Soviet Atomic Energy Vol. 54, No. 4

Body:  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 October, 1983 SATEAZ,54(4) 243-318 (1983) Russian Original Vol. 54, No. 4, April, 1983 SOVIET ATOMIC ENERGY ATOMHAA 3HEPIVIH (ATOMNAYA ENERGIYA) TRANSLATED FROM RUSSIAN CONSULTANTS UUREAU, NEW YUKK Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 SOVIET ATOMIC ENERGY- Soviet Atomic Energy is abstracted or in- dexed in Chemical Abstracts, Chemical Titles, Pollution Abstracts, Science Re- search Abstracts, Parts A and B, Safety Science Abstracts Journal, Current Con- tents, Energy Research Abstracts, and Engineering Index, Soviet Atomic Energy is a translation of Atomnaya Energiya, a publication of the Academy of Sciences of the USSR. An agreement with the Copyright Agency of the USSR (VAAP) makes available both advance copies of the Russian journal and original glossy photographs and artwork. This serves to decrease the necessary time lag between publication of the original and publication of the translation- and helps to improve'the quality of the latter. The translation began 'with the first issue of the Russian journal. . Editorial Board of Atomnaya Energiya: Editor: 0. D. KazaChkovskii Associate Editors: N. A. Vlasov and N. N. Ponomarev-Stepnoi Secretary: A. I. Artemov I. N. Golovin V. I. II'ichev V. F. Kalinin P. L. Kirillov ,Yu. I. koryakin E. V. Kulov B. N. Laskorin V. V. Matveev I. D. Morokhov A. A. Naurnov A. S. Nikiforov A. S. Shtan' B. A. Sidorenko M. F. Troyanov Copyright ,1) 1983, Plsnurh Publishing Corporation. Soviet Atomic Energy partici- pates in the program of Copyright Clearance Center, Inc. The appearance of a code line-at the bottom of the first page of an article in this journal indicates the copyright owner's consent that copies ofithe article may be made for personal or internal use. However, this consent is given on the condition that the copier pay the stated per-copy fee. through the Copyright Clearance Center, Inc., for all copying not explicitly permitted by Sections 107 or 108 of the U.S. Copyright Law. It does not extend to other kindstof 'copying, such as copying for general distribution, for advertising or promotional purposes, for creating new collective works, or for resale, nor to the reprinting of figures, tables, and text excerpts. Consultants Bureau journals appear about tix months after the publication of the original Russian issue. For bibliographic accuracy, the English 'issue published by Consultants Bureau carries the same number and date as the original Russian from which it was translated: For. example, a Russian issue published in December will appear in a Consultants BureautEnglish translation about the following June, but the translation issue will carry the December date. When ordering any volume or particu- lar issue of a Consultants bureau journal, please specify the date and, where appli- cable, the volume-and issue numbers of the original Russian. The material you will receive will be a translation of that Russian volume oi` issue. ? Mailed in the USA by Publications Expediting, Inc., 200 Meacham Ave- nue, Elrriont, NY 11003. POSTMASTER: Send address changes to Soviet Atomic Energy, Plenum Publish- ing Corporation, 233 Spring Street, New York, NY 10013. Subscription (2 volumes per year) ` - I Vols. 52 & 53: $440 (domestic); $489 (foreign) Single Issue: $100 Vols. 54 & 55: $500'(domestic); $555 (foreign) Single Article: $7.50 CONSULTANTS BUREAU, NEW YORK AND LONDON n 0 233 Spring Street New York, New York 10013 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 SOVIET ATOMIC ENERGY A translation of Atomnaya Energiya October, 1983 Volume 54, Number 4 April, 1983 CONTENTS EXPERIENCE ACCUMULATED BY SOVIET NUCLEAR POWER ENGINEERING Nuclear Power in the USSR - A. P. Aleksandrov, A. S. Kochenov, E. V. Kulov, A. G. Meshkov, V. P. Ryazantsev, and V. A. Sidorenko. Experience with the Creation, Operation, and Means of Improvement of Nuclear Power Plants with Water-Cooled-Water-Moderated Reactors VVER) - F. Ya. Ovchinnikov, Yu. V. Markov, V. A. Sidorenko, V. A. Voznesenskii, G. L. Lunin, V. V. Stekol'nikov, G. G. Bessalov, and Yu. V. Vikhorev. Some Characteristics of and Experience with the Operation of Nuclear Power Plants with RBMK-1000 High-Powered Water-Cooled Channel Reactors (RBMK) - N. A. Dollezhal', I. Ya. Emel'yanov, Yu. M. Cherkashov, V. P. Vasilevskii, L. N. Podlazov, V. V. Postnikov, A. P. Sirotkin, V. P. Kevrolev, and A. Ya. Kramerov. . . . . . . . . Development and