 1, accord- ing to estimates. It turns out that such a tritium leak does not result in the ground-level dose of 5 mR?yr-1, established for the population, being surpassed at a distance of 800 m or more. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 The consequences of large discharges were evaluated for a 10-g discharge in the form of the oxide. Indeed, the tritium purifying systems will reduce a discharge into the atmosphere from the building by several orders of magnitude. A tritium discharge of 10 g gives a ground- level dose below the maximum value (25 R) for single events at a distance of 400 m or more. Assessment of all possible accidents leading to a discharge of activated products gave a dose of about 1%, or less, of the dose arising during the discharge of tritium under conditions of ordinary operation. As far as deadlines are concerned, the schedule (Table 4) is fairly tight, but is con- sidered to be feasible provided there are not delays associated with the decisions of the governments concerning the construction of INTOR. It is proposed that the indicated research .and development program, providing the basis for the design, will be carried out at the same time. The program of research and development necessary for the.design, construction, and operation of INTOR was considered in the zero phase of the design project [1]. This program was developed in greater detail in phase I and was supplemented with technological problems, whose solution is necessary for the construction of a tritium-breeding blanket. The conceptual design for INTOR can serve as a basis for the technical design. The de- sign has been drawn up with some margin in the parameters so that present uncertainties in plasma physics and technology would not significantly change the-solutions adopted in the design after further research and development in the nearest future. The design has made it possible to reveal problems which must necessarily be solved in order to build an experimental thermonuclear reactor. Work on the design has reached a stage such that the solution of tech- nological and engineering problems depends on the details of the design. It has become necessary to optimize the design in regard to all of the main subassemblies by working it out in greater detail. The work of the International INTOR Group in phase I made it possible to focus the atten- tion of the scientific community on the solution of the most important problems of both exist- ing tokamaks and those being built. The design gains much by being developed on an interna- tional basis. It concentrates all of the latest world advances in the domain of plasma phy- sics and engineering and technological developments for fusion reactors. It constitutes an example of fruitful international cooperation of scientists and engineers which enabled. the zero and first phases to be carried out in a short time (2.5 yrs). In respect of the level of execution the design surpasses all existing national designs for fusion reactors. LITERATURE CITED 1. INTOR Group, "International Tokamak Reactor - Zero Phase," IAEA Report, Vienna (1980); Yad. Sintez, 20, No. 3, 349 (1980). 2. G. Grieger et al., European Contributions to the INTOR Workshop. EUR FUBRU/XII 501/79/ EDV-50, EUR FUBRU/XII 501/79/EDV-60, Brussels (1979). 3. B. B. Kadomtsev et al., USSR Contribution to the INTOR Workshop - 1979, Kurchatov Institute Report, Moscow (1980). 4. S. Mori et al., Japanese Contribution to the INTOR Workshop - 1980, JAERI Report, Tokai- Mura, Japan. (1980) . 5. W. Stacey et al., US Contribution to the INTOR Workshop - 1979, US INTOR Report, Georgia Institute of Technology, Atlanta, Ga. (1979). 6., B. B. Kadomtsev, Problems of Atomic Science and Engineering. Series "Thermonuclear Fusion" [in Russian], No. 1(15), 3 (1980). 7. EC Conceptual Design Contribution to the INTOR Phase-I Workshop, EC Report, Brussels (1981). 8. Japanese Conceptual Design Contribution to the INTOR Phase-I Workshop, JAERI Report, Tokai-Mura (1981). 9. USA Conceptual Design Contribution to the INTOR Phase-I Workshop, Report INTOR/81-1, Georgia Institute of Technology, Atlanta, Ga. (1981). Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 10. USSR Conceptual Design Contribution to the INTOR Phase-I Workshop, Kurchatov Institute Report, Moscow (1981). 11. INTOR Group, "International Tokamak Reactor - Phase One," IAEA Report, Vienna (1982). COOLING CONDITIONS OF THE REACTOR OF THE BELOYARSK NUCLEAR POWER STATION FIRST UNIT DURING SHUTDOWN OF THE STATION AND DEPRESSURIZATION OF THE EVAPORATIVE CIRCUIT P. A. Gavrilov, V. G. Zakharov, G. A. Zvereva, A. I. Ionov, V. D. Kozyrev, V. I. Mikhan, Yu. I. Mityaev, T. D. Ogina, L. N. Podlazov, N. Z. Rybakov, A. G. Sheikman, and S. V. Shirokov An analysis of the dynamics of the processes in the case of emergency situations with shutdown of the station and depressurization of the circulatory circuit allow the conditions of heat removal in the core to be determined in these regimes (reserves up to limiting power, fuel element temperature), and recommendations to be formulated for the assurance of the'safe operation of the AMB-200 reactor, which has two main (MCP) and two emergency (ECP) circula- tory pumps in each of the two circulation loops [1]. The investigation of an emergency situation associated with total shutdown of the facility is an essential part of the work carried out for substantiating the safety of the Beloyarsk nuclear power stations. In the computational investigations of this emergency, the results of which are presented below, experimental data on the switching-off of the MCP in the Belo- yarsk nuclear power station (Fig. 1) and data on a mechanical breakdown of the turbogenerator used for powering the electric motors of the circulator pumps were used. In the case of trig- gering of the scram system, the experimental data on the transit time of the signal to the equipment and to the control and safety system logic were used. Experiments on tripping the MCP were conducted at a power of 35-45% of the nominal. Before the start of each experiment, stabilization of the thermotechnical parameters of the reactor and the unit as a whole was carried out for 15-30 min, and loop oscillographs were. prepared, having recorded on photographic film the variation of the coolant flow rate and other characteristics of the power generating unit (power, reactivity). After stabilization, the physical and thermotechnical parameters of the unit were recorded and then the MCP were switched off [2] simultaneously with the start of closure of the slide valve at the pump out- let. As a result of this, the flow rate in the circuit changed abruptly over 2-3 sec. In order to ensure nuclear safety and to prevent functioning of the scram system, simultaneously with switching off the MCP, two rods of the standby automatic regulator, operated in the man- ual regulation regime were inserted to a specified depth in the core [2]. When studying the variation of the coolant flow rate after switching off the MCP, it was established that in the cooling reactor the flow rate decreases smoothly without pulsations. In this case, by switching off one MCP in each loop, i.e., two per reactor, a stable flow rate is established over 3.5 sec, and when two MCP in each loop are switched off, i.e.,four per reactor, the flow rate attains a zero value over 3.6 sec. When the MCP are switched off at power, the change of the flow rate has a more complex nature. Three seconds after switch- ing off one MCP in each loop, the flow rate reaches a minimum, and then increases smoothly and after 8 sec becomes stable. The minimum flow rate through the evaporative circuit when one MCP is switched off in each loop at power attained 20% of the initial value. Later it rose to 36% of nominal. In the numerical scheme of the facility, one of the two technological evaporative loops of the reactor with a superheat system is considered [1, 2]. The main plant of the circuit Translated from Atomnaya Energiya, Vol. 54, No. 2, pp. 98-100, February, 1983. Origi- nal article submitted March 18, 1982. 0038-531X/83/5402-0103$07.50. ? 1983 Plenum Publishing Corporation 103 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 2 4 6 8 2', sec Fig. 1. Change of flow rate through the evaporative circuit when the MCP are switched off: 1, 2) one MCP in each loop switched off in reactor cooling and with 44% power, respectively; 3, 4) all 4 MCP switched off in reactor cooling with simultaneous switch-on of the 4 ECP, and only switch-off of all 4 MCP, respec- tively. 