Soviet Atomic Energy Vol. 40, No. 1

Body: _ __.1.-.Awrisk111111111t ? Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 4 ,e" 3 Russian Original Vol. 40, No. 1, January, 1976 b) July, 1976 SATEAZ 40(1) 1-118 (1976) SOVIET ATOMIC ENERGY ATOMHAll 3-HEP11411 (ATOMNAYA kNERGIYA) TRANSLATED FROM RUSSIAN CONSULTANTS BUREAU, NEW YORK Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10 CIA-RDP10-02196R000700070001-0 SOVIET ATOMIC ENERGY Soviet Atomic Energy, is abstracted or in- dexed in Applied Mechanics Reviews, Chem- ical Abstracts, Engineering Index, INSPECT Physics Abstracts and Electrical and Elec- ' tronics Abstracts; Current Contents, and Nuclear Science Abstracts. -/- Soviet Atomic Energy is a cover4o-cover translation of Atomnaya tnergiya, a publication of the Academy of Sciences of the USSR. An agreement ,with the Copyright Agency of the USSR (VAAP) makes available both advance copies 'of the Russianlournal and originsl glossy photcigraphsland artwork. This serves to decrease the necessary, time lag between publication of the original and publication of the translation and helps-to improve the quality of the latter. The translation began with the first_ issue of the Russian journal.- Editorial Board of Atomnaya Energia: Editor: M. D. Millionshchikov ? Deputy Director 1. V. Kurchetov Institute of Atomic Energy Aced?my of Sciences of the USSR Moscow, USSR ? Associate Editor: N. A. Vlasov A. A. Bochvar N. A. Dollezhal' - V. S. F6rsov I. N. 091ovin V. F. Kalinin A. K. Krasin V. V: Matveev M. G. MeshcherySkov V. B. Shevchenko V. I. Smirnov A. P. Zefirov -,Copyright ? 1976 Plenum Publishing Corporation, 227 West 17th Street, New York. -N.Y. 10611.- All rights reserved. No article contained herein may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. . Consultants Bureau journals appear _about six months after the publication of the original Russian-,issue. For bibliographic'-accuracy,- the English issue published by Consultants Bureau -carries the same number and date as the original Russian from which it was translated. 'For example, a ,Russian issue published in 'Deeember 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. 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 $107.50 per volume (6 'Issues) 2 volumes per year _ Prices somewhat higher outside the United States. Single Issue: $50 Single Article: -$15 _CONSULTANTS BUREAU, NEW YORK AND LONDON 227 West 17th Street New York, New York 10011 Published monthly. Second-class postage paid at Jamaica, New 'York 11431. Declassified and Approved For Release 2013/04/10: 6IA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya July, 1976 Volume 40, Number 1 January, 1976 CONTENTS ARTICLES Engl./Russ. Role of Gas as a Coolant in the Development of Nuclear Power Stations ? E. P. Anan'ev and G. N. Kruzhilin Experience in the Use of a Nuclear Reactor in the Noril'sk Mining-Metallurgical Complex ? V. N. Nikitin, V. N. Pavlova, A. I. Petrov, and A. M. Shchetinin. Testing of Experimental BN-600-Type Fuel Elements in the BOB-60 Reactor up to Different Burnups ? M. M. Antipina, Yu. K. Bibilashvili, I. S. Golovnin, V. M. Gryazev, E. F. Dyvydov, G. V. Kalashnik, A. V. Medvedev, T. S. Men'shikova, V. S. Mukhin, A. A. Petukhov, A. V. Sukhikh, V. N. Syuzev, L. I. Sytov, and V. L. Timchenko Predicting the Efficiency (Serviceability) of Oxide Fuel Elements for Fast Sodium Reactors ? I. S. Golovnin and Yu. I. Likhachev In-Reactor Measurements of the Modulus of Elasticity of Uranium Dioxide ? V. M. Baranov, Yu. K. Bibilashvili, I. S. Golovnin, V. N. Kakurin, T. S. Men'shikova, Yu. V. Miloserdin, and A. V. Rimashevskii Hydrogen Embrittlement of Vessel Steels ? V. V. Gerasimova and E. Yu. Rivkin. . . Nonsteady-State Space-Energy Spectrum of Neutrons in a Heavy, Weakly Inhomogeneous Medium, Allowing for Neutron Capture ? E. V. Metelkin Experiments on Cooling by Electrons ? G. I. Budker, Ya. S. Derbenev, N. S. Dikanskii, V. I. Kudelainen, I. N. Meshkov, V. V. Parkhomchuk, D. V. Pestrikov, A. N. Skrinskii, and B. N. Sukhina The Use of Microwave Methods in the Dosimetry of Impulse Fluxes of Ionizing Radiation ? Yu. A. Medvedev, N. N. Morozov, B. M. Stepanov, and V. D. Khokhlov DEPOSITED PAPERS Universal Absorption Curves for a Sinusoidally Modulated Electron Beam ? R. Ya. Strakovskaya, I. R. Entinzon, and G. N. P'yankov Dosimetry on an Object Rotating in an Electron Beam ? B. Ya. Strakovskaya, G. N. Pyankov, and Yu. F. Golodnyi Calculations on Weakly Interacting Systems ? V. P. Ginkin LETTERS TO THE EDITOR Self-Acceleration Experiment of a Strong Electron Beam in a Ferrite Accelerating Structure ? V. V. Zakutin, N. N. Nasonov, A. A. Rakityanskii, and A. M. Shenderovich Determination of the Half-Life of 238PU ? V. G. Polyukhov, G. A. Timofeev, P. A. Privalova, V. Ya. Gabeskiriya, and A. P. Chetverikov The High-Temperature Thermal Diffusivity and Electrical Resistivity of Yttrium and Gadolinium ? I. I. Novikav and I. P. Mardykin Effect of Implanted Space Charge on Particle Range Distribution ? V. S. Remizovich and A. I. Budenko 1 3 9 11 14 16 26 27 37 37 40 40 45 45 50 49 55 53 59 56 60 57 62 57 63 59 66 61 69 63 72 64 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Yields of 99mTc, 96Tc, and 971nTc from Irradiation of Molybdenum and Niobium ? P. P. Dmitriev, G. A. Molin, Z. P. Dmitrieva, and M. V. Panarin Stimulating Isotopically Selective Heterogeneous Reactions with Laser Light ? V. D. Borman, B. I. Nikolaev, and V. I. Troyan Efficiency for Conversion of Electrons into Positrons at 20-70 MeV CONTENTS (continued) Engl./Russ. 75 66 78 69 ? V. A. Tayurskii 80 70 Dependence of Asymmetry in the Photofission of 233U and 239Pu on the Maximum Bremsstrahlung ? M. Ya. Kondrat'ko, V. N. Korinets, and K. A. Petrzhak 83 72 Synthetic Pitchblende: Composition, Structure, and Certain Properties ? V. A. Alekseev and R. P. Rafal'skii 85 73 Measurement of the Energy Dependence of 71233U in the 0.02-1-eV Region ? V. A. Pshenicluiyi, A. I. Blanovskii, N. L. Gnidak, and E. A. Pavlenko . ? ? 89 76 INFORMATION Next Problems in the Development of Oxide Fuel Elements for Fast Power Reactors ? I. S. Golovin 91 78 CONFERENCES AND SYMPOSIA Third Conference on Neutron Physics ? A. I. Kal'chenko, D. A. Bazavov, B. I. Gorbachev, A. L. Kirilyuk, V. V. Kolotyi, V. A. Pshenichnyi, A. F. Fedorova, and V. D. Chesnokova 94 80 Scientific Seminar on the Complex Optimization of Power Installations ? Yu. I. Koryakin 98 82 Soviet?American Seminar on Fast-Breeder Reactors ? E. F. Arifmetchikov 101 84 All-Union Conference on "Development and Application of Electron Accelerators" ? A. N. Didenko and V. K. Kononov 104 85 7th International Conference on Cyclotrons and Their Applications ? N. I. Venikov. 107 87 Conference on Laser Engineering and Applications ? V. Yu. Baranov and N. G. Koval'skii 1.10 89 Soviet?American Working Meeting on Open Traps ? D. A. Panov 113 90 International Congress on Engineering Chemistry, Chemical Engineering, and Seventh All-Union Conference on Scintillation Technology-- O. P. Sobornov 115 92 International Congress on Engineering Chemistry, Chemical Engineering, and Automation ? V. N. Koshkin 116 92 REVIEWS V. I. Sidorov, N. I. Loginov, and F. A. Kozlov ? Fundamentals of Heat Physics in Atomic Power Installations ? Reviwed by M. Kh. Ibragimov 117 94 L. S. Sterman, L. T. Sharkov, and S. A. Tevlin, Thermal and Nuclear Power Stations ? Reviewed by Yu. I. Klimov 118 94 The Russian press date (podpisano k pechati) of this issue was 12/23/1975. Publication therefore did not occur prior to this date, but must be assumed to have taken place reasonably soon thereafter. Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 ARTICLES ROLE OF GAS AS A COOLANT IN THE DEVELOPMENT OF NUCLEAR POWER STATIONS E. P. Anan'ev and G. N. Kruzhilin UDC 621.039.524.46.034.3 Gas-cooled reactors, of both the thermal and fast varieties, are again exciting considerable interest. Gas cooling in general, and helium cooling in particular, is in fact now being considered and developed as a technical alternative to cooling with molten sodium in fast reactors, since as regards the breeding of plutonium and the generation of water vapor for the power cycle the parameters of the two coolants are sufficiently close to make no practical difference. Such difference as exists is mainly of a technological character. There is of course a considerable difference in the arrangements for the cooling of the active zone in any hypothetical emergency, since the heat capacity of the gas in this zone is quite negligible com- pared with that of sodium. As regards thermal reactors, helium cooling appears to be quite a promising solution; in this case the temperature of the gas at the outlet from the reactor may extend to 750-850?C or even higher. There is thus a real possibility of generating steam at high temperatures and pressures, such as those charac- teristic of modern thermal power installations, giving a power cycle efficiency of 41-42% as compared with 30-34% in the case of the saturated-vapor cycle obtained in installations with water-cooled reactors. Clearly, if we take no account of the possible nuclear superheating of the steam in certain types of reac- tors, this effect will be of no great importance in view of the low fuel component of the cost of electrical power in nuclear power stations with saturated-vapor cycles. Nevertheless, the degree of use of the nuclear fuel improves in the presence of a high efficiency factor and the ejection of heat through the con- denser into the cooling water diminishes, i.e., the "thernial contamination" of the water and the corre- sponding harmful effect on Nature is reduced. Considerable interest also arises in relation to high-temperature helium coolants at the present time in view of the special requirements of a number of branches of industry (metallurgical, chemical, etc.) for high-temperature heat; these are branches which at present consume some 25% of recoverable organic fuel. One of the most important problems facing the power industry will be that of directing nuclear power into the multipurpose complex production of electrical power and other types of product [1]. Here we have a real prospect of using high-temperature gas heated in a nuclear reactor for indus- trial purposes. It is interesting to consider gas-cooled reactors of the kind which have been developed most exten- sively in Britain and France. The development of a gas reactor was started in the USSR even at the end of the 40's. However, only a nuclear power station with an electrical power of 150 MW (incorporating a gas-cooled reactor using natural uranium with a heavy-water moderator), built in Czechslovakia, was in fact fully developed. The first British nuclear power station with a gas reactor (essentially a demonstration model) was started in Calder Hall in 1956 (power 42.0 MW) [2]. The reactor in this power station uses natural uran- ium and has a graphite moderator situated in a steel housing 11.2 m in diameter with a wall thickness of 50 mm. The cooling gas is CO2, the fuel elements are metallic uranium (cylindrical elements with a core diameter of 29.6 mm) covered with a magnesium alloy (Magnox) can. Since the melting point of Magnox is 640?C, the maximum temperature of the fuel-element can is only ?450?C, and correspondingly the tem- perature of the cooling gas at the reactor outlet is 345?C. The reactor has four loops with a pipe diameter Translated from Atomnaya Energiya, Vol. 40, No. 1, pp. 3-11, January, 1976. Original article submitted April 9, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, IV. Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 of 1370 mm; the gas coolant circulates through these at a pressure of 7.8 atm. Each loop has a gas blower with a power of 1500 kW and a steam generator (boiler) which produces superheated water vapor at a pressure of 14 atm and a temperature of 320?C. The steam generator is placed alongside the reactor in a vertical steel housing of diameter 5.5 m and is furnished with a drum separator. Repeated circula- tion of the water in the boiler is effected on the La Monte forced-flow principle. The efficiency of the unit is 19.2%. In order to intensify heat transfer, the fuel-element can is ribbed transversely with a rib height of 12 and a step of 7 mm. This ribbing is certainly very original, since the gas flows longitudinally around the fuel elements in the graphite stack. The mean thermal loading is 1.65 MW/ton of U, or 0.19 .106 kcal/m2?h referred to the surface area of the uranium cores. The groups of boiler tubes are flushed transversely by the gas. In order to intensify heat transfer, these tubes (carbon steel, 51 mm in diameter) also have transverse ribbing. This configuration is pro- duced by the automatic welding of sector ribs on to the tube, the weld contour amounting to 1/3 of the length of the outer circumference of the tube cross section. The total surface of the rib is thus roughly four times greater than that of the outer surface of the tube. Clearly the ribbing of the fuel elements and boiler pipes is in this case perfectly acceptable, since the circulating gas does not contain any particles such as might become deposited on the pipes and ribs in the manner of, for example, coal installations. After the British nuclear power station in Calder Hall had been tested, a number of industrial nuclear power stations were built with analogously constructed uranium-graphite gas reactors using fuel elements of natural uranium and Magnox cans; reactors of this kind have become known as the "Magnox" type [2]. Eighteen industrial power units with a total power of 5000 MW have been constructed in Britain. The unit electrical power of these installations has gradually increased from 138 to 590 MW, with a parallel rise in efficiency from 24.6 to 33.6%. These indices were largely achieved by correspondingly enlarging the active zone, raising the pressure.of the cooling gas in the circuit to 27 atm, and increasing the gas tem- perature at the reactor outlet to 410?C, and also by providing for a corresponding rise in boiler steam temperature and pressure, namely, superheating to 393?C and 100 atm. Another important feature was a further improvement to the fuel-element ribbing, in that the transverse ribs were replaced by an opti- mum configuration of inclined ribs (chevrons). The diameter of the fuel-element core was also reduced to 28 mm. Altogether the mean thermal loading of the fuel elements was increased to 3.15 MW/ton of U or 0.35 .106 kcal/m2.h. The increase in the dimensions of the active zone resulted in a corresponding increase in the size of the reactor vessel. This latter increment, together with the rise in circuit gas pressure, necessitated an increase in reactor wall thickness. The nuclear power station in Sizewell, in which the unit power is 290 MW and the gas pressure 18.56 atm, has the largest steel spherical vessel, 19.4 m in diameter, with a wall thickness of 105 mm. Vessels of this kind are made directly on the site in the form of individual sheets welded together manually. This extremely laborious work is executed by hundreds of welders over several months. Under the conditions existing on the building site severe difficulties were encountered in ensuring a metal temperature sufficient for welding and subsequent heat treatment. Such difficulties increased with vessel wall thickness. The next British gas reactors with higher cooling gas pressures were therefore made not of steel but of prestressed reinforced concrete. The first reinforced-concrete vessel was made for the nuclear power station in Oldbury. This cy- lindrical vessel (internal diameter and height 23.5 and 18:3 m, wall thickness 4.5 m) was designed for a pressure of 24.6 atm. The second was a spherical vessel for the Magnox reactor of the nuclear power station in Wylfa, designed for a unit having an electrical power of 590 MW with an internal diameter and wall thickness of 29 and 3.36 m and a gas pressure of 27 atm. In both reactors the boilers and gas blowers are sited in chambers inside the walls of the concrete reactor vessel, and in this sense the com- position of the equipment may be regarded as integrated. Such a compact arrangement is undoubtedly economical and also entirely reasonable from the point of view of ensuring more reliable hermetization of the circuit. An important technological characteristic of the British Industrial Magnox reactors is the method of fuel recharging, which ensures a high use coefficient and improves the physical characteristics of the fuel cycle. 