Experience of Operating Fast Reactors in the Soviet Union - 0. D. Kazachkovskii, A. G. Meshkov, F. M. Milenkov, V. P. Nevskii, L. A. Kochetov, V. I. Kupnyi, B. I. Lukasevich, V. M. Malyshev, V. V. Pakhomov, F. G. Reshetnikov, A. A. Samrkin, M. F. Troyanov, V. I. Sh.iryaev, V. A. Tsykanov, and D. S. Yurchenko. . . . . . . . . Paths for the Development of Fast Power Reactors with a High Breeding Factor - S. B. Bobrov, A. V. Danilychev, V. A. Eliseev, 0. A. Zhukova, Yu. A. Zverkov, V. G. Ilyunin, V. P. Matveev, A. G. Morozov, V. M. Myrogov, A. I. Novozhilov, V. V. Orlov, I. S. Slesarev, S. A. Subbotin, M. F. Troyanov, and B. F. Shafrygin. Standards for Safety of Atomic Power Plants in the USSR - V. A. Sidorenko, 0. M. Kovalevich, and A. N. Isaev . . . . . Radiation Safety of Atomic Power Plants in the USSR - E. I. Vorob'ev, L. A. Il'in, V. D. Turovskii, L. A. Buldakov, N. G. Gusev, 0. A. Pavlovskii, and G. M. Parkhomenko . . . . . . . . . . . . . . . Extraction and Processing of Uranium Ore in the USSR - B. N. Laskorin, V. A. Mamilov, Yu. A. Koreisho, D. I. Skorovarov, L. I. Vodolazov, I. P. Smirnov, 0. L. Kedrovskii, V. P. Shulika, B. V. Nevskii, and V. N. Mosinets . . . . . . . . . . . . . . . . . . . . . . . Experience in Handling Spent Fuel from Nuclear Power Stations in the Soviet Union, Including Storage and Transportation - V. M. Dubrovskii, V. I. Zemlyanyukhin, A. N. Kondrat'ev, Yu. A. Kosarev, L. N. Lazarev, R. I. Lyubtsev, E. I. Mikerin, B. V. Nikipelov, A. S. Nikiforov, V. M. Sedov, B. I. Snaginskii, and V. S. Shmidt . . . . . . . . . . . . . . . . . . . . . . . . . . Engl./Russ. 243 243 251 249 263 257 270 262 279 269 285 273 290 277 301 286 309 293 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 Engl./Russ. Problems of Radiation Safety of Atomic Power Plant Personnel and the Public - E. I. Vorob'ev and 0. A. Pavlovskii. . . . . . . . . . . . . . . 314 303 The Russian press date (podpisano k pechati) of this issue was 3/ 24/ 1983. Publication therefore did not occur prior to this date, but must be assumed to have taken place reasonably soon thereafter. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 EXPERIENCE ACCUMULATED BY SOVIET NUCLEAR POWER ENGINEERING A. P. Aleksandrov, A. S. Kochenov, E. V. Kulov, A. G. Meshkov, V. P. Ryazantsev, and V. A. Sidorenko The Soviet Union has no equal as regards reserves of organic fuel. The total reserves of coal in the USSR are 5.5-6 x 1012 tons and constitute about half of world reserves. In 1981, the volumes of organic fuel production were as follows: oil (including gas condensate) 609 million tons (first place in the world), gas 465 billion m3 (second place in the world), and coal 704 million tons (second place in the world). The development of power engineering in the USSR is supported by our own resources.. Also, oil and gas are exported to countries in Eastern and Western Europe. On the other hand, most of the organic fuel reserves are located in the Asiatic part of the country, while four- fifths of the fuel demand lies in the European part. For this reason, the shipping of fuel from eastern regions into western ones constitutes about 40% of the total goods carried by the country's railroads. In recent years, the Asiatic part has provided almost all the increase in the extraction of organic fuel. Therefore, if power engineering developments were to be based only on the use of organic fuel, there would be an increasing disproportion in the location of the extrac- tion and use of fuel. There would be also an appreciable increase in the costs for trans- porting fuel resources to the European part of the country. This disproportion can be sub- stantially relieved by developments in nuclear power.. Also, nuclear power enables a reduc- tion in the cost of producing electricity in the European part, while relieving railroads from transport load and improving labor productivity mainly by reducing the number of workers required in the extraction industry and in transportation, while also modifying the fuel and power balance by reducing the proportion accounted for by oil and gas.