0 2 It 6 8 90 z', sec Fig. 2. Reserves R = N1im/N up to limiting power of maximally located EC, when the MCP are switched off: 1, 2) one MCP switched off and the turbogenerator maintained at coasting by one MCP in each loop for EC-5 and EC-6, respectively; 3, 4) all 4 MCP switched off, with simultaneous switch-on of 4 ECP for EC-5 and EC-6, respectively. is listed: separator, descending channel, MCP subassembly, distributing collector, and eva- porative channels (EC) and their communications. The superheat system was described as the superheat channel with communications from the separator to the stream takeoff at the turbine. In accordance with the normal electric power supply system for the inherent requirements of the unit, it was assumed that the MCP are fed from busbars,_ supplying voltage from the turbo- generator unit through a transformer. The mathematical description of the facility and the numerical programs on the computer were based on theoretical requisites for choosing the basis of the mathematical models of reactor dynamics, explained in [3].. The calculations performed on the basis of the experiments on the change of coolant flow rate when the MCP is switched off allow the cooling conditions of the EC to be estimated in the station shutdown cycle. In the calculations the duration of the crisis was determined from the time of exceeding the limiting power of the EC. When the MCP is shut off, the coolant circulation continues for a few seconds because of coasting of the MCP, and then it continues with the emergency circulatory pumps, which 2 sec after shutdown develop nominal efficiency. Because of the nonconformity between flow rate and power, no noncritical cooling of the chan- nels is provided during 4-8 sec. The time of stay of the EC of the reactor in the critical regime of heat removal during operation of the ECP amounts to (Fig. 2) 5.5 sec for EC-5 (five-element channels) and 8 sec for EC-6 (six-element channels). The crisis condition of the heat transfer is dependent upon a positive steam reactivity coefficient, because of which the power of individual EC becomes higher than limiting as a result of the coolant flow rate having been reduced at this instant to its minimum value. Because all types of EC enter the crisis cooling cycle, the permissible level of the temperature increase of the fuel elements is used as a safety assessment criterion of this cycle. Data are given below of a calculation of the fuel-element outer-casing temperature Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020002-8 Fig. 3. Cooling of the graphite masonry of the re- actor in the case of breakdown of the distribution water collector (T is the maximum temperature of the masonry). for maximally loaded EC for the worst case during transition after switching off the MCP to operation with the ECP - the temperature before shutdown and the maximum temperature in the crisis, ?C are given: EC-5. . . . . . . . . . . 375 550 EC-6. . . . . . . . . . . 360 520 Taking account of the brevity of the arrival at-crisis, this increase of temperature of the casing can be assumed to not exceed the permissible values in emergency situations [4). An experiment was carried out at the Beloyarsk nuclear power station with an electric power of the turbogenerator of 7 MW, using coasting of the turbogenerator for supplying elec- tric power to the MCP. Before the start of the experiment, the thermotechnical parameters of the power generating unit were stabilized during 20 min, after which the air cut-out of the 110 kV turbogenerator was tripped and the reactor was shut down by the scram system. The shutoff value of the turbine was closed after 20 sec by the normal circuit. As a result of this, the intrinsic requirements of the unit were provided by the coasting of the turbogenera- tor. The nature of the change of flow rate was recorded with a loop oscillograph. Experi- ments on the use of turbogenerator coasting for supplying power for-two MCP of different loops during an emergency with shutdown of the station showed that the flow rate is reduced smoothly to 40% of the nominal, 100 sec after switching off the air cut-out. In this case during the first 20-25 sec the coolant flow rate through the EC was not below nominal. When using tur- bogenerator coasting for supplying power to the MCP motor, all the EC are cooled by the cool- ant during 100 sec, the flow rate of which significantly exceeds the flow rate during opera- tion of the ECP. It was established in the numerical investigations that by using the coasting of the tur- bogeneratorfor feeding one MCP in each loop - two per reactor - the EC in 4-5 sec briefly enter a heat-removal crisis (see Fig. 2). In this condition, as mentioned above, the tempera- ture of the fuel element casings exceeds the permissible values. An accident with breakdown of the distributing collector is the most serious from the point view of ensuring the temperature conditions of the core. In this case the pressure in the collector becomes less than the pressure in the separator, and leads to reversal of the circulation, operation of the check valves in the outlet circuits of the EC, and to their almost instantaneous steaming. The evaporative circuit is rapidly dehydrated (after 12-30 sec). The increase of the steam content causes an increase of the reactivity and power of the reactor. Owing to the special structural features of the normal EC in conditions of total stoppage of the circulation, a simultaneous heating up of the fuel elements and the graphite masonry occurs due to residual heat release. In the numerical investigation of this hazard it was assumed that, the scram system operates by the period or exceeding the power setting, and that the heating up of the fuel elements at the start of development of the hazard takes place adiabatically. After the fuel elements of the normal EC have heated up to a temperature equal to the average temperature of the graphite masonry 630?C (not before 15 sec), a further-change of temperature of the EC takes place quasistationarily together with the graphite masonry of the reactor. Because of the drain of heat to cooling the zone of heating, acceptable cooling conditions are ensured (Fig. 3) as a result of the alternation of the numbers of evaporative and superheat channels of the reactor core. LITERATURE CITED 1. P. I. Aleshchenkov et al., At. Energ., 16, No. 6, 489 (1964). 2. V. P. Andreev et al., At. Energ., 50, No. 6, 381 (1981). 105 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 3. I. Ya. Emel'yanov et al., Control and Safety of Nuclear Power Generating Reactors [in Russian], Atomizdat, Moscow (1975), p. 12. .4. N. A. Dol.lezhal' et al., in: Experience in the Operation of Nuclear Power Stations and the Course for Further Development of Nuclear Power Generation. Report of the Anniver- sary Obninsk Conference [in Russian], Vol. 1, Physicopower Institute (1974), p. 149. L. A. Adamovskii, V. G. Vysotskii, V. V. Golovanov, V. V. Golushko, V. N. Efimov, B. V. Kebadze, E. P. Kozubov, and V. I. Kupnyi The measurement of the sodium flow rates in the heat-transfer loops of a reactor has be- come a serious problem with the development of fast power reactors using sodium coolant. The electromagnetic primary transducers (magnetic transducers) used for this have, along with the known merits (inertia-free, absence of contact with the flow, convenience in combination with secondary devices, and so forth), a number of significant inadequacies associated mainly with the difficulties of their operational calibration and verification [1]. The necessity of providing direct measurements of the flow rate and calibration of re- gular flowmeters at the operating site has become especially evident during the period of start-up adjustments and power-up of the BN-600. The problem has been solved by the correla- tion method using magnetic transducers [2]. The essence of the method reduces to recording the average time for hydrodynamic inhomogeneities of the flow to cover the distance between detectors, for which identical magnetic transducers are used which are mounted on a straight section beyond the local hydraulic drag - a bend in the piping, a gate, a valve, etc. (Fig. 1). Small random fluctuations of the signals from the magnetic transducers which are correlated with the passage of hydrodynamic inhomogeneities enter into the calculation of the mutual correlation function after amplification and shaping; the time to transport the sodium between the first and second cross sections is determined from the shift in the maximum of the cor- relograms on the time delay axis. The basic positions to be calibrated were the bypass line of the.reactor with IRMU-50 flowmeters* (four positions, first circuit) and sections of the steam generators (SG) with IRMU-300 flowmeters mounted at the exits from the evaporators (24 positions, second circuit). The bypass line of the reactor contains a calibrated nozzle which accounts for .95% of the hydraulic drag of the bypass; therefore, one can indirectly estimate the accuracy of the hy- draulic calculation of the first circuit by measuring the flow rate in this line. It is important for SG to measure the total flow rate, which is necessary for estimation and comparison with the calculation of the hydraulic characteristics of the pump and circuit as a whole, and to control the cross-sectional distribution of the flow rate. The parameters, of the measuring sections (see Fig. 1) were selected on the basis of experience from correla- tion measurements on similar BOR-60 piping and sodium test stands [2]. Regular and auxiliary IRMU-300 flowmeters identical to them served as the transducers. The signals were processed by a 100-channel polar correlometer implementing an algorithm for calculating the sign mutual correlation function (T ? (t) ? T ~ sign x (t) sign y (t+ T) dt, 0 where x(t) and y(t) are the variable components of the signal of the first and second magne- tic transducers, respectively; *IRMU-50 - a magnetic unified flow rate indicator for piping 50 mm in diameter. Translated from Atomnaya Energiya, Vol. 54, No. 2, pp. 100-103, February, 1983. Origi- nal article submitted April 26, 1982. 