2 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 The Magnox reactors were constructed and are being used on a two-purpose principle: for breeding plutonium and for producing electrical power. The fuel consumption is accordingly relatively small and amounts to 3000-3600 MW ? day/ton of U. The campaign of the fuel elements in the reactor lasts up to six years [3]. Reliable hermetization of the fuel elements is required for such a long cam- paign. In the few cases in which the hermetization of the fuel ele- ments is breached, the release of the carbon dioxide is quite a slow process. There is accordingly no need for any urgent discharging, and the recharging is planned as an ordinary operation. Up to 1971 there had been only one emergency in Magnox reac- tors involving the melting of the fuel elements in one channel of the Chapel Cross reactor, probably due to the breakdown of a graphite bushing, with the partial closing of the channel cross section. An important point noted in 1968 was the considerable corrosion of the carbon steel in the CO2 atmosphere [3]. It was found that this gas had no effect on the open surfaces, including the vessel and the boiler pipes, but that serious corrosion took place at the contacts between nuts and bolts. Magnetite (Fe304) was formed in volumes two or three times greater than the volume of the actual metal, and in some cases this led to the fracture of the bolts. Carbon steel bolts are used for fixing purposes in the reactors and boilers of all Magnox nuclear power stations. It was therefore decided in 1969 to reduce the pressure at the reactor output to 360?C, and this reduced the power of these nuclear power stations by 20-25%, i.e., to a total value of about 1000 MW. The considerable experience gained in the building and practical exploitation of the British Magnox reactors formed a basis for the manufacture of improved powerful gas reactors of the AGR series, using enriched uranium. The planned cost of electrical power in nuclear power stations using these reactors was estimated as being 40% below the best nuclear power stations with Magnox reactors and 10% below power stations using organic fuel. Typical of this series is the AGE reactor in the Hinckley Point B nuclear power station with a unit power of 660 MW, and cylindrical fuel elements 1016 mm long made of enriched (2.06-2.57% [4]) uranium dioxide (core diameter 14.5 mm). The mean depth of burn-up is 18,000 MW ? days/ton of U. The cooling gas (CO2) circulates at a pressure of 43 atm. The gas temperature at the inlet into the active zone is 282?C and at the outlet 665?C. As a result of this the boilers produce superheated steam at 170 atm and 540?C; secondary superheating of the steam is also effected at 41 atm and up to 540?C. The turbine operates with a condenser pressure of 0.041 atm with a cooling water tem- perature of 12?C. The efficiency is 41.7%. It follows from Fig. 1 that the rod-type fuel elements are assembled in a cassette with a cylindrical graphite housing having internal and external diameters of 190 and 238 mm. The cassette contains 36 fuel elements. Eight cassettes are connected together vertically, using a special suspension, and the group is let down into the channel of the graphite stack through a special adjusting tube. The diameter of the channel is 263 mm, the gap between the channel and the graphite casing of the cassette 12.5 mm, so that the cassette passes freely into the channel. Mounted on the suspension are biological shieding units, sensors measuring the temperature at the outlet from the channel, and a mechanism for changing the flow of gas through the latter. Altogether there are 308 channels; the average electrical power of one channel is 2.13 MW. The step between the channels is 460 mm. The average thermal loading of the fuel elements is 12.2 MW/ton of U (0.67.106 kcal/m2.h). The maximum temperatures in the center of the fuel element and on the can are 1500 and 800?C. These tem- peratures are reached in approximately the middle cross section of the channel, at which the gas tempera- ture is ?480?C, so that the can/gas temperature drop is ?320?C. Taking account of this, we may consider that the heat-transfer coefficient from the can to the gas averages ?2100 kcal/m2.h ? ?C. This large value for the gas coolant is achieved by circulating the gas under high pressure (43 atm at a mean velocity of 12 m/sec) and by creating artificial roughness on the fuel-element can, in the form of a fine rectangular thread with a pitch of 2 mm. The reactor vessel is cylindrical, made of prestressed reinforced concrete with an internal diameter of 18.9 m. The boiler and gas blower are integrated, as in the Oldbury power station. The four direct- flow boilers associated with each reactor are sited in individual chambers, the gas circulation through each being effected by means of two gas blowers. The eight gas blowers ensure a gas circulation with a Fig. 1. Cross section of a fuel cassette in the Hinckley Point nuclear power station. 3 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 5 2 3 total flow rate of 3512 kg/sec for a pressure drop of 2.5 atm. The gas- blower power is 3.5 MW (28 MW for all eight) amounting to 4,1% of the power of the whole unit. The heating surface of the direct-flow boiler is made of three kinds of steel. The economizer part is made of mild carbon steel tubes, in view of the fact that their surface temperature will never exceed 350?C (this is based on corrosion considerations). The evaporating part of the boiler is made of chromium steel tubes with a chromium content of 9%. In the steam-generating part, austenitic chromium-nickel steel tubes are employed. From considerations of the strength and welding requirements, transitional sections of Nicone1-600 are placed between the chromium and austenitic steel tubes; if the tubes were welded directly there would be harmful effects associated with their different thermal expansions and also with carbon diffusion. In order to avoid stress corrosion moisture must be prevented from falling into the austenitic steel tubes. In order to intensify heat transfer the first two kinds of tube are provided with dif- ferent transverse ribbing. No ribbing is provided on the austenitic steels. It is interesting to note that the temperature difference between the gas leaving the reactor and the superheated steam in nuclear power stations of this type is ?130?C, while in Magnox power stations, in which the boilers are made of ribbed carbon tubes, this difference is only 17?C. The temperature of the graphite stack in reactors of this type is under 500?C. This limit is chosen because at higher temperatures the graphite starts oxidizing seriously as a result of interaction with the oxygen released by the decomposition of the CO2 (partly due to radiation). In addition to this, methane is added as an inhibitor to retard the oxidation of the graphite. In order to en- sure the required temperature the graphite is gas-cooled. Hence after leaving the gas blower at a tem- perature of 200?C the gas first pagses through the graphite stack, including the annulae gaps between the channels in the stack and the graphite housings of the fuel cassettes. Then the gas passes through the fuel-element cassettes, in which it is heated to 670?C. The reinforced-concrete reactor vessel is cooled with water passing through a special tube system, the temperature of the concrete being held below 70?C. The construction of the vessel is quite specific owing to the large number of metal tubes with wire cables ?100 mm in diameter passing through them, the tension in these creating the stressed state of the vessel. The calculated tensile-strength reserve factor is taken as three. When testing model vessels on 1/8 and 1/5 scales, the maximum loads were applied to the models. When the pressure rises in the vessel, cracks start propagating, and the rupture process is of a completely different character from that in which walls made of brittle materials fracture. It is therefore quite reasonable to assume that a hypothetical accident (with the swift collapse of the vessel and a total loss of coolant) is much less likely in this case than in that of a system with a steel vessel. As far as radiation safety is concerned, it is especially important to consider the problem of pre- venting the active zone from melting. In reactors of this series, this requirement is chiefly guaranteed by the low temperature of the active-zone graphite. Thus if the gas circulation ceases as a result of (for example) accidental disconnection of the gas blower (when emergency shielding also comes into action) the remaining heat evolution of the fuel elements will be largely carried out into the "cold" graphite stack as a result of radiant heat transfer. According to calculations, in this emergency situation the fuel element can will only rise to 1000?C after 4 h (at which point it may well break down, although its melting point is actually 1400?C). In this time it should be possible to correct the fault in the electrical supply system which caused gas circulation to cease. It is interesting to note that during the development of nuclear power stations containing AGR reac- tors the boiler tubes vibrated; this was prevented, after certain experiments, by appropriate clamping of the tubes. The vibrations arose both from the mechanical pressure of the dense gas on the tubes and from acoustic resonance in the gas itself. In view of the latter the number of gas blowers had to be in- creased to eight (3.5 MW each as in the Hinckley Point station). Even so during tests on the first unit of this nuclear power station the severe noise from the gas blowers gave serious problems; The regulating valves at the outlet from the gas blowers were damaged and severe vibrations of the adjusting tubes took place. It was indicated in August 1974 [5] that because of this the initiation of the unit would be delayed for 20 weeks. So far nothing further has been said about this. // 0 P,3 4 Fig. 2. Block-type fuel element of the reactor in ? the Fulton nuclear power station. 4 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 TABLE 1. Basic Characteristics of Certain Reactors with Gas Cooling Indices Peach Bottom, USA Fort St. Wrayne, USA Fulton, USA Hinckley Point, Britain Coolant Power, MW: thermal electrical Efficiency, % Temperature, ?C: at the inlet at the outlet Height of active zone, m Diameter of active zone, m Internal diameter of vessel, m Gas pressure, atm Helium 115 40 35 340 715 2,28 2,79 ? 24 Helium 842 330 39 405 780 4,75 5,94 ? 49 Helium 3000 1160 38,6 318 740 6,34 8,47 ? 51 CO2 1500 625 41,7 282 665 ? 18,9 43 Earlier it was indicated that the AGE reactor (unit) Dungeness B would be first to start (in 1970). Then in 1972 it was proposed to start the reactor of the Hinckley Point nuclear power station and later to initiate other AGE reactors. According to the data of the International Agency on Atomic Energy [6], however, all the AGE reactors will only just now have started operation; this implies a delay of 5-6 years, involving direct economic losses and certain difficulties in estimating the prospects of such reactors. This is why the discussions regarding future nuclear-power developments held in Britain in 1973-1974 were so acrimonious and prolonged. Itwas pointed out [7] that the main reason for the delay in starting the AGE was the necessity of replacing certain carbon steel parts in these reactors. A number of the first reactors of this series were already largely completed as regards the construction of the reinforced con- crete vessels when it became known that dangerous corrosion of carbon steel took place in Magnox reac- tors working in an atmosphere of CO2 at 360?C. The carbon steel therefore had to be replaced by more heat-resistant varieties in the already completed structures, a task by no means simple. In addition to this [8] difficulties arose with the fuel-element cassettes, since earlier tests on these in the experimental Windscale AGR reactor did not entirely correspond to the operating conditions of in- dustrial reactors in the power system. Subsequently further tests were made, the thermal loads being varied cyclically from 40 to 100%. The tests showed that, under these conditions, and subject to high fuel burn-up values, the uranium dioxide fuel-element cores swelled. However, fuel-element cores with a central aperture swelled in the direction of the aperture and may therefore be regarded as more prom- ising than solid versions. The aperture in the core promotes a more vigorous diffusive evolution of gaseous fission products from the fuel; this might be compensated by changing to coarser 1302 granules in preparing the cores by the sintering method. These tests also showed that carbon was deposited on the fuel-element cans. This is an undesirable feature, since under nominal loads the can temperature may rise to 800?C. In the opinion of the former Director of the British Atomic Energy Authority Sir Christopher Hinton it will be necessary to reduce the gas temperature and pressure very considerably in the AGR reactors, reducing the power to some 60% of the original plan [9]. Naturally this raises the question as to whether it might not be better to use helium as a reactor coolant [10], a