106 0038-531X/83/5402-0106$07.50 ? 1983 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Fig. 1. Schematic diagram of the measurements: 1, 2) magnetic transducers; 3) unit for analog processing of the signal (filter- ing, amplification); 4) shaper of the polar signal; 5) shift re- gister; 6) logic multiplication unit; 7) memory accumulator unit; 8) digital-analog converter; 9) oscillograph; 10) potentiometer for measurement of the signal of the regular flowmeter; 11) polar correlometer; 12) evaporator of the steam generator (D is the inner diameter of the piping, D = 300 mm; a) identical height mark- ings of the levels). I I VI I 9 0,2 0, 4 0, it 0,5 Time, sec Fig. 2. Examples of measured correlograms: 1) bypass line of the reactor, D = 50 mm, L z 6D, G = 17.5 m3/h; 2) section B4 of steam generator 1, D = 300 mm, L 8D, G = 1670 m9/h. ( sign x (1), sign y (t) + 1 for = f` _ 1 for x (t), y (t) > 0, x(t), y(t)> 1, which for a relatively small length of the plate which equalizes the magnetic field creates the prerequisites for the appearance of an end effect and nonlin- earity of the characteristics. The form of the approximation dependence can be indicated on the basis of the assumption that the distortion of the primary magnetic field by induced currents is proportional to the induced emf: E = kG (Bo - aE), where Bo is the induction of the primary field, and k and a are proportionality coefficients. From this we get F, = kB0G/(1 + kaG). It is evident that as G -} the induced emf tends to some limiting value Elim = Bo/a. If the operating characteristic is nonlinear but sufficiently far from saturation, which is Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 TABLE 1. Main Components of the Systematic Measurement Error Parameter xi Error Si, Source of estimate Distance between detector Measurement conditions in electrodes (baseline) L the piping External diameter of the Tolerances on pipes accord- piping De ing to engineering condi- tions Wall thickness A ?1.0 Time shift of the correlation ?0.5 Engineering data of the 100- maximum Tmax Correction coefficient Kc = 1 Temperature correction of the measurements KT channel correlometer Measurement experience on test stands Calculation observed under our conditions, then with kaG 0.8 MeV) at 510?C (---) and 530?C (-). S, x-S2/S1 77` x--/I2/ni where S1, S22 111, and 112 are the initial cross sections and perimeters of the wire samples. A sensor of similar design had been used already in similar investigations [2]. The accuracy of the method is independent of the intensity of the neutron flux and temperature variations during irradiation. However, at high temperature (?700?C) the electrical resistance changes not only in consequence of the formation of the oxide film, but also because of the dissolu- tion of oxygen in the metal. In addition, the relation (1) was obtained for plane geometry of the sample. In the case of experiments at a temperature of ''500?C, the resulting measure- ment error caused by these two factors amounts to \10% and can be estimated by means of moni- tored metallographic analysis of the wires after completion of the experiment. Wire samples of alloys Nos. 1, 2 with a diameter of 0.39 and 0.19 mm were prepared for the investigations; after repeated deformation during rolling, these were annealed in vacuo for 2 h at 600?C. The texture of the wires was not determined. It is relevant to note here that in the majority of the published data the effect of this factor on the corrosion of samples of different configuration is not taken into consideration. The ampul with the sensor was irradiated in a channel of the RBT-6 reactor. At the same time, a similar sensor was tested in an autoclave without irradiation. The autoclave and the ampul were connected by pipelines; the rate of supply of superheated water vapor with a low oxygen content ( 1,4 4,86.10-2 0,122 15,0 3,64111-11 1,119.111-6 1,6 2,99.10-2 0,106 16,1) 1,09.10-'1 5,08.10-11 1,8 3,431()-2 9,57.10-2 17,0 7,74.111-7 2,3:3.111-6 2,0 :1,01;.10-2 8,88. 111-2 18,0 3,57111-7 1,59. (0-r, 2,2 2,7:3.111-2 7,'10.111-2 Neutron fields in the IR-100 have been studied systematically since 1976, and the spec- tral characteristics obtained from these measurements show good reproducibility. We present below the results of these studies based on data taken during these years. 1. The errors in the determination of the neutrons spectrum V1o (E) in the CEC and OEC- 4 (Table 1) are as follows (confidence coefficient P = 0.95) for the various neutron energy ranges: 0.5 eV-1 keV, no more than 10%; 1 keV-0.3 MeV, 20%; 0.3-10 MeV, 7%; 10-18 MeV, 15%. Application of the procedure in [5] to the integral measurements gave s = 0.06 ? 0.015 for the CEC and S = 0.04 ? 0.015 for the OEC-4, which agree with the spectrum c00(E). 2. The epithermal parameter in the form r T> T , where T is the effective temperature of the thermal neutrons and To = 293.4?K, is equal to 0.056 for the CEC and 0.0094 for the OEC-4. The error was estimated as 7% (P = 0.95). 3. The normalized relative neutron 'flux ~Po is 0.447 and 7.17 neutrons/cm2 sec for the CEC and OEC-4, respectively. The error of the determination does not exceed 5%. 4. The scale factor for use with the sulfur detector-monitors is 3.75 x 1025 RS for the CEC and 1.18 x 1026 RS for the OEC-4. The random error in the determination of KM does not exceed 1% (P = 0.95). From studies of the neutron fields in the IR-100 reactor, standard neutron sources OI-R-3 and OI-T-12 were constructed, forming the basis of the system of standardizing neutron mea- surments in this reactor. LITERATURE CITED 1. Yu. M. Bulkin et al., At. Energ., 21, 363 (1966). 2. R. D. Vasil'ev, E. I. Grigor'ev, and V. P. Yaryna, in: Metrology of Neutron Measure- ments at Nuclear Physics Facilities [in Russian], Vol. 2, TSNIIatominform, Moscow (1976), p. 43. 3. E. I. Grigor'ev and V. P. Yaryna, Izmer. Tekh., 9, 61 (1980). 4. E. I. Grigor'ev et al., Techniques, Technology, Economics. Interindustrial Handbook, Ser. T, No. 49, RD-16/110 [in Russian], (1980). 5. V. P.. Yaryna, [2], p. 17. 140 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 KINETICS OF TWO STRONGLY COUPLED PULSED REACTORS A. V. Lukin UDC 621.039.51 A point approximation has been developed for-describing neutron transition processes in coupled reactor systems [1-3]. The kinetics-of two coupled pulsed reactors has been consi- dered by many researchers (4, 5] in a two-point approximation. An attempt to derive applic- ability criteria of the single-point approximation for analyzing the kinetics of such a sys- tem was made for the first time in [3]. The goal of the present work is to derive and estab- lish the. applicability limits of the single=point model of neutron transition processes in .two strongly coupled pulsed reactors. Let us consider transition processes of prompt neutrons. The processes are assumed to occur in two coupled pulsed reactors, the kinetics of each of which can be described in the single-point approximation. The fission intensity ni(t) (i = 1, 2) in the reactor satisfies the following equations [3]: 2 t lioiti (t)=AKi (t) ni (t) + 2 Kii J dt'fij'(t--t') n1(t')+Kigi (1); ni (t=0)=into, i=1, 2 (1) j=1 0 where AKi = Ki(l - Si) - 1; lio and Ki are the lifetime of the prompt neutrons and the effec- tive neutron multiplication coefficient for the i-th isolated reactor; Kij, coupling parameter of the i-th and j-th reactors, the'parameter resulting from the prompt neutrons only; fij(T), functions normalized to unity and describing the intensity of 'the neutron interaction of the i-th and j-th reactors; and si, qi(t), effective fraction of delayed neutrons and the inde- pendent neutron'source in the i-th reactor. According to [3], the straight and conjugated conditionally critical problems can be stated as follows: (AKi-SK)gi-I-LJ K1jgj=0; i=1, 2; j=1 2 (AKi-6K) g.i -f Kjigj =0; i=1, 2, j=1 ``2, 2 Gj gi = 1; j gigs = 1. i=1 i=1 The subcriticality 6K of the coupled system is defined as the maximum-magnitude root of the equation 6K2-(AK1+Ak,) 6K?AK1AK2-K12K21=0, (5) where AKi = AKi + Kii. When the deviation of the system from the state of criticality is small, 6K is defined in the following fashion within the theory of small perturbations [6] at unchanged Kij values: 2 6K= +6K L11K216K1+LoK126Kz gi0 jBto= L K -1K i=1 0 12 ~- L-0 21 The notation is interpreted as follows: Lo = g20/g10; 6Ki = AKi - AKio is the perturba- tion of the subcriticality ?of the i-th reactor; and the subscript "0" denotes the state of criticality of the coupled system (6K = 0). Translated from Atomnaya Energiya, Vol. 54, No. 2, pp. 125-127, February, 1983. Origi- nal article submitted October 22, 1981; revision submitted September 23, 1983. 0038-531X/83/5402=0141$07.50 Q1.983 Plenum Publishing Cerpora'tion 141 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Let us assume that when the integrals on the right side of Eq. (1) are calculated, the form of the functions fij(T) makes it possible to limit the expansion of nj(t') in a Taylor series of the variable t - t' to two terms. In this approximation Eq. (1) assumes the form lini (t) = Ali i (t) n{ (t) +Kijtzi (1) +Ktg{ (1), j t, j = 1, 2; fi =1io+,K{iT,{ -)-Ki7Tifni (t)/n{ (t), j -# i, i=1, 2; iii= dTTfi1 (T), f, i= 1, 2. u We represent the fission intensity -n the reacto s in the form of products ni(t) _ pi(t)n(t) (i = 1, 2), where n(t) = n1(t; + n2(t) is she total fission intensity in the cou- pled system. By subtracting the sum of Eqs. (3) (the first one of which was first multiplied by nl(t), and the second by n2(t)) from the sum of Eqs. (7) (the'first of which was multi- lied by gi(t), and the second by gz(t)) and assuming that the inequality Ipi(t)/pi(t)Ik is found for this spectrum, and a new value of v is determined by formula (6), etc. The fission spectrum for v is found by linear interpolation of the tabulated values corresponding to the values assigned as the starting data for v. The output data of the PRIDAN.program can be formulated in the form of input data for different programs for the neutron-physics calculation of fast reactors: MED, FAST, ANALIT, etc. These programs, like PRIDAN, are written in FORTRAN IV language. A standard fast reactor [7, 8] was investigated, which is a simplified spherical model of a fast power reactor with sodium coolant and oxide fuel. The volume of the core is 2.5 m3. The isotopic composition of the reactor, geometrical dimensions, and number of points of the spatial network of the calculations are given in Table 1, and the fuel composition for versions A, B, and C are given in Table 2. In addition to the BNDL constants [5], the same library was used in the calculations, but with the recalculated data for 239Pu and 238U (BNDL- M) [8]. The effective neutron multiplication coefficients keff and the integrated neutron flux over the zones were calculated by the MED program [9]. The maximum number of energy groups in the program was 26, the maximum number of spatial points was 170, and they were distributed in seven zones. The critical masses for the plutonium isotopes, contribution to the reactivity of 240Pu and 241Pu, normalized integrated spectra, reaction velocities, neutron balance, and plutonium breeding characteristics were calculated by the FAST program (10). It can be seen from the data of Table 3 that the values of keff calculated using the BNDL constants are markedly higher than those obtained by means of the BNDL-M constants. The re- sults of the calculation of the neutron balance, using the BNDL-M constants, are given in Table 4 for P, F, and C - the production, fission, and capture reaction velocities, respectively. In order to estimate the effect of sodium losses on the reactivity, keff was calculated for version A with a reduction of the amount of sodium in the core and shield by a factor of 2. As a result, Ok/k = -0.43%. The calculated critical masses of 299Pu for versions A, B, and C amount to 991.4, 1051.9, and 1002.3 kg, respectively. The parameters defining the plutonium breeding - the excess breeding coefficient Cj for the zone j, the physical breeding coefficient Bj, and breeding factors BF - are given in Table 6. In order to study the effect of the neutron fission spec- trum on keff, the spectrum corresponding to v = 2.9 for the core and v = 2.88 for the shield was replaced by the spectrum corresponding to v = 2.4. For all versions and systems of con- stants BNDL and BNDL-M, keff was reduced by almost 0.4%. In order to estimate the effect of the multigroup constants, version C was also calculated using the UKNDF constants [11] for 240Pu. Although the principal results of the calculations proved to be quite close (critical mass of 239Pu = 947.7 kg, BF = 1.35; according to BNDL: 1002.3 kg, 1.37, respectively), it was was advantageous to correct the BNDL data for 240Pu. The calculations for version B were performed with different fission product capture cross sections (Table 6): with cross sections from BNDL (B-1), with cross sections from RCN [12] system of constants (B-2), and with cross sections obtained from BNDL with the addition of the long-lived fission product cross sections for 113Cd, 149Sm, 1s1Sm, 155Gd, 157Gd (B-3). From a comparison of the versions B-1 and B-3 it can be seen that the capture and fission product cross sections for energies below 160 eV have no significant effect on the neutron parameters of the reactor. This is explained by the fact that there are almost no neutrons in this energy range [3]. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 The results of the calculations of version A showed that keff increases by 2.0 and 2.9%, with increase of the temperature of the medium from 300 to 2100?K, when using the BNDL-M and BNDL systems of constants, respectively. Calculations using the procedure of [13] for the LADY GODIVA fast reactor, performed for two zones in four-group diffusion approximation, showed the excellent agreement between the experimental and calculated data for keff and the critical radius of the reactor. The use of different finite-difference methods [14] for the solution of multigroup dif- fusion equations is accompanied by the appearance of errors, due to the approximation of the analytical differntial operators by finite differences.. With specified conditions, the multi- group diffusion equation can be solved analytically by means of a relatively uncomplicated method [13], which is also suitable for practical application. Based on this method, the ANALIT program has been developed for the calculation of criticality in plane geometry. Cal- culations of the French fast. reactor MASURCA were performed by this procedure. The results of calculations by the ANALIT program are compared in Table 7 with the results obtained with the program developed on the basis of the finite-central-differences program, for which the errors are shown to be minimal. 1. 2. 3. 4. 5. I. D. Iordanov, Yad. Energ., 15, No. 8 (1981). R. Kidman et al., Nucl. Sci. Eng., 48, 189 (1972). I. D. Iordanov and N. A. Antonov, Yad.Energ., 18 (l982).* I. D. Iordanov and N. A. Antonov, Yad. Energ., 18 (1982).* L. P. Abagyan et al., Group Constants for the Calculation of Nuclear Reactors [in Rus- sian], Atomizdat, Moscow (1964). 6. G. Cecchini, M. Cosimi, and M. Salvatores, The Pravda Code for the Generalization of Effective Capture Cross Sections [in Italian], RT/FI (72) 48, CNEN-Rome. 7. A. Baker and A. Hammond, Calculations for a Large Fast Reactor, A Comparision of Results Organized by the IWG on Fast Reactors of the IAEA. TRG Report'2133 (R) (1971). 8. L. P. Abagyan et al., Calculations of the Characteristics of a "Standard" Fast Reactor, Preprint FEI-525, Obninsk (1974). 9. I. D. Iordanov,. Yad. Energ., 6, No. 5 (1978). 10. I. D. Iordanov and N. A. Antonov, Yad. Energ., 18 (1982).* 11. E. Menapace, M. Motta, and G. Panini, A 26-Group Library and Self-Shielding Factors for Fast Reactor Calculations from the UK Nuclear Data File. RT/FI (73)15, CNEN-Rome. 12. H. Gruppelar, RCN-l Pseudo Fission Product Capture Group Cross Section. RCN-205 (1974). 13. I. D. Iordanov and N. A. Antonov, Yad. Energ., 18.(1982).* 14. I. D. Iordanov, "Analytical solution of the multigroup diffusion equation," Doctoral Thesis, Bucharest (1975). *As in the Russian original - Publisher. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 A UNIVERSAL NEUTRON IRRADIATOR WITH A 252Cf SOURCE. M. A. Bak*, A. S. Krivokhat-skii, V. A. Nikolaev, B. D. Stsiborskii,. and B. M. Shiryaev The widespread use of the method of activation analysis using neutrons of various energies requires sources with different neutron energy spectra. A more efficient and convenient method is the use of a single irrad3..t -,chose averagE energy and shape of energy spectrum in the region of irradiation of the spec.. liens can be c anged fairly rapidly and simply. In order to obtain such a source of neutrons, we have developed and constructed a labora- tory irradiator with a 252Cf source. The energy spectrum of the neutrons in the irradiation cavity is shaped by choosing the necessary combination of hydrogenous moderator around the ampoules with californium and the irradiated specimen and regulating the distance between them. Figure 1 shows the construction of the irradiator. It consists of three coaxially ar- ranged right circular cylinders. The inner cylinder and the spaces between the cylinders can be filled with a neutron moderator, and we chose water for this purpose. The moderating properties of water are slightly less favorable than those of materials such as polyethylene, but water has a number of advantages for this type of construction: accessibility, simple introduction into the irradiator cavity and removal from it, and the possibility of shifting the sources in it. Six metal ampoules with 252Cf are placed on special holders in the inner cylinder. The total mass of the californium may be as much as 100 pg. The purpose of dis- tributing it into six ampoules is to produce a more uniform neutron field at the center of the irradiator; in addition, it is more convenient to handle lower-intensity sources when they are being transported and loaded into the irradiator. The placement of the ampoules into the irradiator and their removal from it can be accomplished without dismantling it. The limiting amount of