possibility not seriously considered before on account of the difficulty of making a circuit with a reinforced-concrete vessel sufficiently gastight. Thermal reactors with helium cooling are intended to heat the gas to higher temperatures than AGE reactors. This leads to a considerable contraction in the heating surface of the boilers, which is quite substantial and expensive when using gas coolants. An increase in the gas temperature also means an increase in the temperature of the fuel elements. Thus in high-temperature reactors fuel elements with stainless steel coatings (as employed in the AGE) cannot be used; for these reactors fuel elements of a special construction with graphite coatings have been developed, and the temperature of such coatings may exceed 1000?C. This means that CO2 cannot be used as a coolant, since it reacts vigorously with graphite even at under 500?C. Hydrogen cooling is also impossible, since at temperatures above 700?C it starts reacting with carbon to form hydrocarbons. Thus helium cooling of the high-temperature reactor is the most favorable method. However, even helium has to be purified from H2, CO, and CO2, since these gases appear in the coolant, for example, when water leaks out of the boiler. Helium is a comparatively rare gas and quite expensive. It is therefore very important to prevent losses such as might occur if the circuit is not hermetized and leaks occur. We may note that the rein- forced-concrete vessel of the Magnox reactor in Oldbury, for example, gave a CO2 leakage of about 1000 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 C_ (TA ,250?C J kg per day at the very beginning, [11], which later of course increased. For helium cooling this is quite unacceptable. Some interesting developments were started by S. M. Feinberg in cooperation with other scientists on a helium breeder reactor of module construction [12]. Helium-cooled reactors have perhaps been most persistently de- veloped in the United States (Table 1). An experimental helium reactor with a power of 40 MW (electrical) was constructed and started in 1966 in Peach Bottom. Then a demonstration helium reactor with an elec- trical power of 330 MW was constructed in Fort St. Wrayne, being in a state of imminent operation in 1974 [10]. Recently the design of a large power reactor with helium cooling and an electrical power of 1160 MW was completed for the Fulton nuclear power station; construction is expected to be complete in 1982 [13]. According to design the Fulton nuclear power unit will have six direct-flow boilers (steam pressure 177 atm, temperature 513?C) and two turbogenerators of 600 MW each. It is well known that US industry already produces turbogenerators of 1300 MW with high steam param- Fig. 3. Construction of eters. In the Fulton nuclear power station design we thus see a clear a fuel element for a high- departure from the principle of the monolithic unit hitherto ruling in temperature helium-cooled United States nuclear power. reactor. In the Fulton nuclear power station (as in Fort St. Wrayne) the fuel elements are made of a graphite block of hexagonal cross section with gage dimensions of 359 mm and a height of 793 mm (Fig. 2). A block-type fuel element of this kind contains 128 channels 16 mm in diameter accommodating the fuel 1, and 72 open channels 21 mm in diam- eter through which the helium passes 2. In addition to this there are six channels 12.7 mm in diameter, containing burning-out absorbent. Eight such fuel elements are arranged over the height of the active zone in the Fulton nuclear power station. These are joined by means of projections 3 and corresponding depressions 4. The central upper depression 5 serves for taking hold of the block when recharging. The fuel employed comprises UC and ThC granules ?0.8 mm in size coated successively with layers of pyrolytic graphite and silicon carbide to a total thickness of ?0.2 mm. The granules are mixed with graphite and resin and pressed into rods 16 mm in diameter and 6 cm long, which are then loaded into the channels of the graphite block. The coating of the granules with layers of graphite and silicon carbide fails to make them completely hermetic; some of the fission fragments diffuse into the coolant, which becomes radioactive. In order to reduce the radioactivity, devices are incorporated to trap the radioactive contaminants with activated char- coal, in con junctionwith cryogenic traps. According to experience gained in the use of this experimental type of reactor, the total radioactivity of the gas in the circuit stabilizes at a level of 20-30 Ci. This indicates that no developing discontinuities occur in the layered can for burn-ups of up to 100.103 MW .day/ton of U, and in this case the cans operate reliably for the temperatures indicated in Table 1. This is an extremely important and favorable result. At higher temperatures, however, when the diffusion of the fission fragments through the solid wall becomes substantial, the radioactivity of the helium in the circuit will be higher. A rise in temperature may also lead to damage in the fuel-element cans. These circumstances (especially the second) prevent increasing the gas temperature when using such fuel elements. The new type of fuel-element construction illustrated in Fig. 3 is perhaps a more promising version; in this the fuel gx:anules are placed in protec- tive graphite cans over the height of the fuel elements in the form of free layers, having the high porosity characteristic of a layer filling. The fuel element is cooled by gas passing through the layers of granules, there being a relatively low temperature difference between the gas and the can. Hence the gas tempera- ture may lie close to the permissible can temperature, exceeding 1000?C. Unfortunately fuel elements of this kind have not yet been developed. However, it is easily seen that the construction will not be a simple matter, since in view of the high gas temperature the housing of such a fuel element will have to be made of heat-resistant ceramic. The development of such fuel elements will take time and effort. No experi- mental data have yet been obtained as to the permissible upper temperature of the can. Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 ID 77 Attention should also be paid to certain other characteristics of high- temperature gas reactors. In particular, the motion of the gas from bottom to top used for the active zones of the Magnox and AGR reactors must now be replaced by motion from top to bottom, since if the gas in the active zone moved from bottom to top and had an outlet temperature of 750?C the current 015 servicing of the upper part of the reactor, the recharging of the fuel elements, the functioning of the control and safety-rod system, and the regulation of the gas flow along the channels would be very difficult. One interesting feature is the transition to a new, compact construc- tion of the direct-flow boiler in the Fulton station. The power-station unit provides for six boilers, each producing ?600 tons/h of steam. The total boiler heating surface is therefore very large. The over-all dimensions nevertheless have to be limited, since the whole equipment of the circulating circuit is made in integrated form in a common prestressed-concrete housing. According to design the boiler will therefore be made from coils wound around a cylindrical surface. In the concrete reactor vessel, this fairly powerful boiler is to be sited in a cylindrical cavity only 3.2 m in diameter. It would thus be very difficult to repair the boiler if any leaks developed. In the majority of cases it would be necessary to extract the whole boiler and spend a great deal of time on servicing. It is therefore vital that the boiler should be made as reliable as possible. The question of radiation hazard is especially important, since the cooling gas of these reactors contains fission fragments diffusing through the cans of the fuel granules. This complicates operations in the recharging of the fuel, and also repair operations within the circuit. Problems of general safety are here approximately the same as in the AGR reactors. There is also a great deal of graphite in the active zone of these reactors. Hence if the gas circulation is accidently cut off the residual heat will pass into the graphite stack without damaging the granule cans for some considerable time. The systems are also provided with special emergency cooling systems. Fast helium-cooled reactors with a prestressed reinforced concrete vessel are also planned. These also envisage integrated construction, i.e., all the equipment resides inside the reinforced concrete ves- sel. Owing to the absence of a moderator the active-zone dimensions of a fast reactor are much smaller. In the design for the fast helium reactor GBR-4 with an electrical power of 1200 MW [14] the diameter and height of the active zone are 4 and 1.4 m. The fuel elements are cylindrical, the can is of stainless steel 7.7 mm in external diameter. As indicated in Fig. 4, the outer surface of the fuel-element can has a fine rectangular thread to intensify heat transfer. Such fuel elements are assembled in a hexagonal cassette with a gage dimension of 213 mm and are arranged in a triangular lattice with a step of 11.65 mm, i.e., with a gap of about 4 mm. Altogether the cassette contains 321 fuel elements. It is well known that the main characteristic of a fast reactor is its extremely high thermal loadings. In the GBR-4 reactor the maximum loads are about 400 MW/ton of U or 1.5 -106 kcal/m2-h of the surface of the fuel element. In view of this the problem of cooling the active zone in the case of an accidental dis- connection of the gas blowers is extremely vital. In the reactor design this problem is intended to be solved by means of three loops capable of operating with natural circulation of cooling water and cooled gas. When an emergency stoppage occurs lathe gas blower, the reactor is protected by continued circula- tion of the gas through the active zones, promoted by the difference between the densities of the gas in the active zone and the sets of tubes forming the emergency loop. For the normal gas pressure in the circuit (90 atm) this gas circulation is sufficient to prevent melting of the active zone, at least for a while, so giving time to restore the gas circulationby means of the gas blowers [14]. In addition to this, these emer- gency loops have their own gas blowers, driven by emergency Diesel generators, which may be automati- cally brought into action at short notice. With this arrangement the chief danger clearly arises in the very first moments, in which the residual heat has to be carried away by natural gas circulation. Another original feature of the design is the fact that the lower part of the fuel-element assembly is filled with activated charcoal, designed to absorb gaseous fission fragments. In addition to this, in a fast gas-cooled reactor it is desirable to use fuel elements made from granules with a graphite can (Fig. 3). The search for other coolants (apart from those already known) to be used in fast reactors has so far met with little success. As an alternative it has been proposed to use dissociating nitrogen tetroxide Fig. 4. Longitudinal section of the fuel-ele- ment can of a fast helium-cooled reactor. 7 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 N204 [15]. According to A. K. Krasin et al., a reactor with a coolant of this kind may provide a doubling time equivalent to that of sodium reactors. Thus it may now confidently be stated that gas-cooled reactors will shortly be capable of effective use not only in nuclear power stations but also as sources of high-temperature heating for technological processes in industry. The development of the latter aspect has only just started, and cannot of course be very rapid, since it depends on both reactor development and special technologies. Nevertheless, it is becoming more and more evident that nuclear reactors are excellently suited to high-temperature gas heating. It should furthermore be noted that extremely heat-resistant materials (molybdenum, niobium, tantalum, and their alloys) may also find employment in future high-temperature reactors. It may well be that these new materials will be required for the heat exchangers, in which high-temperature helium from the reactor will heat the gas directly used in the technological process. Finally it is appropriate to note that, in the possible development of the technological aspect, there is yet another specific limitation which has to be taken into account; namely, that arising from radioactiv- ity; it is accordingly essential to ensure that the practical exploitation of the nuclear reactor should be characterized by great efficiency and reliability. It is reasonable to assert that there should not be too many technological undertakings with nuclear reactors, and the latter should therefore have a very high unit power, or a correspondingly high thermal power, at least of the order of 1000 MW. The technological undertakings should be correspondingly powerful. LITERATURE CITED 1. A. P. Aleksandrov, At. nerg., 25, No. 5, 356 (1968). 2. Brit. Nucl. Export Executive Rev., No. 1 (1966); No. 2 (1967); No. 3 (1968). 3. R. Kutter, Fourth Intern. Conf., Geneva (1971), Rep. 49/p./468. 4. Nucl. Engng. Intern., 20, No. 229, 412 (1975). 5. Nucl. Engng. Intern., 19, No. 219, 623 (1974). 6. Power and Research Reactors in Member States, IAEA, Vienna (1974). 7. Appl. Atomics, No. 857 (1972); No. 936 (1973). 8. Nucl. Engng. Intern., 19, No. 220, 689 (1974). 9. Electr. Rev., Feb.8 (1974); Electr. Rev., Oct. 25 (1974). 10. J. Nucl. Energy, 26, No. 1, 49 (1972). 11. Brit. Nucl. Export Executive Rev., No. 3, 75 (1968). 12. S. M. Feinberg, At. Energ., 37, No. 1, p. 3. 13. V. Boyer and J. Gibbons, Nucl. Engng. Intern., 19, No. 219, 635 (1974). 14. Nucl. Engng. Intern., 19, No. 218, 566 (1974). 15. A. K. Krasin et al., 4th Geneva Conf., Soviet Paper No. 431 (1971). Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 EXPERIENCE IN THE USE OF A NUCLEAR REACTOR IN THE NORIL'SK MINING-METALLURGICAL COMPLEX V. N. Nikitin, V. N. Pavlova, UDC 621.039 A. I. Petrov, and A. M. Shchetinin The diverse analytical problem in the practical training of the Norilisk mining-metallurgical complex requires constant development, modernization, and unrestricted industrial application of the most modern chemical, physicochemical and physical methods for investigating the elementary composition of the nat- ural raw material and the products of its technological processing. The correct technology for process- ..ng the raw material and for the extraction of metals depends on the speed and quality of the analysis, and the detection and elimination of channels of loss of industrially valuable elements depends on the sen- sitivity of the methods [1]. Among the physicochemical and physical methods which have become classical, activation analysis possesses a number of additional potentialities with respect to sensitivity, rapidity and efficiency. There- fore, an activation analysis laboratory was built in the complex with an RG-1M research nuclear reactor [2,3], which in the first place should solve the problems of the high-sensitivity determination of the con- tent of metals of the platinum group and the rare elements, should ensure a higher efficiency for the analysis of geological and technological materials in nonferrous and rock-forming elements and should carry out radioisotopic investigations of technological processes. The thermal-power design of the reactor (5 kW) was increased during the start-up adjustment period to 30 kW [4, 5] and in April 1970 the RG-1M reactor was brought on stream. As a result of redesign after two years of operation and after the investigations carried out, simultaneously with the installation of a five-channel pneumatic rabbit system a further increase of its power was achieved [6]. The thermal neu- tron flux at the center of the reactor core is now 2.7.1012 n/ (cm2. sec) and, by using a neutron trap, it is doubled. The heterogeneous RG-1M reactor, of the swimming-pool type, has 11 vertical experimental chan- nels in ten of which the thermal neutron flux amounts to 0.4 to 2.7.1012 n/ (cm2. sec) and in one channel the fast neutron flux (energies greater than 5 MeV) has a value of 108 n/ (cm2 ? sec). Five channels are equipped with a pneumatic rabbit system for transporting samples into the reactor and out of the reactor to the store position after irradiation or measurement. The times of irradiation, holding and address for con- veyance of the samples in ampoules are provided automatically. The pneumatic rabbit equipment, de- veloped and manufactured in the All-Union Scientific-Research Institute of Reactor Technology, has shown high operating qualities during two years. The magnitude of the neutron flux in the reactor channels and the pneumatic rabbit equipment permit neutron activation analysis to be carried out with a sensitivity and accuracy, satisfactory for industrial requirements, in two alternatives: radiochemically with the use of long- and average-life isotope-tracers and nondestructively with the use predominantly of short-lived activities. Single-shift operation of the reactor was found to be sufficient for the irradiation of samples being analyzed according to a manufacturing scheme. The activation analysis laboratory has six work points for radiochemical treatment of irradiated samples, in combination with two scintillation spectrometers based on multichannel pulse analyzers AI- 128 and USD-1 universal scintillation data units, and two industrial facilities for instrumental activation analysis, including NTA-512 and LP-4840 multichannel pulse analyzers with a bank of scintillation and Translated from Atomnaya Energiya, Vol. 40, No. 1, pp. 11-16, January, 1976. Original article submitted March 3, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 9 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 TABLE 1. Characteristics of Radioactiva- tion Analysis Procedures Element determined Thresh. of sen- sitiv. for 5-h ir- ad., 10 No. o Wt. of sam- sample pies to beak in sin alyzed, g gle batch Time of analysis of single batch, h Sample output per week Radiochemical version Au 10-7 0,5--1,0 18 18 36 Pt 5.10-5 0,2--0,5 6 18 12 Pd 2.10-6 0,2--0,5 6 18 12 Jr 2-10-7 0,5--1,0 20 36 20 Ru 2.10-6 0,5--1,0 20 36 20 Os 5.l0- 0,5--1,0 20 36 20 Re 10-6 0,1--1,0 24 36 24 Ag 10-4 0,1--1,0 30 36 30 Instrumental version Cu (in core samples) 10-2 up to 200 1-2 0,15 up to 200 Co(6?Co) 2-10-5 1-2 40-50 168-336 200-250 Co("mC0) 10-3 0,2-1 2 0,18 200 Si 10-4 0,5-1 2 0,18 200 Al 10-2 1-2 1-2 0,15 240 Au 10-4 1-5 40-50 72-120 200-250 Ir 10-5 1-5 40-50 168-240 200-250 semiconductor Ge (Li) detectors. In addition to this, the laboratory is equipped with an AI-4096-3M amplitude?time analyzer, an "Angara" coincidence spectrometer, an MIR-1 electronic computer, and a satisfactory amount of electronic ?physical, radiometric, and dosimetric equipment. Considerable assistance in equipping the laboratory with modern equipment and plant was afforded by the com- bine of the organization GKAE (State Committee for the Utilization of Atomic Energy) of the USSR, treating the acti- 2 7 ? vation analysis laboratory of the Noril'sk Mining-Metallurgi- cal Complex as an experimental-industrial proving ground for the application of achievements in the field of applied nuclear physics to the national economy. Many instruments 19 74 and methods, developed in the All-Union Scientific-Research Institute for Reactor Technology (VNIIRT), are introduced and used in the Noril'sk Mining-Metallurgical Complex, and joint scientific-research and systematic work are undertaken. About 20 procedures have been developed and intro- duced into analytical practice in the laboratory during a year of experimental-industrial operation (industrial since 1972). The characteristics of the continuously used procedures are given in Table 1. When solving ana- lytical problems by radioactivation analysis methods, the requirement for high sensitivity was taken into account (especially when determining the content of metals of the platinum group and of the rare elements), which was lacking in the methods previously used. Radioactivation analysis provides better characteristics of productivity and rapidity and replaces the more laborious classical methods of analysis. As a result of this, the accuracy of the analysis is increased, especially for industrial products, by which a technological balance is set up. Only the use of this method makes it possible to determine the content of elements in samples of very small weight, in unique samples or in undisintegrated samples, when the use of other methods is impossible; moreover, the list of elements being determined is supplemented. The advantages of activation analysis in sensitivity and productivity in comparison with chemical methods have permitted two fields of its industrial application to be defined; for determining the content of microquantities of elements of the platinum group and the rare elements in so-called low-grade pro- ducts [7,8] and for the bulk analysis of geological specimens and products of the average monthly techno- logical assay on the content of ferrous metals and certain elements of the silicate group. Thus, the 1970 1971 1972 1973 Fig. 1. Radioactivation analyses by means of the RG-1M reactor (thousands of element-determinations). 0) Radio- isotope analysis; ?) instrumental acti- vation analysis. 10 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Fig. 2. The RG-1M reactor building of the Noril'sk complex. TABLE 2. Economical Efficiency of Exper- iments Carried Out on the RG-1M Nuclear Reactor, Thousands of Rubles Type of expts. 1970 1971 1972 1973 1974 Radiochemical analyses 2,0 4,8 31,9 33,0 39,0 Instrumental activation analyses -- 2,8 4,5 6,5 18,0 Reduction of sample ir- -- -- 36,0 48,0 - 67,0 introduction of procedures for the radiochemical deter- mination of the Ir, Os, Ru, and Re content has signifi- cantly supplemented the capabilities for the analytical service of the complex, and the use of instrumental activation analysis of ores and industrial products for the content of cobalt and aluminum has increased sig- nificantly the productivity of the analytical operations, has increased their quality, ,andhas shortened the times. The determination of the microelemental composi- radiation cost tion of the environmental contamination, the analysis of dust samples and of certain industrial solutions, the Total 2,0 7,6 72,4 87,5 124,0 nondestructive analysis of large bulk-samples, etc., il- lustrate the merits of this method. Problems of deter- mining the degree of contamination of the environment, the effects of copper? nickel production waste on the plant and animal world of Zapolyar'ya atthe present time are acquiring particular importance in connection with the increase of the volumes of industrial output. The capabilities of activation analysis are of great importance in these investigations. A series of analyses of the snow cover, selected in the Noril'sk region, carried out by the method of nondestructive multiele- mental analysis using Ge (Li) detectors with a volume of 60 cm3 and a resolution of 3 to 4 keV, permitted the content of more than 20 elements to be estimated quantitatively, including Hg, As, Se, Te, Go, NI, Cu, Fe, Cr, etc. The instrumental determination of the content of native copper in core samples with a weight of up to 200 g permitted the error, due to the nonrepresentativeness of the balance used in chemical and x-ray spectral methods, and losses of metal on grinding the samples to be reduced. The use of both versions of the neutron-activation method in analytical practice of the Norirsk com- plex is shown by the histogram (see Fig. 1) and at the present time an increase is being restrained only by the shortage of laboratory accommodation. By means of radioactive isotopes, it has become possible to use successfully the two versions of radioisotopic investigations: the introduction of labelled materials Into a production process with the subsequent removal of technological samples for measurements under laboratory conditions and radiometric measurements directly on industrial plants. The first method was used to investigate the depletion process of converter slags and the behavior of the noble metals during the conversion of nickel-containing white matte, when the compounds being studied, 11 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Fig. 3. The reactor hall. Fig. 4. Reactor control desk. labelled with short-lived radioactive isotopes of Cu, Co, Ni, Au, Pt, Pd, Os, Se, and other elements, were introduced into materials being processed in the industrial plants. Analysis of selected samples assisted in drawing the appropriate conclusions concerning the distribution of these metals in the metal- lurgical products [9]. In the second case, direct examination was accomplished without sampling of the calcination of nickelous oxide in rotating tubulai furances and recommendations have been made for the optimization of this technological process. Similar investigations are being planned for the future. The first stage for the introduction into industry of applied methods of nuclear physics and equip- ment, combined with overcoming difficulties of an organizational-technical and psychological nature, can be assumed to have been completed successfully. In making an attempt to assess the economical efficiency of the all-round utilization of a nuclear reactor (Table 2), it is still not possible to talk about the consider- able achievements which, to a known degree, are explained by the insufficiently complete utilization of all its capabilities and the mass use of activation analysis first of all because of the absence of an industrial- experimental basis. The fraction of activation analyses (about 20 thousand elemental-determinations per year) in comparison with the total volume of the annual analytical operations of the complex (more than 2 million element-determinations) amounts to about 1%; even among similar operations carried out by other ? methods, this contribution still amounts to 5-6%. Clearly, tendencies are seen to increase the quantity and efficiency of the operations carried out by means of a reactor, to approach self-repayment of the annual costs on the content of the laboratory which, according to our forecasts, will be reached in 1976. Over 5 years, more than 36 thousand element-determinations have been carried out. A further increase ? of the volume of activation analyses by a factor of ten may even give an annual economy of about 0.5 million rubles. A new stage of development of activation analysis and radioisotpe investigations at the Norirsk Mining- Metallurgical Complex using the RG-1M nuclear reactor provides for a widening of the circle of industrial materials to be analyzed (including solutions) and an increase of the number of elements to be determined; the mass use of the procedures introduced; the development partially or completely of automated analytical cycles with data processing on a computer and, in the first place, of the multielement instrumental version of analysis; the application of activation analysis for investigating plant samples, animal tissues and other biological items; the carrying out of analyses for ecological investigations; the expansion of radioisotope investigations of technological processes for their optimization and the reduction of the loss of valuable metals, etc. [101. Five years of industrial operation of the research nuclear reactor at one of the largest-scale mining- metallurgical plants of the country and the experience built up permit a positive conclusion to be drawn concerning the prospects for further extension of operations by means of nuclear-physical methods of in- vestigation in-the application to problems of nonferrous metallurgy and of the mining industry, and give a 12 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 basis for optimistic conclusions concerning their increasing industrial-practical importance. LITERATURE CITED 1. B. I. Kolesnikov, in: The State of Technology for the Extraction and Analysis of the Content of Metals of the Platinum Group in the Process for the Concentration of Copper ? Nickel Ores [in Russian], Izd. TsNIIinformtsvetmet, Moscow (1967), p. 3. 2. Yu. M. Bulkin et al., At. Energ., 21, No. 4, 319 (1966). 3. A. S. Shtani, At. Energ., 33, No. 4, 858 (1972). 4. V. I. Alekseev et al., At. Energ., 32, No. 4, 315 (1972). 5. A. M. Benevolenskii, V. T. Shentsev, and A. M. Shchetinin, in: Collection of Scientific Proceed- ings of NVII No. 15. Physicotechnical Issue [in Russian], Izd. Krasnoyarsk Polytechnical Institute, Krasnoyarsk (1973), p. 5. 6. A. M. Shchetinin et al., At. Energ., 38, No. 2, 97 (1975). 7. V. N. Pavlova et al., Zh. Analit. Khim., 29, No. 11, 2088 (1974). 8. V. P. Razhdaev and V. N. Nikitin, Zh. Analit. Khim., 29, No. it, 2172 (1974). 9. The Use of Isotopes and Ionizing Radiations in the National Economy of the Urals. Data for Reports at the 3rd Zonal Conference on the Use of Isotopes in the National Economy of the Urals (Thesis of Reports) [in Russian], Sverdlovsk (1973). 10. V. N. Nikitin, in: The 9th All-Union Conference on the Chemistry, Analysis, and Technology of the Noble Metals (Thesis of Reports) [in Russian], Krasnoyarsk (1973), p. 183. 