californium used is determined by the conditions of radiation safety of the service personnel; the radiation shielding for them consists of the outer layer of moderator in the irradiator, which weakens the neutron flux, and the layer of lead, which weakens the y radiation. When low-power neutron sources are used, the lead layer between the inner and outer cylinders may be removed, and the space thus left may be filled with modera- tor or left empty. This reduces the mass of the apparatus. The ampoules containing the californium can be brought closer to the center or moved away from it symmetrically within the limits of the inner cylinder. They can be moved into a rigorously fixed calibrated posi- tion by using a mechanical control to rotate the axial tube, which bears a rotation indica- tor. At the center of the inner cylinder, at the lower end of the axial tube, there is a polyethylene chamber for the material being irradiated. The axial tube, together with the chamber, can easily be inserted into the irradiator and removed from it. The dimensions of the central region of the moderator are chosen with due regard for the requirement that the radius of the inner cylinder must be larger than the slowing-down length of the source neu- trons (in our case this is 5-6 cm) and the diffusion length of the thermal neutrons (of the order of 3 cm). Under these conditions, leakage of neutrons from the moderator should not lead to any substantial attenuation of the neutron flux at the center of the irradiator. On the basis of this, the diameter of the inner cylinder was chosen to be 30 cm. The total dia- meter and height of the irradiator may be as much as 1 m. In order to reduce its mass, the shape of the external tank may be made nearly spherical. The irradiator was tested with 235U and 238U targets. The fissions were recorded with solid-state detectors. We measured the fissionability of the 295U and 238U (as the fission- ability characteristic we took the number of fissions taking place in specimens of fission-. Translated from Atomnaya 9nergiya, Vol. 54, No. 2, pp. 132-133, February, 1983. Original article submitted February 8, 1982. 152 0038-531X/83/5402-0152$07.50 ? 1983 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020002-8 6/Y 5 6 7 Fig. 1. Schematic diagram of the construction of the irradiator: 1) lead shielding; 2) chamber for the specimen being irradiated; 3) cover of the inner tank in which the mechanism for moving the sources is placed; 4) central rotatable tube; 5) middle tank; 6) polyethylene packing; 7) outer layer of moderator; 8) inner tank; 9) isotope sources; 10) isotope-source holders. Fig. 2. Cadmium ratio for 235U (- - -) and ratio of the fissionabilities of 235U and 238U ( ) as functions of the distance r between the center of the chamber and the neutron sources. able material with unit mass per unit of time), the cadmium ratio for the 235U (the ratio of the fissionability of the 235U without shielding of the target with a cadmium layer to its fissionability when the cadmium was used), and the fissionability of the 295U and 238U by fast neutrons when the moderator was removed, so that the lead cavity was left empty, and the neutron sources were inserted until they touched the chamber. Figure 2 shows the results of some of the measurements. To estimate the possibilities of using the irradiator in neutron activation analysis, we may use the measured fissionability of the 235U. When neutron sources with a total of 100 pg of 252Cf (neutron flux of 3.108 sec-1) are placed at an optimal dis- tance of r = 2.5 cm from the center of the chamber and all the cavities of the irradiator are filled with water, the fissionability of the 235U is 1.7.103 mg 1?miri 1. The ratio of the fissionability of the 235U and 238U increases with increasing r, from 450 (for r = 17.5 mm) to 6000 (for r = 75 mm) when all the cavities are filled with water. When the water is removed from the inner cavities of the irradiator, this ratio decreases to about 5. When the water was removed from the central and intermediate cylinders, the fissionability of the 238U and the cadmium ratio for it varied according to a law close to i/r2, which indicated that the flux of fast neutrons does not play a significant role after .Lc is scattered by the various structural components. The fissionability of 235U shielded with cadmium was found to be lower under these conditions than in the case when the space was filled with water; this was evidence of the considerably larger contribution of 2 SU Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 fissions caused by supercadmium neutrons when the space was filled with water. If the space between the inner and intermediate cylinders was not filled with water, the flux of slowed- down neutrons for large values of r was about 10% less than with a water filling, while the flux of supercadmium neutrons remained unchanged. The measurements showed that an irradiator constructed on the basis of an isotopic neu- tron source makes it possible to form neutron fields in which the effective neutron energy may vary considerably. Such an irradiator can be used not only as a stationary device, but also as a transportable apparatus for activation analysis using neutrons of various effec- tive energies. The total mass of the irradiator is about 1 ton. EFFECTS OF HYDROGEN-ION BOMBARDMENT ON THE STRUCTURE AND COMPOSITION OF A NICKEL-RICH ALLOY M. I. Guseva, A. N. Manaurova, 0. S. Naftulin, Yu. V. Nikol'skii, P. A. Fefelov, and 0. I. Chelnokov We have examined the effects of hydrogen-ion bombardment on the structure and chemical composition of a nickel-rich alloy (18-20% Cr, 53-56% Ni) containing molybdenum and tungsten. The alloying elements make this alloy considerably superior in properties to an austenitic stainless steel of type Khl6Nl5 and Kh18N10, and in particular it has much higher long-term strength and plasticity. Targets were bomarded at 7509C by scanning a monoenerletic beam of hydrogen ions of energy 10 keV over the surface, the intensity being 6 x 10 S ions?Cm 2?sec-1 at the dose range from 1019 to 5 x 1020 ions ?cm 2, these values corresponding to the working lives of certain :proposed large fusion systems. The experiments were performed with the ILU ion-beam accelerator [1]. Auger spectroscopy was used to examine the effects of hydrogen-ion bombardment on the chemical composition of the surface. Serial etching with argon ions of energy 2 keV was used to determine the depth distributions of the major components. The surface topography was examined in a scanning electron microscope for various doses of these hydrogen ions of energy 10 keV, and it was found that there were pores of size about 100-200 nm, as well as sharp grain boundaries appearing as a result of the ion etching. The pores occurred within the bodies of the grains and at the. rain boundaries. The pore density in the grain bodies increased with the dose and at 5 x 102 ions ?cm 2 was about 109 Cm-2. Fusion of the pores at the grain boundaries produced continuous cracks, and at a dose of1019 ions 'cm 2 the cracks occurred only in individual.-parts, whereas at 5 x 1020 ionscm2 the cracking at the grain boundaries occurred virtually throughout the irradiated surface. It has previously been observed [2] that a porous surface is produced by implantation of high doses of helium ions in nickel-rich alloys, and the same has been observed for high- temperature helium blistering at Tirr > 0.5 - 0.6Tmp [3]. Das and Kaminsky [4] irradiated polycrystalline Nb with D ions (energy 250 keV) to a dose of 1.25 x 1019 ions?cm 2 at 700?C and found that blisters were formed at the surface. In the case of a nickel-rich alloy ex- posed to H+, the effects were as in the bombardment with He+ at high temperatures, in that a spongy structure was produced by diffusion of the implanted ions and defects, which led to diffuseness in the distribution curves, and also to virtually identical stresses throughout the thickness of the implanted layer from the surface. In that case the gas bubbles were also uniformly distributed over the thickness of the layer and linked up at high doses to form channels emerging from the surface, along which the gas was released from the target. One assumes that this extensive branched porous surface will favor reduced tritium loss from the volume. Translated from Atomnaya Energiya, Vol. 54, No. 2, pp. 134-135, February, 1983. Origi- nal article submitted April 9, 1982. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 C Fig. 1. Distributions of the major com- ponents in a nickel-rich alloy after ir- radiation by hydrogen ions of energy 10 keV (Tirr = 750?C) at doses of 1019 (a), 5.1019 (b), and 1020 ions?cm 2 (c); A - Cr; i.i - Fe; ^ - Ni; x - Mo; A - W; +) 0; s- C. Figure 1 shows distribution curves for the major components and the alloying ones after various hydrogen-ion doses at 750?C obtained by Auger spectroscopy with sequential etching. The bombardment increases the chronium concentration in the surface by a factor of 2-3 as the dose increases from 1019 to 1020 ions cm 2, while the thickness of this chromium-enriched layer increases by a factor of about five. At the same time, there is a reduction in the chromium content in the deeper parts of the specimen, and the thickness of this depleted re- gion increases with the dose. This is characteristic of radiation-induced segregation asso- ciated with nonequilibrium impurity vacancies [5, 6]. The complexes between vacancies and impurity atoms move to the surface, where one gets a nonequilibrium but reasonably stable segregation of the impurity atoms after vacancy annihilation [6]. The surface contains a con- siderable amount of oxygen on account of the oxidation of the chromium. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Figure 1 shows that the surface layer lacks.tungsten and molybdenum atoms, and the thick- ness of this depleted layer increases with the dose. The irradiation times at 750?C varied from about 1.5 to 15 h. The lack of W and Mo in the surface layer cannot be explained as due to sputtering by ion bombardment, because Fe and Ni sputter preferentially, i.e., they are components having higher sputtering coefficients [7]. Also, on increasing the dose from 101 to 5 x 1019 or 1020 ions ?cm-2 (Fig. 1), the thickness of the sputtered layer is substan- tially less than the observed displacement of the distribution curves for W and Mo. This surface depletion for large impurity atoms whose concentrations do not exceed the solubility limit' is in agreement with concepts on ra&ation-induced impurity segregation in fcc metals [8, 9] as derived from tine 0'! 'noto-Widersi i model [10]. On this model, large atoms in substitution positions diffuse i the opposit sense to the radiation vacancies and leave the surface'zone. Therefore, the redistribution of the major components and alloying elements produced by ion bombardment in this nickel-rich alloy is favorable as regards preventing the plasma from contamination by heavy elements such as tungsten and molybdenum, whose permissible concentra- tions in a deuterium-tritium plasma are 0.1 and 1%, correspondingly [11]. The W and Mo have a favorable effect on the heat resistance of the nickel-rich alloy, which opens up prospects for using it as a material of the first wall in a fusion reactor instead of expensive molyb- denum alloys. 1. V. M. Gusev, N. P. Busharov, and S.. M. Naftulin, Prib. Tekh. Eksp., 64, 19 (1969). 2. V. M. Gusev.et al., Rad. Effects, 40, 37 (1979). 3. V. M. Gusev et al., Preprint IAE-3133, Moscow (1979). 4. S. Das and M. Kaminsky, "Radiation blistering in metals and alloys," in: Rad. Effects on Solid Surf., Washington (1976), p. 119.. 5. V. F. Zelenskii et al., Nuclear Science and Engineering: Series Radiation-Damage Physics and Radiation Materials Sciences [in Russian], Issue 3 (14), Moscow (1980), p. 48. 6. R. Hanneman and T. Anthony, Acta Metalurgica, 17, No. 9, 1133 (1969). 7. G. Wehner, in: Proc. 2 Coll. Int. de Pulver. Cathodique, Nice (1976), p. 1. 8. L. Rehn et al., J. Nucl. Mater., 74, 242 (1978). 9. R. Piller and A. Marwick, J. Nucl. Mater., 71 309 (1978). 10. P. Okamoto and H. Widersich, ibid., 53, 336 (1974). 11. N. V. Pleshivtsev, Physical Problems in Cathode Sputtering: IAE Survey [in Russian], Moscow (1979), p. 23. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 REMOVAL OF MONO-2-ETHYLHEXYLPHOSPHORIC ACID FROM SOLID EXTRACTION AGENTS WITH DI-2-ETHYLHEXYLPHOSPHORIC ACID V. B. Dedov, P. S. Trukhlyaev, UDC 543.544.4 B. S. Kalinichenko, and I. K. Shvetsov When solid extraction agents are used with di-2-ethylhexylphosphoric acid (D2EHPA), the organic. phase accumulates the destruction product from the extractant: mono-2-ethylhexyl- phosphoric acid (M2EHPA), which has higher extraction performance and thus hinders the sol- vent-extraction and chromatographic separation of the transplutonium, rare-earth, and other elements (V. S. Smelov et al., Atomnaya Energiya, 51, No. 4, 231 (1981)). One can remove the M2EHPA from the solid extraction agent containing D2EHPA to regener- ate it by washing the column with distilled water. Figure 1 shows that the content of the M2EHPA and also that of the D2EHPA in the matrix decreases when the matrix is chloromethylated styrene and distilled water is passed through the column. Clearly, washing with 100 column volumes of water can reduce the M2EHPA content in the extraction agent to 0.6% from an initial concentration of 6%, and at that level the effects of the component will be inappreciable, while the loss of D2EHPA will not exceed 10-12%. To remove the M2EHPA from a solid extrac- tion agent prepared on a polysorb matrix it is sufficient to wash with 10 column volumes of water. The elution of the extraction agent from the matrix no matter what the method of produc- tion indicates that the bond between these two components is not strong. This has been es- tablished qualitatively for all solid extraction agents. X90 1 90 900 900 10000 No. o f column volumes of wash water Fig. 1. Dependence of the content A of a com- ponent in the solid extraction agent on wash- ing with water: 0- D2EHPA:? - M2EHPA. Translated from Atomnaya Energiya, Vol. 54, No. 2, p. 135, February, 1983. Original article submitted June 14, 1982. 0038-531X/83/5402-0157$07.50 ? 1983 Plenum Publishing Corporation 157 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 NEUTRON ACTIVITY OF THE EARTH AND THE Cl-Ar NEUTRINO EXPERIMENT I. R. Barabanov, V. N. Gavrin, UDC 539.12.164 G. T. Zatsepin, I. V. Orekhov, and L. P. Prokop'eva In connection with radiochemical experiments for the detection of solar neutrinos which have already been conducted [1] as well as those being planned [2], the urgency of the ex- perimental investigation of the flux of fast neutrons from rocks has increased. Incident on a detector, fast neutrons give by means of the two-stage reaction (n, p), (p, n) the very same isotope as do neutrinos as the result of the reaction (Z, N) + v + (Z + 1, N - 1) + e-. The results obtained in more than 10 years with a chlorine-argon detector of solar neu- trinos at the Brookhaven National Laboratory of the USA have shown that the flux of solar neutrinos is at least 3-4 times less than the expected value, and the questions of background effects require attentive study. The results of an investigation of the flux of fast neutrons from the ground are supplied in this article, and their possible background contribution to the neutrino experiment is discussed. Approximate estimates show [3] that the flux density of fast neutrons from rocks with the usual uranium and thorium content should be '-.l neutron/ (cm2?day) but can vary greatly de- pending upon the content of light elements (Li, Be, B) and water in a rock. The main source of neutrons are (a, n) reactions with the elements C, N, 0, Mg, Al, and Si under the action of a particles of the uranium and thorium series. Fast neutrons arising as a result of the spontaneous fission of 296U still make approximatley a 10% contribution. The isolation and measurement of such a small flux density of fast neutrons as 1.5 neu- tron/(cm2-day) is scarcely possible with the help of the usual neutron detectors (scintilla- tors, helium (9He) counters, etc. [4]) against an emission back ?round greater by a factor of 106. Therefore, the radiochemical method using the reaction ?Ca(n, a)97Ar was selected for measurement of the flux of fast neutrons from rocks. This reaction is convenient in that its cross section [5] increases sharply at E = 3 MeV and is actually of a threshold nature. The technique for the extraction of the small amount of 97Ar formed (right down to tens of atoms) and the measurement of its activity has been well developed. People usually use an aqueous solution of some calcium salt or other in detectors of this kind, and they extract the 37 Ar formed by blowing a gas through the liquid, e.g., helium. However, water is a good moderator of fast neutrons, which leads to a decrease in the sensitivity of such detectors. Taking account, moreover, of the fact that the detector must contain '100 kg of Ca to achieve the required sensitivity, it would be difficult to work with such a detector under subterran- ean conditions. Therefore, a different way was selected to create the detector. It has been shown previously [6] that the 37Ar formed as the result of neutron irradia- tion can be extracted with practically 100% efficiency from some solid compounds of calcium organic compounds, e.g., from CaC204 (calcium oxalate), by a trap with activated carbon at a temperature of 196?C. A radiochemical detector of fast neutrons [7] was created on the basis of this result. The detector consists of a hermetically sealed cylinder 60 cm in diameter and 120 cm in height filled with CaC2O4 dried in advance, containing 60 kg of Ca. A valve is mounted on the top end of the cylinder to evacuate the detector and extract the 37 Ar formed. A manometer tube is mounted in the lower part of the cylinder to control the pressure. After exposure of the detector in the neutron flux being measured, a trap with activated carbon ('Ll g) at a temperature of -196?C, in which the 97Ar formed is adsorbed, is connected to the upper valve. Subsequent purification of the 97Ar and measurement of its activity are described in [8]. Calibration measurements with neutrons from a Pu-Be source have shown that the process of extracting 37 Ar from the detector is subject to the law Q = Qo[l - exp (-t/to)], where to = 1.42 h. Thus, 95% of the 97Ar is extracted in 4 h. Translated from Atomnaya Energiya, Vol. 54, No. 2, pp. 136-137, February, 1983. Origi- nal article submitted June 16, 1982. 158 0038-531X/83/5402-0158$07.50 Q1983 Plenum Publishing Corporation Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 1 R- 107, 37Ar atoms/neutron 110 18 01 / I 9Z Z 4 6 8 1OEn, 20 4,0 6,o H,0 1(;0 E, MeV A-Me V n ' Fig. 1 Fig. 2 Fig. 1. Dependence of the efficiency of the calcium detector on the neutron energy (a monoenergetic isotropic source distri- buted uniformly over the detector surface). Fig. 2. Dependence of the 97Ar yield per neutron R on neutron energy in the detec- tor of R. Davis. Fig. 3. Cumulative energy spectrum of neu- trons in granite from (a, n) reactions and spontaneous fission of uranium; a, b) con- tributions of the (a, n) reactions are ob- tained with assumptions 1 and 2, respec- tively. The spectra were calculated for the concentration ratio (mm : mU = 5 : 1). - The dependence of the detector efficiency on the neutron energy has been calculated by the Monte Carlo method. The computational results are presented in Fig. 1. The sensitivity of the detector created to the flux density of fast neutrons is 't'3 x-10-3 neutrons/(cm2?day) [7] and is sufficient, with a large reserve, for measurement of neutrons from terrestrial soil. Measurements have been made with the help of,this detector in subterranean rooms of the Baksansk Neutrino Observatory [9] at a depth of 'ul000 m below sea level. The average rate of formation of 37Ar in the equipment turned out to be (0.89 ? 0.09) 37Ar atoms/h. If one approximates the detector efficiency as shown in Fig. 1 by the value 1.7% for E > 3 MeV and 0% for E < 3 MeV, then we obtain N = 0.045 day-1'cm :2 for the fluxof neutronswith E~> 3MeV. The data obtained were used to estimate the background of a radiochemical detector of solar neutrinos from the neutrons of rocks. The Brookhaven chlorine detector is a tank 10 m long and 6 m in diameter filled with 610 tons of C2C14 (perchloroethylene). The reaction recorded is 37C1(v, e-)97Ar. The background under the action of fast neutrons arises mainly as a result of the chain of reactions 31C1(n, p)35S and 35C1(p, n)87Ar. The efficiency of formation of 97Ar in this detector by neutrons of different energies was calculated by the Monte Carlo method. The elastic and inelastic cross sections of neutrons as well as the nu- clear reactions by all the nuclei making up the detector were taken into account in the cal- culation. All necessary neutron cross sections were taken from [5]. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 The spectrum of the protons generated as a result of the (n, p) reaction were calculated from the statistical model developed in [10]. According to this model, the spectrum of the protons emitted from a compound nucleus is described by the Weiskopf formula P (E) _ mh3 aE 1W (f)/W(i)1; (1) as recommended in [10], the empirical dependence a=a7(1+C)(I-KV/E) (2) is taken as the cross section a of the inverse reactic=i, where aq is the geometric cross sec- tion, C and K are constants dependent or 'she charge oS the nucleus, and V is the coulomb bar- rier. For the nuclei of interest to us i. = 0.368 and = 0.516 in accordance with [10]. The yield of 97Ar per proton (I, nucleus/proton) was calct.-ated from the formula I =ao W (E) I ld?/dx~ de dE, (3) Ethr Ethr where W(E) is the proton spectrum and Qpn(E) is the perturbation function of the (p, n) re- action with 97C1 (taken from [11]). The criterion for the correctness of the calculation was a comparison of the calculated 37 Ar yield under the action of neutrons with the spectrum of a Ra-Be source with the experimentally obtained yield upon irradiation of the detector by the Ra-Be source. The experimentally obtained value is equal to (7.5 ? 0.4) X 10-7 37Ar atoms/neutron, and the calculated value was 5.2 x 10-7 97Ar atoms/neutron. Taking account of the indeterminacy of the neutron constants and the error of the experimental value, such agreement should be considered to be rather good. But one can achieve absolute agreement by selecting the constant K in formula (2). With K = 0.634 ? 0.03 agreement is obtained between the calculated and experimental yields. The value K = 0.634 was used in the subsequent cal- culations. The calculated dependence of the 97Ar yield on neutron energy is presented in Fig. 2. It is evident from a comparision of Figs. 1 and 2 that the dependence of the effective for- mation of 97Ar in perchloroethylene and CaC2O4 under the action of neutrons is different, generally speaking, and it is necesary to use data on the spectrum of neutrons from the ground based on the data of the calcium detector to calculate the background from fast neu- trons in the chlorine detector. Neutron generation spectra in granite were calculated in [12] with two assumptions re- garding the energy carried off by the neutron from a compound nucleus: 1. The emitted neutrons have a Maxwellian energy distribution in the center-of-mass system. 2. The neutrons carry off energy in the center-of-mass system equal to the energy of the output channel. The first case is evidently similar to the actual one, whereas the second case gives an extremely "hard" spectrum. The spectra in Fig. 3 were normalized to the results obtained with the calcium detector; the integral of the product of the spectrum normalized by the efficiency of formation of 97Ar in the chlorine detector (`see Fig. 2) gives the desired back- ground of the chlorine detector under the action of fast neutrons. The results of the inte- gration for spectra a and b, respectively, are equal to 0.079 and 0.098 37 Ar atoms/day. 'Thus, the amount of background in the chlorine experiment depends weakly on the assump- tion about the fast neutron spectrum. Even with extreme assumptions about the spectrum, the results differ by only 20%. One should compare the values obtained to the rate of 97Ar for- mation under the action of neutrinos obtained by R. Davis (0.40 97Ar atoms/day),-i.e., the background from fast neutrons of granite without shielding in the detector of R. Davis should amount to 20% of the effect. According to the estimates of R. Davis, the background from fast neutrons was ti10% prior to installation of shielding in the chamber where the detector is located. Emax E Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 LITERATURE CITED 1. R. Davis, D. Hammer, and K. Hoffman, Phys. Rev. Lett., 20, 1205 (1968). 2. G. Zatsepin, in: Proc. Int. Conf. on Neutrino Physics and Neutrino Astrophysics, Vol. 1, Baksansk Valley (1978), p. 20. 3. I. A. Arshinov, V. I. Glotov, and G. T. Zatsepin, in: Proc. Int. Conf. on Low Radioact. Meas. Appl., Bratislava (1977), p. 29. 4. G. V. Gorshkov, V. A. Zyabkin, and R. N. Yakovlev, Yad. Fiz., 13, No. 4, 791 (1971). 5. D. Gullen et al., The LLL Evaluation Nuclear Data Library (ENDL), UCRL-50400 (1976), p. 15. 6. I. R. Barabanov et al., Izv. Akad. Nauk SSSR, Ser. Fiz., 40, 1050 (1976). 7. I. R. Barabanov et al., At. Energ., 47, No. 4, 273 (1979). 8. I. R. Barabanov et al., At. Energ., 37, No. 6, 503 (1974). 9. I. R. Barabanov et al., At. Energ., 50, No. 1, 59 (1981). 10. I. Dostrovsky et al., Phys. Rev., 116, No. 3, 683 (1959). 11. R. Kishore et al., ibid., 12 No. 1, 21 (1975). 12. V. I. Glotov, Kr. Soobshch. Fiz., No. 7, 75 (1970). ION SYNCHROTRON COMPLEX OF V. G. KHLOPIN RADIUM INSTITUTE N. A. Perfilov, V. P. Shilov, V. P. Eismont, V. L. Auslender, V. N. Lazarev, and B. L. Faktorovich Existing accelerators can deliver heavy ions with an energy E x..10 MeV/nucleon and E 1 GeV/nucleon. The range of intermediate energy values (several tens and hundreds of mega- electron volts) remains open: at the present time there are no special accelerators capable of producing continuously variable ion beams of such energy. Moreover, this range is now of extreme interest, since hopes for research into the most fundamental properties of nuclear matter and product nuclei are linked with it: the form of the equation of states (including the problem of "superdense" isomers), the existence of a critical nuclear. temperature, maxi- mum rotational moments, the nature and magnitude of nuclear "friction," etc. At an energy of several tens of megaelectron volts per nucleon, important limits of the physics of heavy ions are reached, e.g., the sound barrier (E 15-20 MeV/nucleon) and the Fermi energy (30- 40 MeV/nucleon). It is expected that hydrodynamic effects of compression of nuclear matter begin at the first, while at the second the mechanism of nuclear collision changes signifi- cantly, i.e., there is a transition from interaction between nuclei as a whole to interaction between the nucleons constituting them. It is not surprising, therefore, that the aim of almost all heavy-ion accelerator pro- jects now is to attain energies of tens and hundreds of megaelectron volts per nucleon. Such projects exist in our country [1, 2] and abroad [3]. In this paper we report on an ion-synchrotron complex for an energy of 50 MeV/nucleon, under construction at the V. G. Khlopin Radium Institute in Leningrad. Besides the synchro- tron, the complex includes large experimental facilities and systems for automatic control of the accelerator and research on it. The ion synchrotron [4, 5] has been designed to produce beams of accelerated protons and completely ionized atoms (Z/A'- 0.5) in the continuously variable ranges 10-200 and 0.3- 50 MeV/nucleon, respectively. The energy homogeneity of the extracted beams is not less than 1% with a current-pulse duration of 20 psec and pulse-repetition rate of 50 Hz. The inten- sity of the extracted proton beam is up to 1011 particles/sec, while that of the beam of mul- tiply charged ions is no less than 108 particles/sec. The synchrotron is a weak-focusing ring with four straight gaps and an equilibrium-orbit radius of 1.4 m. Bending-focusing magnets, an' accelerating cavity, and a system for particle injection and extraction are in- stalled inside a common vacuum-tight housing with an outer diameter of "4.2m. The design average vacuum in the synchrotron orbit is no less than 6.7-10-" Pa (ti5.10 ' torr) [61. Translated from Atomnaya Energiya, Vol. 54, No. 2, pp. 137-139, Feburary, 1983. Origi- nal article submitted July 9, 1982. 