13 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 TESTING OF EXPERIMENTAL BN-600-TYPE FUEL ELEMENTS IN THE BOR-60 REACTOR UP TO DIFFERENT BURNUPS M. M. Antipina, Yu. K. Bibilashvili, I. S. Golovnin, V. M. Gryazev, E. F. Dyvydov, G. V. Kalashnik, A. V. Medvedev, T. S. Men'shikova, V. S. Mukhin, A. A. Petukhov, A. V. Sukhikh, V. N. Syuzev, L. I. Sytov, and V. L. Timchenko UDC 621.039.542.342:621.039.548 Tests in the BOR-60 of experimental fuel elements ? prototype fuel elements for high-capacity power reactors ? complete the work on the creation of a fuel-element design. These tests and post-ir- radiation investigations permit the operating qualities of the design to be analyzed, the assumed design and technological solutions to be refined or confirmed, and also the route for further optimization of the fuel element to be designated. In this present paper, the generalized results are given of an investigation of three experimental bundles, with fuel elements based on oxide fuel, which were irradiated in the BOR-60 reactor to burnups of 4.3, 8, and 10.3% of heavy atoms. The principal structural and thermal parameters of the fuel elements did not differ from the working parameters of the BN-600 fuel elements. However, the neutron fluence for the claddings of the experi- mental fuel elements are lower than is required for the BN-600 fuel elements (-7.5.1022 instead of -3.5. 1023 n/cm2). This difference is quite significant and must be taken into account in the analysis of the re- sults of the tests. All the experimental fuel elements achieved the stated burnup without damage to the completeness of the cladding and loss of hermiticity. The results of the post-radiation investigations permitted the degree and nature of the chemical in- teraction of the cladding and the fuel with the fission fragments to be estimated, the magnitude of the de- formation stored up after the run as a result of the mechanical action of the fuel core and swelling of the steel, and it permitted the gas release from the fuel to be measured and the structural changes in the core ? to be determined. Structure and Manufacturing Technology of the Fuel Elements. In the principal structural character- istics (diameter and thickness of the cladding, pitch of positioning in the bundle, effective fuel density and density of the sintered pellets, the cladding-fuel gap), the experimental fuel elements irradiated in the BOR-60 reactor are similar to the regular fuel elements of the BN-600 reactor [1]. The structure of the experimental fuel element is shown in Fig. 1. The outside diameter of the cladding, ofOKH16N15M3B steel from electroslag remelting, is 6.9 mm and the wall thickness is 0.4 mm. In the single cladding are located the active section and the upper and lower reflectors. The length of the core of the active section, in the form of sintered sleeved pellets of 90% enriched uranium dioxide, is 500 mm. The density of the pellets is 10-10.6 g/cm3. The nominal value of the effective fuel density in the fuel element is 8.25 g/cm3. The upper and lower reflectors, with a length of 50 m, are made from de- pleted uranium dioxide with a density of not less than 10 g/cm3 and directly touch the core of the active section. Translated from Atomnaya Energiya, Vol. 40, No. 1, pp. 16-27, January, 1976. Original article submitted April 28, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 14 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Deformation Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 2 50 500 A 50 1100 7 421 0178 A?A 44 B?B Fig. 1. Structure of an experimental fuel element of the BN-600 reactor; 1) upper cap; 2) spring; 3) reflector; 4) spacer band; 5) active section of fuel element; 6) sleeve; 7) fuel element cladding; 8) lower cap. 2000 4000 6000 8000 1000 1200 Time, h Fig. 2. Nature of change of mechanical deformation over the run in the average cross section of the fuel element (375 mm from the lower end of the cap) on the time of operation at nominal power when y0 = 10.3 g/cm3; 1,2,3) fuel elements of Type I and II (see [2]) zones of high enrichment and zone of low enrichment, respectively. The pellets are prepared by a cold pressing method, using a rotary automatic press, and with subsequent sin- tering at a temperature of 1650?C during 3 h in metal heater furnaces. Before loading the pellets in the fuel element cladding, chemical, x-ray and metallographic analysis was carried out in order to determine the con- tent of uranium, oxygen and other admixtures. The char- acteristics of the pellets are as follows; the content of fluorine, carbon and other admixtures to a total of 1100?K. These investigations were prompted by the small number of published data, their large discrepan- cies, and the narrowness of their temperature range. The thermal diffusivity was measured by the method of radial temperature waves based on one variety of regular thermal field of the'third order. The thermal diffusivity can be determined either by means of the phase of the first harmonic of the temperature oscillations, or by means of the difference between the times corresponding to the maximum temperature and the moment of switching on the periodic electronic heating of the specimen [2]. We investigated yttrium of 99.8% purity and gadolinium of 99.75% purity, in the form of hollow cy- linders 7 cm long with external diameters of 15 mm and internal diameters of 6 mm. The impurity con- tent of the yttrium was as follows (in percent): Gd, Tb, Dy, Ho < 0.1; Fe < 0.01; Ca < 0.03; Cu < 0.05; Ta 0.06; that of the gadolinium was: Y ? 0.08; Tb 0.07; Eu 0,04; Cu ? 0.025; Fe ?-? 0.02; Ca < 0.004. The systematic error of the resits on the thermal diffusivity was about 5%. Data on the dc electrical resistivity P were obtained by the four-probe method. We used heating by electron bombardment of the internal surface of the hollow cylindrical specimens (external diameter 15 mm, internal 6 mm, length 7 cm). We also measured p by heating in a resistance furnace on solid cylin- ders, 6 mm in diameter and 7 cm long [3]. The systematic error of the results on p was about 2% for the solid phase and about 3% for the liquid. Figure 1 plots the results of measurements of the thermal diffusivity of yttrium and gadolinium be- tween 1100 and 1700?K. The rms deviation of the individual points from the smoothed values is about 3%. The results for Gd are close to those in [4]. The discrepancy is about 6% at T 1400?K or higher, i.e., close to the total systematic error. TABLE 1. High-Temperature Thermal Con- ductivities of Y and Gd [A ? 102 W/(m ? deg K)] Element T,*IC Gd 1100 1300 1500 1700 0,17 0,18 0,19 0,20 0,17 0,19 0,21 0,20* Estimated from data for p on the basis of the Wiedemann?Franz law. Translated from Atomnaya fnergiya, Vol. 40, No. 1, pp. 63-64, January, 1976. Original article submitted March 31, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 69 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 o 012 ;;-? 0,1f 0710 0,09 1000 1200 1400 1600 T, ?K 210 ? 200? a 180 I 1000 1200 1400 1600 T, ?K Fig.1 Fig. 2 Fig. 1. High-temperature thermal diffusivity of Y and Gd. 0) Yttrium; 0) gadolin- ium. Fig. 2. High-temperature electrical resistivity of Gd. A) Data from [4]; ?) data from [5]; 0) present authors' data for solid cylinders; 0) for hollow cylinders. The results of measurements of the electrical resistivity of gadolinium are plotted in Fig. 2. Be- tween 1100 and 1500?K they are comparatively close to the data in [4, 5]. At the phase transition point Ta ?i3= 1537?K [5], the structure of gadolinium changes from hcp to bcc, but there is little change in the electrical resistivity. According to Dennison et al. [6] there is a marked change in the specific heat. For the 13 phase, p - 200 p.,2 ? cm. The random errors are rather larger than for the a phase at T 1500?K, so that one could only conclude that there is a singularity in the behavior of p in this region after a marked improvement in the accuracy. We can note a deviation from linearity in pa) which is typical of the lanthanides [3]. The electrical resistivity of liquid gadolinium is about 205 ?2 ? cm, and between the meltingpoint Tmp and 1800?Kit is independent of the temperature, pi /ps - 1.04. A slight change of p on melting has been ob- served [7] for transition metals. The data in [1] on the volumetric specific heat of gadolinium and yttrium, together with our results on the thermal diffusivity, were used to obtain values for the thermal conductivities of these elements (Table 1). The published information on the thermal conductivities of y and Gd at about 1200?K or higher is very limited; we know only the results in [4, 8] on the thermal conductivities of Y and Gd at 900-1500?K, and the data given in [9] concerning the thermal conductivity of Y at 350-1150?K. It is difficult to compare these results, because the thermal conductivity of Y depends on the hydrogen content [9]. The thermal conductivites of Y and Ga increase with rise of temperature. The data obtained for P and A were used to estimate the Lorenz numbers L. The results for L are higher than the theoretical values, possibly owing to the phonon contribution to the thermal conductivity. A high value of this compo- nent is one feature of the lanthanides. Using the data in [10] between 90 and 310?K and in [4] between 900 and 1400?K, as a result of our experiments we can infer that there is a monotonic rise in the thermal con- ductivity of gadolinium in the paramagnetic region (270-1500?K). Similar inferences were drawn for the specific heat and electrical resistivity. Similar behavior of the thermal properties was noted for yttrium [1,8]. Our discussion of the data on p was based on the assumption that elastic scattering processes play a leading role above the Curie and Wel points (and the Debye temperature), and that Matthiessen's rule applies. The observed behavior of p in the paramagnetic region is largely governed by scattering of elec- trons by disordered spins [11]. Analysis of the thermal conductivity is based on the assumption that the Wiedemann? Franz law holds and that the thermal conductivity can be expressed in the form of a sum of the phonon and electron components. This approach to the transition elements meets objections in [12]. We must also reckon with the existence of other contributions to the thermal conductvity [10], in- cluding that of magnetic ordering in the high-temperature region. A complete examination of the behavior of the properties of the lanthanide elements is made difficult by the lack of complete information on the structures and energy spectra of these elements. 70 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 LITERATURE CITED 1. I. I. Novikov and I. P. Mardykin, At. nerg., 37, No. 4, 348 (1974). 2. L. P. Filippov, Investigation of the Thermal Properties of Solid and Liquid Metals at High Tempera- tures [in Russian], Izd. MGU (1967). 3. I. P. Mardykin, Teplofiz. Vys. Temp., 13, No. 1, 211 (1975). 4. V. E. Zinov'ev et al., Fiz. Tverd. Tela, 14, 2747 (1972). 5. F. Spedding, J. Hanak, and A. Daane, J. Less.Comm.Met., 3, 110 (1961). 6. D. Dennison, K. Gschneidner, and A.Daane, J. Phys. Chem., 44, 4273 (1966). 7. A. R. liege'', in: The Structures and Properties of Liquid Metals [in Russian], Izd-vo Akad. Nauk SSSR, 3 (1959). 8. V. E. Zinov'ev and P. V. Gel'd, Fiz. Tverd. Tel., 13, 2261 (1971)., 9. Y. S. Touloukian (editor), Thermophysical Properties of High Temperature Materials, Vol. 1, Macmillan, New York? London (1967). 10. D. Chuah and R. Ratnalingam, J. Low Temper. Phys., 14, 257 (1974). 11. T. Kasuya, Progr. Theor. Phys., 16, 45 (1956). 12. M. Laubitz, High Temper. ?High Press., 4, 379 (1972). Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 71 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 EFFECT OF IMPLANTED SPACE CHARGE ON PARTICLE RANGE DISTRIBUTION V. S. Remizovich and A. I. Rudenko UDC 539.124.17 If the thickness of a layer of matter is sufficiently great, heavy charged particles (ions, protons) are decelerated because of interactions with atoms of the material and then are stopped, forming a dis- tributed space charge. The amount of this charge increases as the radiation time increases and the in- tensity of the electric field created by the implanted charge can reach significant values (of the order of 106 V/cm) [1]. There is evidence [2, 3] that the macroscopic electrostatic field can have a significant effect both on the penetration of charged particles into matter and on the mechanical properties of the material itself. At the same time, because of the presently important problem of ion implantation in materials, there is interest in a calculation of the depth distribution of the implanted particles for various irradiation times. In this paper, a solution of the transport equation is obtained for heavy charged particles including the effect of the self-consistent field of the space charge produced in the target in the case of plane geo- metry. Let a broad beam of monoenergetic, nonrelativistic particles having a velocity v? directed along the normal to the surface of a plane-parallel plate (along the x axis) be incident on this plate, which is made of a nonmetallic homogeneous material, starting at the time t = 0. For heavy particles, one can neglect velocity deviations from the original direction and fluctuations in energy loss during deceleration [4]. The transport equation for the distribution function is then written in the form Of (x, v, t) Of F(x, t) Of 1 8 - Ox s OS M as [8(v) f (x, v, t)] = 0, Ot 7n (1) where 7(v) is the average energy lost per unit path length by a particle with velocity v; m is the mass of a particle; Fx is the projection of the force acting on a moving particle. In a number of cases the speci- fic energy loss 7 depends very slightly on particle velocity so that in first approximation it can be as- sumed constant ? Fo, where F 0 = mv20R-01/2 (here, R0 is the range of particles with an initial velocity v? in the absence of space charge). The boundary condition for Eq. (1) takes the form 1(0, v, vo (2) where 10 is the incident particle flux density. The time from entrance to a complete stop is to = mv0(e0)-1 in the absence of electrostatic interac- tions between particles. In actual cases, to 10-16-10-13 sec. The electrostatic force produced by the presence of an implanted space charge decreases the stopping time. Therefore the total number of par- ticles (per unit surface area) moving in the material at any time is no greater than 40 = Ienv0(60)-1. The moving particles create a field which acts on a particle entering the medium with a force not greater than F = 271-I0(ze)2mv0(xe0)-1 (here, ze is the charge of the incident particles and x is the dielectric constant of the material). The magnitude of this force even for the most intense continuous sources is many orders of magnitude less than the stopping power of the medium, i.e., F /0 (ze)2 nivo a ? ? 2n < 1. E0 (3) Translated from Atomnaya nergiya, Vol. 40, No. 1, pp. 64-66, January, 1976. Original article submitted August 7, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, NY. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 72 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 This circumstance makes it possible to neglect the interaction between moving particles a.nd to assume that the decelerating electric field is created only by the stopped particles, i.e., by the implanted charge. It is impossible to neglect the contribution of the latter to the field since their number increases with time. Since the mobility of the stopped particles in a dielectric is very small [5], one can neglect a slight displacement of these particles after stopping because of the forces of electrostatic repulsion. Then the force Fx in Eq. (1) describes only the effect of implanted charge on the moving particles. The presence of a decelerating electric field leads to the fact that particles entering the material at later times experience greater deceleration than particles entering earlier. Therefore the total stopping range R (t) decreases as t increases. Thus at the time t the stopped particles are distributed with a cer- tain density over the range of depths R (t) Ls x S Ro and the moving particles are found in the region 0 < x < R (t). If the time dependence R (t) of the range is known, the distribution of the implanted charge, p (x, t), can be calculated in the following manner: in the time interval from t to t+dt, dN = Iodt particles enter the material which stop in the layer dx = ?(dR/dt)dt at the depth x = R (t). Therefore I+ , dl I \-1 " k jt(x) R (t) < x < Ro; 0; x > to. Therefore in the time range t>> to of interest to us, a(t) lot, since the stopped particles outnumber the moving particles. As a result, Fx is written in the form (4) Fx= ?2n (ze)2 The solution of Eq. (1) with Eqs. (2) and (5) taken into consideration takes the form where 2/0 2a f (E, u, uo TO ? 2a (1?u) ; u=_!_; = ?_L; (t) =T.? 1-F accr [1 (i )2] (1 + CCT)2 (5) (6) (7) and the parameter a is defined by Eq. (3). In the limiting case = 0), we obtain from Eq. (6) an expres- sion for the distribution function fo (E, u, T) =210-2-- (1?u) 0 (x-1- u ? 1) 8 (1? u2 (8) vo Equation (8) represents a solution of the transport equation (1) in the absence of the decelerating force Fx. Since a 1. Setting u = 0 in Eq. (9) we find the time dependence of particle range: Ro Ro Ro R (0? 1+6,T f+c, to ? i? ? (9) (10) The particle range decreases as the time increases. We find R = (1/2)R0 when t = t* = to/a. Now using Eqs. (4) and (10), we find the density of implanted charge: Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 73 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 mqx , R0 > Es.) Figure 4 shows curves for < 02+(E-, E+, T)>, the mean square spatial angle of positron emergence, as a function of E+ and T. Within the limit of statistical error, < 02+(E_, E+, T) > can be considered independent of E_ and also Gaussian, as shown by analysis of the an- gular distributions. In order to judge the reliability of the results obtained, the Monte Carlo calculations were compared with experimental data [1-4]. Spectra for d2N+/dE4d2 from [1,3, 5] are compared in Fig. 5 with the spectrum obtained for E- = 25 MeV and T = 0.2 radiation lengths. The spectrum from [1] was increased by a factor 2.7 in accordance with the note at the end of that paper; points on the curve were obtained by interpolation of data for T = 0.15 and 0.3 radiation lengths. The spectra from [1] and [5] and the Monte Carlo calculation are in good agreement but are roughly a factor of two lower than the results of [3]. However, it was pointed out in [3] that the absolute scale of the experimental curves may contain a factor of the order of two. On the whole, the comparison of the Monte Carlo calculations with experi- mental data [1-4] showed that the agreement between them is good. In conclusion, the author takes great pleasure in thanking B. V. Chirkov, B. I. Grishanov, R. A. Salimov, and A. D. Bukin for valuable discussions. (1) LITERATURE CITED 1. M. Bernardini et al., CEA Report N 2212 (1962). 2. C. Jupiter et al., Phys. Rev., 121, 866 (1961). 3. L. Katz and K. Lokan, Nucl. Instrum. and Methods, 7, 7 (1961). 4. T. Aggson and L. Burnod, ORSAY, Preprint LAL-27 (1962). 5. V. Jacobs et al., Nucl. Instrum. and Methods, 61, 166 (1968). 6. R. Sund and R. Walton, Nucl. Instrum. and Methods, 27, 109 (1964). 7. V. A. Tayurskii and B. V. Chirikov, Preprint 73-73, IYaF, SO Akad. Nauk SSSR (1973). 8. F. M. Izrailev et al., Preprint 63-73, IYaF, SO Akad. Nauk SSSR (1973). 9. S. Z. Belen'kii, Cascade Processes in Cosmic Rays [in Russian], Gostekhizdat, Moscow (1948). 10. W. Messel et al., Nucl. Physics, 39, 1 (1962). 11. H. Nagel, Zeits. fur Physik, 186, 319 (1965). 12. M. Tamura, Prog. of Theor. Phys., 34, 912 (1965). 13. M. Ya. Borkovskii and S. P. Kruglov, Yad. Fiz., 16, 349 (1972). 82 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 DEPENDENCE OF ASYMMETRY IN TH.E PHOTOFISSION OF 2 3 3 U AND 239Pu ON THE MAXIMUM BREMSSTRAHLUNG M. Ya. Kondrat'ko, V. N. Korinets, UDC 539.173.3 and K. A. Petrzhak The fission yields of certain products of the symmetric and near-symmetric fission of 233U and 23213u by means of the bremsstrahlung from a betatron are determined over a range of energy maxima E0 = 10- 24 MeV. The procedure for similar experiments is described in [1-5]; irradiation of targets using special equipment, inserted in the accelerating chamber of the betatron; buildup of radioactive fission products in the form of nuclear recoil; radiochemical analysis, and measurements of 13 activity in proportional 4Tr counters. The fission products were identified by their half lives. Experimental decay-buildup curves were fitted by the method of least squares, utilizing tabular values for the half lives and branching ratios. The yields were determined with respect to the standard 149Ba. Cumulative yields of identified isotopes were scaled to the values of the total yields of branches with a corresponding mass number. In addition, numerical estimates of individual yields of daughter isotopes and published data on the branching of decay with the formation of short-lived isomers were utilized. The results are presented in Tables 1 and 2. The values of the relative yields, the error in which has not been noted, are determined with a relative error of 5-7%. The error in the maximum energy E0 is due mainly to a drift in the standard values during irradiation and consisted of approximately ?150 keV. The yield ratios of 139Ba and 149Ba, equal to 1.11 ? 0.06 and 1.13 d= 0.05 when E0 = 16-20 and 24 MeV, re- spectively, are also determined for the photofission of 239Pu. A comparison of the results shows that the ratios Y115Cd/Y149Ba for 233U and Y117Cd/Y149Ba for 239Pu, representing the case of the most symmetric fission, are similar in magnitude and energy dependence. A systematic reduction in the yields is observed for 235PU upon switching from near-symmetric (111Ag) to most-symmetric (vict') fission. However, in the photofission of 233U, the relative yields of 113Ag are less than with 115417Cd. It is possible that this irregularity is connected with the manifestation of a central peak in the mass distribution of the symmetric fission. A similar phenomenon has been observed in the photofission of 235U [5]. TABLE 1. Relative Yields of 233U Photo- TABLE 2. Relative Yields of 235Pu Photo- fission Products fission Products E0, MeV Y118Agnr140Ba Y115Cd/Y140Ba Y117 cd/Y140Ba 10 0,0031?0,0008 0,0078?0,0010 0,005?0,001 12 0,022?0,005 0,034?0,002 0,032?0,002 14 0,037 0,060 0,060 16 0,053 0,068 0,061 20 0,068 0,095 0,086 24 0,100 0,126 0,128 Eo, MeV Y111Ag/Y140Ba., Y113Ag/Y140Ba ,, 12 - . ?, . 10 0,046?0,005 0,020?0,006 0,013?0,002 0,010?0,002 12 0 ,075?0 ,005 0 ,031?0,003 0,027?0,002 0 ,024?0 ,003 14 0,100 0,058 0,059 0,049 16 0,144 0,081 0,073 0,071 20 0,176 0,099 0,102 0,094 24 0,202 0,121 0,127 0,127 Translated from Atomnaya Energiya, Vol. 40, No. 1, pp. 72-73, January, 1976. Original article submitted May 21, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 83 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Fig. 1. Yield ratios "peak to trough" for photofission. Data of present paper for the reaction 233U (Y f): 0) YttoBa/Y115Cd; ? ) Y 14oBa/Y113Ag; for the reaction 238Pu (y, f): ) A5(1234,43jid_. Ta....t)a n i40[B1a]/f;r h i it5ced ; reac- tionin[6] for the reaction 238U (y, f): ? ? ? ? ?') Yi4oBa/Y115Cd? 10 12 14 16 18 20 22 24 Eo, MeV ? The energy dependences of the "peak to trough" ratios are given in Fig. 1. The Y14?Ba/Y117Cd ratios for the photofission of 238U [1] and 238U [6] are given there also. A reduction in the "peak to trough" ratio 5? is observed in the series 233u 23-u-238u, i.e., with a given nuclear charge Z, the probability of the most symmetric fission is lower, the greater the mass number A. LITERATURE CITED 1. M. Ya. Kondrat'ko and K. A. Petrzhak, At. Energ., 23, No. 6, 559 (1967). 2. M. Ya. Kondrat'ko, 0. P. Nikotin, and K. A. Petrzhak, At. Energ., 27, No. 6, 544 (1969). 3. M. Ya. Kondrat'ko, 0. P. Nikotin, and K. A. Petrzhak, Pribory i Tekh. Eksperim., No. 3, 47 (1964). 4. M. Ya. Kondrat'ko, V. N. Korinets, and K. A. Petrzhak, At. Energ., 34, 52 (1973). 5. M. Ya. Kondrat'ko, V. N. Korinets, and K. A. Petrzhak, At. nerg., 35, No. 3, 214 (1973). 6. R. Duffield, R. Schmitt, and R. Sharp, in: Proceedings of the 2nd International Conference, Geneva, Vol. 15 (1958), p. 678. 84 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 SYNTHETIC PITCHBLENDE: COMPOSITION, STRUCTURE, AND CERTAIN PROPERTIES V. A. Alekseev and R. P. Rafaltskii UDC 542.65:549.514.8:549.12 Pitchblende with a highly defined collomorphic structure has been synthesized previously under hy- drothermal conditions by the reduction of hexavalent uranium by elementary arsenic [1]. However, inves- tigation of the products of the synthesis was not accompanied by systematic determinations of the 0/U ratio and water content. Additional experiments were conducted in order to obtain this information. A wafer of natural arsenic was placed in a quartz ampoule with a solution of UO2SO4, which was maintained at a fixed temperature after being evacuated and sealed. After rapid cooling, the ampoule was opened and we washed the wafer, ? coated with a thin layer of pitchblende, with water and alcohol and dried it in vacuum at room temperature. We carefully removed the thin layer; we used part of it for de- termining the content of tetravalent arid ordinary uranium and water, as well as for x-ray analysis. A polished section was prepared from the remainder. We calculated the 0/U ratio from the data of chemical analysis performed by the ferrophosphatevan- adate method [2]. The material subjected to analysis contained from 2.4-7.2 mass % of arsenic; however, it has been established that the presence of arsenic and its oxides do not have any effect on the accuracy of determining the uranium. We determined the water content by the Penfield method. During the heating of the sample, not only the water, but also arsenic trioxide, which together with the water was condensed in the form of a white deposit in an enlarged section of the tube, was driven off. During the drying of the latter in vacuum (110?C; 0.5-1 h), evaporation of As203 did not occur, a fact which was established by control weighing. In order to prepare the polished sections of the thin layer, we introduced an epoxy resin, after the solidification of which we produced the thin section and the polishing. Investigation of the polished sec- tions was followed by structural pickling in a 20% solution of iron chloride. We measured the reflecting TABLE 1 Expt. No. Exptl. conditions Composition and properties of the nth. pitchblende t, 'C heating ,. n me, tiuranium initial ,I concn. of in sol.. giliter volume of -sol. at 25?C, mli 0113 water content, To reflecting pow er, % abs. micro- hardness, kgf/ mm2 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 150 150 150 150 150 150 150 150 200 200 200 250 300 320 320 10,5 17,5 16,5 33 32 9 33 33 32,5 9 6 8,5 31,5 8,5 8,5 2 3 2 2 4 21,8 21,8 21,8 4,1 6 21,8 21,8 4,1 21,8 21,8 5 5 12 5 100 10 16 18 100 10 10 7 100 7 7 2,38 9,62 2,36 I 5,88 2,28 I 4,67 2,32 I 6,90 2,261 3.35 2,35 INot determined 2,36 I 2,33 I 6,27 2,25 2,66 2,19 I 2,33 2,19 I 1,84 2,21 0,96 2,12 I 0,12 2,161 0,76 2,09 0,79 Not determined D D 10,1 9,5 Not determined 10,5 12,4 13,3 11,9 Not determined 16,9 14,3 14,7 Not determined D D I 160 I 310 ot determined D I 185 I 410 I 420 I 320 jNotdeterrnined 650 i 450 505 Translated from Atomnaya nergiya, Vol. 40, No. 1, pp. 73-76, January, 1976. Original article submitted May 21, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y.10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 85 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 R,% 18 16 14 12 10 e o 200 H, kgf/mm2 400 600 Fig. 1. Dependence between reflec- ting power and microhardness of uranium oxides; 0) Synthetic pitch- blendes (microhardness is calcu- lated from optimal load); 0) the same, microhardness is calculated for 14-12 diagonal of replica; GO natural uraninites and pitchblendes. their reflecting power and microhardness between 0/U and CH20 (0.93), as well as between R and H (0.89 and 0.93 when calculating the microhardness according to the optimal loads and for a constant diagonal length, respectively). Calculation of H by the second method should convert the systematic error due to the effect of the surface layer into a constant quantity which is smaller the greater the diagonal length. If one omits the anomalous point corresponding to the minimum reflectingpower, the correlation coefficient in the last case increases to 0.98. At the same time, the dependence between the microhardness and the reflecting power (Fig. 1) is expressed by the following equation of a straight line, the coefficients of which are calculated by the method of least squares; power in_a POOS device in air in the visible region of the spec- trum (435-660 nm). The etalon was STF-2 silicon glass. The deviation in the arithmetical mean of the value of R for a single sample equaled 3.6% due to the inhomogeneity of the material. We utilized the value of R at A = 580 nm for comparison with the microhardness. We measured the microhardness in a PMT-3 device, calibrated to halite, at loads of 15-100 gf. The deviation in the arithmetical mean of the value of H with a single load com- prised 16% due to the inhomogeneity of the material. For com- parison with the reflecting power, the microhardness was de- termined by two methods; at optimal load for a given class [3] and for a constant length of a diagonal of the replica. In the latter case, we found the microhardness from the points of intersection of the straight line, obtained by utilizing the well- known formula H = (1.8544P)/d2 (kgf/mm2) and the corresponding diagonal length at 14 pt (the magnitude of which is limited by the size of the thin layers), with curves describing the H? P de- pendence for individual samples. The conditions under which the experiments were con- ducted and the results of the study of the synthetic pitchblendes are given in Table 1. In spite of significant fluctuations in the 0/E ratio