0038-531X/83/5402-0161$07.50 ? 1983 Plenum Publishing Corporation 161 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 2 1k , 15 ,16 Fig. 1. Layout of accelerator and experimental facilities: 1) accelerator hall; 2) experimental hall No. 1; 3) experimental hall No. 2; 4) auxiliary pre- mises containing the power equipment of the magnet power supply; 5) protective gates and doors; 6) walled-up openings; 7) cable channels; 8) ion source; 9) injector; 10) main ring of accelerator; 11) magnet power supply; 12) bending magnets; 13.) quadrupole lenses; 14) "on-line" mass spectrometer; 15) ion source of mass spectrometer; 16) ion guide; 17) "Globus" universal nuclear-reaction chamber. Particles are injected into the synchrotron from a pulsed rf linac [7], formed by a quarter-wave coaxial resonator shortened by the capacitance of the drift tube. The energy of protons at the exit from the resonator is 1.5 MeV, while that of'ions is 0.6-1 MeV/nucleon with Z/A 0.5. For use of the injector in the domain of radiation research, a multibeam electron-ion accelerator was suggested [8] and designed [9] on its basis; this accelerator makes it pos- sible to obtain, along with an ion beam, one or several electron beams whose total power at an energy of "0.6 MeV is 10-12 W. In order to expand its capabilities, the accelerator has been supplemented with a second drift tube [10] and at the present time a version of the injector with two independent ion-acceleration channels is being built. Protons are extracted from a source with a Penning discharge with a cathode needle in the discharge chamber [11]. The duration of the proton-current pulse is 4 usec and the cur- rent is 90 mA at an extraction voltage of 50 kV and a pulse-repetition rate of 50 Hz. A "warm" version [12] of an electron-beam'source, suggested by E. D. Donets [13] and developed at the Institute of Nuclear Physics, Siberian Branch of the Academy of Sciences of the USSR, was used as the source of multiply charged ions. The intensity of the source of multiply charged ions is up to 10? nuclei/pulse for elements up to argon. The ionization cycling rate is up to 50 Hz. Just like the multibeam version of the pulsed rf linac, intro- ducing a new quality into radiation research and technology, the heavy-ion source is also of great importance in itself. By means of two bending magnets and several quadrupole-lens doublets, the beam accelerated in the synchrotron is divided and directed into two indepen- dent experimental halls (see Fig. 1). Large experimental facilities in these halls consti- tute an essential part of the complex being built. It is intended to use them for a wide range of basic research on the dynamics of the interaction of complex nuclei, for the study Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 of the properties of compressed and heated nuclear matter, for the study of the decay of pro- ducts formed in reactions,. and for applied work on radiation physics and chemistry. The principal pieces-of equipment are the "Globus" automated nuclear-reaction chamber, a high- resolution mass spectrometer, a double beta-spectrometer with a superconducting magnet of the. solenoid type, and a high-vacuum chamber for studying the effect of heavy ions on matter [4]. The ion-synchrotron complex is a unique automated system whose component parts - the accelerator and the experimental facilities - are controlled and linked together by a mini- computer. Automation of the accelerator control improves the physical characteristics of the machine; eliminates unproductive losses of time, especially for resetting the accelerator re- gimes; and substantially improves the quality of investigations by increasing their accuracy and the volume and reliability of the data obtained. At the present time the accelerator is controlled with an M-6000 computer [14], but in the future it will be replaced with an SM-2 general-purpose computer. An automated system for scientific research on the synchrotron is being built on,the basis of two SM-2 general-purpose computers. At the present time the main systems of the ion synchrotron are undergoing comprehen- sive alignments at the Institute of Nuclear Physics, Siberian Branch of the Academy of Sciences of the USSR. The complex is to be put into operation in 1984. LITERATURE CITED 1. A. M. Baldin at al., Preprint No. 9-11796, Joint Institute for Nuclear Research, Dubna (1978). 2. Yu. Ts. Oganesyan, Preprint No. R9-12843, Joint Institute for Nuclear Research, Dubna (1979). 3. J. Martin, IEEE Trans. Nucl. Sci., NS-26, No. 3, 3645 (1979). 4. V. L. Auslender et al., Preprint No. RI-56, Radium Institute, Leningrad (1976). 5. V. G. Abdul'manov, in: Proceedings Tenth Int. Conf. on High-Energy Charged-Particle Accelerators [in Russian], Vol. 1, July, 1977, Serpukhov (1977), p. 345. 6. V. L. Auslender et al., Preprint No. 79-5, Institute of Nuclear Physics, Novosibirsk (1979). 7. V. G. Abdul'manov et al., Preprint No. 78-25, Institute of Nuclear Physics, Novosibirsk (1978). 8. V. L. Auslender et al., Inventor's Certificate No. 641851, Byull. OIPOTZ, No. 3, 276 (1980). 9. V. L. Auslender et al., in: Proceedings Third All-Union Conference on the Application of Charged-Particle Accelerators in the National Economy [in Russian], Vol. 1, Leningrad (1971), p. 114. 10. V. L. Auslender et al., Inventor's Certifice No. 720838, Byull. OIPOTZ, No. 9, 305 (1980). 11. V. L. Auslender, V. N. Lazarev, and A. D. Panfilov, Prob. Tekh. Eksp., No. 4, 33 (1979).. 12. V. G. Abdul'manov et al., in: Proceedings Sixth All-Union Conference on Charged-Particle Accelerators [in Russian], Dubna, October 11-13 (1978),-Dubna (1979), Vol. 1, p. 98. 13. E. D. Donets, Inventor's Certificate No. 248860, Byull. OIPOTZ, No. 24 (1969). 14. V. L. Auslender et al., Preprint No. 78-89, Institute of Nuclear Research, Novosibirsk (1978). Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 ,O~ Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020002-8 RUSSIAN JOURNALS IN THE PHYSICAL AND -MATHEMATICAL SCIENCES AVAILABLE IN ENGLISH TRANSLATION ALGEBRAAND LOGIC HYDROTECHNICAL CONSTRUCTION Algebra i Logika Gidrotekhnicheskoe Stroitel'stvo Vol. 21, 1982 (6 issues) ... ..... ............... . ... $270 Vol. 16, 1982 (12 issues) ........................ :... $305 ASTROPHYSICS INDUSTRIAL LABORATORY Astroftzika Vol. 18, 1982 (4 issues) .............................. $320 Vol. 48, 198 (12 issues) ..!....................... $400 AUTOMATION AND REMOTE CONTROL Avtomatika i Telemekhanika Vol. 43, 1982,(24 issues) issues) ...................... $495 COMBUSTION, EXPLOSION, AND SHOCK WAVES Fizika Goreniya i Vzryva Vbl. 18, 1982 (6 issues) . $345 Kosmicheskie Issledovaniya Vol. 20, ,1982 (6 issues) ......... .................. $425 CYBERNETICS Kibernetika Vol. 18, 1982 (6 issues) ....................... .... $345 DIFFERENTIAL EQUATIONS Di,Uerentsial `nye Uravneniya Vol. 18; 1982 (12 issues)............................. $395 DOKLADY BIOPHYSICS' Doklady Akademii Nauk SSSR Vols. 262-267, 1982 (2 issues) ....................... $145 FLUID DYNAMICS Izvestiya Akademii Nauk SSSR, Mekhanika Zhidkosti i Gaza INSTRUMENTS AND EXPERIMENTAL TECHNIQUES Pribory i Tekhnika Eksperimenta .Vol. 25, 1982 (12 issues) ............................ $460 JOURNAL OF APPLIED MECHANICS AND TECHNICAL PHYSICS Zhurnal Prikladnoi'Mekhaniki i,Tekhnicheskoi Fiziki Vol. 23, 1982 (6 issues) . ............................ $420 JOURNAL OF APPLIED SPECTROSCOPY Zhurnal Prikladnoi Spektroskopii Vols. 36-37 (12 issues) I ....... .. $420 l JOURNAL OF ENGINEERING PHYSICS Inzhenerno fz cheskii Zhurnal Vols-42-43, 1982 (12 issues) ........:..:............ $420 , JOURNAL-OF SOVIET LASER RESEARCH' ` A translation of articles based on the best Soviet research in the, -field of lasers. 'Vol. 3, 1982 (4 issues) ............... .................... $95 JOURNAL OF SOVIET.,MATHEMATICS A translation of Itogi Nauki i Tekhniki and Zapiski Nauchnykh Seminarov Leningradskogo Otdeleniya , Vol. 17, 1982 (6 issues) ............. :............. :. $380 t FUNCTIONAL ANALYSIS AND ITS APPLICATIONS - Funktsional nyi Analiz i Ego Prilozheniya Vol. 16, 1982 (4 issues) ............................... $320 Matematicheskogo Instituta im. V. A. 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