in pitchblendes obtained at the same temperature, a distinct tendency of this ratio to fall with a rise in temperature is observed. Simultaneously with a decrease in the oxygen co- efficient, the water content in the pitchblendes is reduced and increases. The largest correlation coefficients are obtained R 0.0137H +7 .71 with the variances of the coefficients Sa = 0.0056 and Sb = 0.92. The straight line corresponding to this equation agrees in direction and passes throug the center of the correlation ellipse of the values of R and H for natural pitchblendes and uraninites [4]. It is note- worthy that the reflecting power and microhardness of synthetic pitchblendes with the highest 0/U ratio (2.33) are the same as for natural pitchblendes with a 0/U ratio equal to -2.70. The dependence of the morphology of the separating out of the synthetic pitchblendes on the tempera- ture presents considerable interest. Pitchblendes synthesized at 150?C and characterized by the highest 0/U ratios form principally small ?Mites of concentrically zonal structure (Fig. 2a); a collomorphically banded microtexture is observed in particular cases. The latter moreover is most characteristic of pitch- blendes obtained at 200?C (Fig. 2b). The pickling reveals a thin stratification (Fig. 2c); the individual layers are .1 Am thick and extend over the entire thin layer. Of the characteristics of the crystalline structure of pitchblendes synthesized at 150 and 200?C, it is not observed during microscopic study. How- ever, thin layers of pitchblendes with the lowest 0/U ratios, formed at 300 and 320?C, have a distinct crystalline structure. At the same time, they are combined with spheroidolites (Fig. 2d), sometimes changing into dendrites (Fig. 2e). As is known, the rate of chemical reactions increases with an increase in temperature (2-4 times every 10?C on the average) and the concentrations of the original reagents. In this connection, one would expect that at higher temperatures and concentrations of uranium in the initial solution the dispersion of 86 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Fig. 2. Morphology of the separating out of synthetic pitchblende (the ordinal number of the experiment in Table 1 and the magnification are indicated); a) 5, 440; b) 9, 340; c) 11, 440, pickled; d) 13, 750, pickled; e) 13, 440, pickled (the filler is indistinct). pitchblende deposits should increase due to the increase in the rate of reduction of hexavalent uranium and the formation of large quantities of the deposit per unit time. In fact, with an increase in temperature, the dispersion is reduced and the deposits have an increasingly distinct crystalline structure. The experi- mental results indicate the absence of an appreciable effect of the initial concentration on the size of the particles and the morphology of the seParating out of the pitchblende, which agrees with the experimental data obtained in [1]. One can explain these discrepancies by the formation at different temperatures of pitchblendes with a distinct 0/U, which is also the most important factor determining the structure and properties of the products of the synthesis. The higher this ratio, the higher the dispersion of the deposits and their water content and the lower the reflecting power and microhardness. This dependence is in good agreement with the results of the investigation of natural uraninites and pitchblendes. The effect of the 0,/U ratio is probably connected with the ordering of the crystalline structure UO2 +x, which occurs with the reduction of this ratio. Unfortunately, pitchblendes synthesized by the method described are distin- guished by an increased crystal lattice parameter (up to 5.56 I), which hinders the determination of the dependence ,of the parameter on the 0/U ratio. It had been assumed previously [1] that the increase in the parameter is connected with the presence of water in the pitchblendes; however, it did not eliminate the possibility that the presence of As203 affects the parameter [5]. The available data does not permit one to solve this problem. 87 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 One can assume that the values of 0/U given in Table 1 do not reflect stable phase ratios under ex- perimental conditions and should decrease with an increase in time of the experiments. This question re- mains unclear: is the formation of pitchblendes with 0/U > 2.38 possible at lower temperatures or for shorter experimental periods? This maximum 0/U ratio, obtained in the experiments described, agrees with the limiting 0/U ratio in the cubic phase UO2+ x, synthesized by a quite different method [6]. How- ever, the possibility that this agreement is accidental is not excluded. LITERATURE CITED 1. R. P. Rafal'skii, The Physicochemical Study of the Conditions of Formation of Uranium Ores [in Russian], Gosatomizdat, Moscow (1963). 2. V. K. Markov et al., Uranium, Methods of Its Determination [in Russian], Gosatomizdat, Moscow (1960). 3. S. I. Lebedeva, Trudy IMGRE, No. 6, _89 (1961). 4. M. V. Soboleva and I. A. Pudovkina, Uranium Minerals [in Russian], Gosgeoltekhizdat, Moscow (1957). 5. Yu. M. Dymkov, The Nature of Pitchblende [in Russian], Atomizdat, Moscow (1973). 6. R. P. Rafal'skii et al., Dokl. Akad. Nauk SSSR, 224, No. 5, 105 (1975). 88 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 MEASUREMENT OF THE ENERGY DEPENDENCE OF 71 2 3 3 U IN THE 0.02-1-eV REGION V. A. Pshenichnyi, A. I. Blanovskii, UDC 539.125.5 N. L. Gnidak, and E. A. Pavlenko Accurate data on the effective number of low-energy fission neutrons r1233U are important in the de- sign of slow-neutron breeder reactors since the principal uncertainty in the breeding ratio for such reac- tors is introduced by the indeterminancy of ?233U. In view of this, the energy dependence of 11233U over the 0.02-1-eV range was determined in the VVR-M atomic reactor of the Institute of Nuclear Problems of the Academy of Sciences of the Ukrainian SSR with an accuracy of 1-2%. The measurements were conducted using the time-of-flight method with a. resolution of ?12 pisec/m and normalized to the value when E = 0.0253 eV, assuming that 77 = 2.297 at this point. The neutron beam from the reactor was passed into a pulsating mechanical chopper with a rptor 300 mm in diameter and slots 2 mm wide. The sample and de- tectors were located 500 5 cm from the chopper. The fission neutrons were recorded by a bank of 50 SNM-37 helium-filled counters with a moderator in the geometry near 2r. A cadmium shield was placed between the moderator and the counters in order to shorten the lifetime of the neutrons. The flux of inci- dent neutrons was measured by the y quanta from neutron captures by a sample of cadmium or indium. Such a (n, y) detector consists of an NaI(T1) crystal 70 x 70 mm in size and a PM-49. The transmission was measured by a SNM-5 boron-filled counter or by three SNM-37 helium-filled counters. Theoretical corrections for the resolution, for the energy-dependence of the flux detector's sensi- tivity, and for multiple scattering of neutrons in the sample were introduced into the experimental results. The magnitude of the correction for multiple processes in the fissionable sample, calculated by a Monte Carlo method, was about 2.5% in the 0.025-eV energy region and increased to 3.5% in the vicinity of 1 eV. The correction for the resolution was about 2% in the 0.2-eV energy region relative to the value when E = 0.0253 eV. Inexact knowledge of this correction introduced a systematic error of the order of 0.5%, al- most comparable with the statistical error in the measurements of ?0.7%. E, eV ' Channel number Fig. 1. Energy dependence of 77233U in the 0.02-1-eV region; a) cadmium; 0) indium. Translated from Atomnaya Energiya, Vol. 40, No. 1, pp. 76-77, January, 1976. Original article submitted May 26, 1975. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 89 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 The energy dependence of 7233U (the flux was measured by an (n, y) detector with samples of cadmium and indium) is shown in Fig. 1. Systematic discrepancies in the data obtained in the determination of the flux with indium and cadmium detectors are evidently connected with some indeterminancy in the flight distance in measurements with indium due to its greater thickness. In the last five years, measurements of 77233U were conducted in the 0.02-1-eV energy range [1, 2]. The results of these papers agree; the reduction in the value of 71 by 4.0-4.2% at 0.16 eV from the value at E = 0.0253 eV is emphasized. The energy dependence of II was determined by the simultaneous mea- surement of the fission and capture cross sections. In this paper, another technique is utilized: the energy dependence of the quantity Vol is measured indirectly. The results agree to within 1% with the data of the papers cited, although the reduction in the value of T1 at 0.16 eV, obtained in the measurement of the flux in cadmium, amounted to 5% of the value of n at E = 0.0253 eV. This agreement indicates that with the same precision the average number of fission neutrons remains constant in the thermal-energy region. LITERATURE CITED 1. L. Weston et al., Nucl. Sci. Engng., 42, 143 (1970). 2. J. Smith and S. Reedez, in; 2nd Conference on Nuclear Cross Sections and Technology, Washington, 1968, Vol. 1, p. 591 (Proc. NBS Spec. Publication 299). 90 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 INFORMATION NEXT PROBLEMS IN THE DEVELOPMENT OF OXIDE FUEL ELEMENTS FOR FAST POWER REACTORS I. S. Golovnin One of the chief features in the current development of nuclear power is the creation of nuclear power stations with fast reactors. Considerable advances have been made in a number of countries as to the technical design of such reactors. Experimental industrial stations with powers of the order of 300 million (electrical) kW each are already in the stage of adoption in the USSR, Britain, and France. Experience in the use of these installations and the more powerful versions now being built in the USSR and USA (BN-600 and FFTF) should over the next decade provide all the data required to create and in- troduce large electrical power stations of optimum parameters with fast sodium breeder reactors using mixed oxide fuel. At the present time scientists are assiduously exchanging the results of research and technical ex- perience. Several recent large international conferences have, in effect, summarized progress up to 1975 and pointed the way to future developments. This specially applies to the winter session of the American Nuclear Society (Washington, October 27-31, 1974),* the First European Nuclear Conference (Paris, April 21-25, 1975), t and the Congress of Specialists from the International Agency on Atomic Energy Regarding Possible Damage to Fast-Reactor Fuel (Seattle, USA, May 12-16, 1975). Earlier in- vestigations were mainly aimed at demonstrating the potentialities of fast reactors and elucidating the basic problems, especially those concerning the fuel elements. The limited amount of information available as to the influence of reactor conditions on the proper- ties of materials prevents all the effects which began to appear on increasing the power, the neutron flux density, and the burn-up of the fuel in the experimental installations from being taken into account at the? same time. One of the most important effects, which has not as yet been studied over a sufficiently wide range of irradiation (up to 4 .1023 neutrons/cm2 at E > 0.1 MeV), is the softening of the construction mate- rials, i.e., the fall in long-term strength and ductility, the acceleration of high-temperature embrittlement (austenitic stainless steels), and creep. Effects which have been discovered include the neutron-induced swelling of construction materials and the effect of the initial purity of the fuel and the accumulation of fission fragments on the compatibility of the fuel with the material of the fuel-element can. On the whole these effects lead to a certain (though not critical) reduction in the efficiency of fast reactors as compared with the original optimistic estimates, although the exact extent of the effect cannot yet be assessed. Apart from design and technological developments, important features in the creation of efficient fuel elements include the study of materials and structures (both inside the reactor and after removal), analytical investigations into efficiency based on experimental data regarding the properties of materials, and an analysis of emergency situations from the point of view of their influence on the efficiency of the installation as a whole and its safety under service conditions. Experimental investigations based on both cooperative and independent programs are being carried out by scientists in the United States, Britain, France, West Germany, Japan, Italy, and the USSR. Many years' experience in the use of the first fast sodium reactors has shown that the number of fuel-element failures involving oxide and mixed oxide fuel is less than 1% for burn-ups of more than 10% of the heavy atoms. In individual cases fuel elements re- main efficient up to a burn-up of 15-18% of the heavy atoms. Damage only appears visually in 0.1% of the total number of fuel elements studied. *See At. nerg., 38, No. 4, 268 (1975)? fSee At. nerg., 39, No. 3, 230 (1975). Translated from Atomnaya fnergiya, Vol. 40, No. 1, pp. 78-80, January, 1976. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, NY. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 91 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 There are two main causes of damage: production defects and the exhaustion of efficiency. Experi- ence in the manufacture of 25,000 fuel elements for the Phenix reactor (France) revealed that, subject to qualified technological monitoring, production defects could be completly eliminated. As yet it has not been found possible to predict the limit of efficiency exhaustion with a sufficient accuracy, since damage criteria have not been finally established, nor have the damage mechanisms been clearly formulated. Furthermore, experimental data regarding the influence of irradiation on the properties of materials are as yet insufficiently reliable. Hence the limiting degree of burn-up of the fuel in the fuel elements is to a large extent chosen on an approximate or empirical basis. For example, in the French Rhapsodie, Phenix, and Superphenix reactors the burn-up is specified as 8, 6.5, and 9% of the heavy atoms respec- tively. Criteria relating to can damage depend on the means of loading, and may be determined, firstly, by reference to the degree of nonuniformity of the plastic deformation of the material and the ductility limit, secondly from the deformation rate, thirdly from the steady creep velocity, and fourthly from the yield stress of the material. Detailed study has been devoted to the interaction between the mixed oxide fuel cores and the austen- itic stainless steel cans which arises at temperatures exceeding 500?C, mainly as a result of the action of volatile cesium, iodine, and tellurium fragments accumulating in the reaction zone at the high oxidizing potential of the medium. The rate of intercrystallite penetration is high until roughly 3% of the heavy atoms have been consumed, after which it rapidly diminishes. For almost stoichiometric fuel composi- tions the depth of intercrystallite penetration may amount to 70-120 J1 for deep burnup, but it is much less for substoichiometric fuel. For superstoichiometric fuel, however, a uniform frontal oxidation of the inner surface of the can occurs to a depth of 10 ?, without intercrystallite rupture. A reduction in the density of the fuel, a rise in the linear power of the fuel element, and an increase in the gap between the can and the core, especially in the presence of eccentricity, intensify the interaction. West German re- search showed that the intercrystallite penetration depended on the composition of the steel. German types of steel may be placed in the following series of increasing interaction (the brackets indicate foreign analogs to the West German brands of steel): 1.4988 (AISI-318, FV-548); 1.4919 (AISI-319); 1.4981 (010116N15M3B); 1.4970 (12R72HV). The difference in the depth of intercrystallite penetration amounted to as much as 20% for different types of steel. The interaction due to the presence of fission fragments does not lead to any direct damage to the oxide-fuel fuel elements, but it weakens the can. A change to the use of carbide fuel fails to eliminate interaction: In this case carburization of the inner surface of the can may occur to a depth approaching 40 ?. Visible damage to the fuel elements consists of longitudinal cracks in the upper and central parts of the cans arising as a result of the exhaustion of efficiency in the material. The central cracks correspond to the region of maximum diametral deformation of the fuel elements and are most probably associated with a gradual increase in the mechanical action of the swelling core and the gas pressure inside the fuel ele- ment. The cracks in the upper, hot part, in which the oxide fuel is almost completely softened, are as- sociated with the transient operating conditions of the fuel element, involving a considerable loss of ductil- ity by the can (thermal ratchet effect). These mechanisms go hand in hand, and it is as yet hard to say which is the most serious. For boosted irradiation conditions, cracks in the hot part of the fuel element predominate. Cracks in the middle are usually formed after diametral distortions of over 1%. The per- missible deformation of the material in the upper part of the element can hardly be taken as greater than 0.1%. Under normal service conditions the initiation and development of damage in one fuel element does not have any major effect on the whole assembly. The time between the loss of hermetic properties and the opening of a crack may extend to several months. Investigations into the behavior of fuel elements with artificial defects have revealed a relationship between the development of damage and the formation of sodium uranoplutonate under the can. Only if the coolant flow is reduced or stops altogether, or if in- dividual open cross sections of the assembly become clogged, is it possible for damage to propagate to the neighboring fuel elements; in the case of oxide fuel the emergency is localized within a single as- sembly. The individual mechanisms leading to the exhaustion of the efficiency reserve of the fuel elements are quite closely associated, for example, with the redistribution of the fuel within the interior of the fuel element, the accumulation of fission fragments, gas evolution under the can, and so on. At the same time the influence of the neutron-induced swelling of the cans, which predominates over mechanical deforma- tion at temperatures below 525?C, and also the role of radiation-induced creep, have not yet been studied anything like sufficiently. One has the impression that the main effort should be directed at creating very strong and ductile construction materials, although combating the swelling effect is still a matter of no 92 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 mean importance. It is quite possible that it will prove desirable in the future only to use low-swelling materials for the housings of the whole fuel-element assemblies. Theoretical investigations play a major part in establishing the mechanisms of fuel-element damage. Mathematical models and computing programs developed by scientists in various countries have in general proved perfectly adequate. Differences between various theories have in the main been associated with the interpretation and assessment of individual properties, as well as the computing speed and accuracy. Some programs have been developed in great detail, and account for quite refined phenomena (the effect of changes in the isotopic composition of the fuel, the continuous change in contact thermal conductivity, radial nonuniformity of the neutron-induced swelling, etc.). These useful computing programs cannot yet be fully exploited in view of the absence of adequate experimental data relating to the properties of the materials. This factor also impedes the establishment of the limiting and statistically averaged values of the fuel-element efficiency criteria. Special attention should be given to the behavior of fuel elements under transient conditions of opera- tion. These include the rapid transient processes associated with the triggering of the emergency protec- tion system, failure in the pumps of the first circuit, a surge in the power of the reactor, partial or com- plete blocking of the coolant flow through the fuel-element assembly, and finally repeated changes in power level. The aim of these investigations (the importance of which it is hard to overestimate) is to establish the limitations which will enable fuel-element efficiency to be maintained right up to the onset of an emer- gency situation. Such criteria include, for example, the limiting plastic deformation of the construction materials in various temperature ranges, for various loading rates, and for various doses of neutron ir- radiation and degrees of chemical interaction with the ambient; the permissible power surges; the per- missible scales of fuel melting; the rupture of the can by gas pressure on overheating, and so on. The extension of steady-state and quasi-steady-state theoretical models to the case of rapidly changing param- eters requires considerable development and careful comparison between the analytical results and experi- ment. It is important, for example, to allow for the transient nature of gas evolution, the redistribution of porosity in the fuel, mass transfer on melting, the transience of the mechanical interaction between the core and the can, the growth rate of the gas pressure (which may amount to hundreds of atmosphere), and so on. A number of western countries and also Japan have undertaken combined and independent programs for studying the efficiency of fuel elements under transient conditions, including experiments inside the reactor involving power surges and coolant losses, and investigations aimed at establishing the influence of rapidly changing processes on the hydraulic and thermophysical characteristics of the fuel-element assemblies. It is also intended to carry out experiments on the heating of fuel elements irradiated to various burn-ups outside the reactor. The programs envisage the use of the TREAT pulse reactor (USA), which enables power surges to be introduced in a controlled way, with a visual assessment of the rupture process, and also loops of the experimental French thermal reactors, in which it is intended to carry out similar experiments with fuel elements previously irradiated in the fast Rhapsodie reactor. Individual experiments have already been carried out. The forming of the irradiated fuel on melting and the rupture of the can by gas pressure at points of overheating have been studied, and the rise in gas pressure in the fuel element following a power surge in the TREAT reactor has been estimated. Preliminary experiments have shown that the rupture of the fuel elements under transient conditions may be associated with the rapid evolution of gas from the molten fuel. Gas evolution from the solid fuel begins playing a major part in the transient process before the onset of melting, leading to additional mass transfer; the onset of can rupture depends on the original structure of the core. A brief power surge does not cause any serious damage to the fuel-element cans. The effect of the efficiency of the fuel elements on the question of safety in the running of fast reac- tors is of importance in connection with the possible extension of any emergency situation (including that associated with anomalous working conditions) beyond the confines of the fuel-element assemblies, to em- brace part or the whole of the active zone itself. Models are being developed to represent the mechanisms underlying the development of such hypothetical emergencies in order to establish the conditions under which the irregularity may be entirely contained within the fuel-element assemblies, so that a complete damaged assembly may be replaced without disturbing the running characteristics of the reactor as a whole.. It is considered that when using oxide fuel it should be perfectly possible to make a reliable determination of the location and extent of the damage and to replace the fuel-element assembly quite safely, subject to the development of the necessary measuring devices recording the presence and type of activity of the fis- sion fragments in the reactor. 93 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 CONFERENCES AND SYMPOSIA. THIRD CONFERENCE ON NEUTRON PHYSICS A. I. Kal'chenko, D. A. Bazavov, B. I. Gorbachev, A. L. Kirilyuk, V. V. Kolotyi, V. A. Pshenichnyi, A. F. Fedorova, and V. D. Chesnokova The conference took place June 9-13, 1975 in Kiev. The participants were 300 Russian scientists from 42 institutes and science centers of the USSR and 50 foreign representatives from 16 countries. Ninety-two reports, including reviews and single communications, were presented. The conference took place in the form of 7 sections in which the most important modern problems of neutron physics were discussed. Requirements in Regard to Nuclear Data and Their Evaluation. The Conference began with a presen- tation of the requirements in regard to nuclear data for reactor technology, thermonuclear reactors, as- trophysics, and reactor physics. The reports put into evidence that to date the requirements in nuclear data and, even more, in neutron data by many fields of science and technology have now been formulated, have received the technological and economical foundation, and have been brought to the scientists. These requirements are heaviest in reactor technology and shielding from penetrating radiation. L. N. Usachev (Physics and Power Institute, Obninsk) reported on mathematical studies of the Physics and Power Institute and on a set of programs of experimental investigations. He based his con- siderations on the conditions required for obtaining the desired accuracy of nuclear data with a minimum of expenses. N. M. Nikolaev (Physics and Power Institute, Obninsk) reported on the strategy of obtaining the required accuracy of neutron data; for this purpose, microscopical and integral experiments are com- bined in the best possible manner. Work was described which is done in the Soviet Union on experimental setups capable of satisfying a large number of the requirements of highest priority, G. E. Shatalov (I. V. Kurchatov Institute of Atomic Energy, Moscow) outlined problems related to the influence which nuclear constants have upon calculations of the blanket of a thermonuclear reactor in terms of neutron physics. The interest which scientists developing thermonuclear reactors have in neutron data increases from year to year. The expansion of the work on the evaluation of nuclear data caused great interest. Two complete files (data sets) on 235U (Institute of Nuclear Physics of the Academy of Sciences of the Belorussian SSR, Minsk) and on iron (Physics and Power Institute, Obninsk) and some other papers dealing with the evalua- tion,of the cross sections of nickel, chromium, gold, carbon, and other important construction materials and of stable nuclei (fission fragments) were presented. Noteworthy was the contribution which the report of G. Salvi (France) made to the new information on the evaluation of nuclear data. The report dealt with a method of calculating the cross sections of cap- ture, fission, and inelastic scattering in the case of heavy nuclei in the energy range 3 keV-1 MeV and calculating the cross sections of n, xn and n, xrif processes at energies of 2-20 MeV. It follows from several reports that it is now possible to automatically obtain grouped constants from the files of microscopical data (mainly from the SOKRATOR system); methods have been developed for matching nuclear data obtained from integral experiments on the basis of the grouped constants. A report by N. A. Vlasov (I. V. Kurchatov Institute of Atomic Energy, Moscow) on neutron reactions in stars was received with interest.* It was shown that it is possible to accurately determine the cross sections of radiative neutron capture at energies of 30-200 keV for solving the fundamental problem of the origin of elements. *See also At. Energ., 39, No. 2, 103 (1975). Translated from Atomnaya Energiya, Vol. 40, No. 1, pp. 80-82, January, 1976. ?1976 Plenum Publishing Corporation, 227 West 17th Street, New York, N.Y. 10011. No part of this publication may be reproduced, stored in a retrieval system, or transmitted, in any form or by any means, electronic, mechanical, photocopying, microfilming, recording or otherwise, without written permission of the publisher. A copy of this article is available from the publisher for $15.00. 94 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Declassified and Approved For Release 2013/04/10: CIA-RDP10-02196R000700070001-0 Fundamental Properties of Neutrons. This *section assembled for the first time. The topics of this section put into evidence that it is justified to introduce this section into the program of the Conference, because, in addition to applied problems, problems involving important principles must be discussed. V. M. Lobashev (B. P. Konstantinov Leningrad Institute of Nuclear Physics) presented a review on the pres- ent state of research on the electric dipole moment of the neutron and the general knowledge of this matter. He reported on an experimental setup for measuring the dipole moment of the neutron with the aid of ultra- cold neutron