* Oil and gas cannot remain in the basis of the world's fuel and power for long, since re- serves are limited, and therefore oil and gas should be considered primarily as valuable chemical raw materials and eliminated as far as possible from the fuel balance [1]. There are also difficulties with other traditional forms of fuel in many parts of the world. On the other hand, the increasing nuclear power during the 1970s was much less than was forecast at the end of the 1960s. Major factors that hold back development at the pres- ent time are the following: inadequate development in specialized engineering, the. lack of a practical solution to fuel breeding, delay in certain aspects of fuel reprocessing and in storage of high-activity wastes, and finally lack of preparation in public opinion. On the other hand, the efforts being made in certain countries lead one to believe that all these problems will be resolved in the next 10-15 years In the Soviet Union, the first designs for nuclear power stations began to be devised at the end of the 1940s. In 1950 it was decided to construct the first nuclear power station at Obninsk on the basis of a channel uranium-graphite reactor. This was commissioned on June 27, 1954. Operating experience showed it was reliable and safe for the staff and for the surrounding population. This gave a clear demonstration that nuclear power can be used- to produce electricity. The State program for the development of nuclear power in the USSR could not be based on nuclear power stations of a single type, since this would not provide the necessary re- liability and stability. On the other hand, the exploitation of each type of power reactor * In this section the editors publish in journal form some of the papers presented by Soviet researchers at the IAEA International Conference on Experience Accumulated in Nuclear Power (Vienna, September 13-17, 1982). In addition to the papers, the editors print two surveys of the current state of foreign nuclear power engineering. Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 243-249, April, 1983. 0038-531X/83/5404-0243$07.50 ? 1983 Plenum Publishing Corporation 243 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 on a commercial scale requires a considerable time interval and considerable financial and material resources. A scientific program was drawn up by the State Committee on the Use of Atomic Energy in the USSR to define the most suitable and economical effective power reactors for this country. The program envisaged research on various power reactors: pressurized-water and boiling- water ones, channel uranium-graphite ones, reactors with organic moderators and coolants, etc. Much attention was given to fast reactors, since from the start it was clear that large- scale nuclear power is inconceivable without them. During the research, some forms of power reactor were not brought to the stage'of pro- totype building for various reasons. It was found, e.g., that reactors with organic modera- tors and coolants are suitable only for low-power stations. Research in this area led to the creation in 1963 of a block-transportable nuclear power station of electrical output 750 kW. On the other hand, research on certain types of thermal power reactor have led to the building of prototype commercial units. A water-cooled and water-moderated reactor (the VVER-210) was built at Novyi Voronezh nuclear power station with an electrical output of 210 MW, while a pressurized-water reactor (the VK-50) was built at Dmitrovgrad with electrical out- put 50 MW, and at Beloyarsk nuclear power station a channel uranium-graphite reactor was built (with nuclear steam superheating and an efficiency of 36%). In the course of the fast-reactor program, several experimental reactors and test facil- ities were constructed for simulating and researching the physical and engineering charac- teristics. The first experimental reactor with plutonium fuel was built in 1955. In the later reactors, the fuel has been metallic plutonium, plutonium dioxide, uranium monocarbide, and uranium dioxide. The main attention has been given to designs with sodium cooling. The power outputs of the experimental reactors have gradually increased. In 1969, an experimental reactor was commissioned at Dmitrovgrad that employed fast neutrons with sodium cooling - the BOR-60 (electrical power'12 MW). The types of power reactor optimum for this country were defined during the development, construction, and operation of these prototype units. During the second half of the 1960s, the accumulated experience led to the decision to develop nuclear power engineering on the basis of two types of thermal reactor: pressurized-water ones and channel uranium-graphite ones cooled by boiling water, since at that time several VVER units had been built together with the second unit for the uranium-graphite channel reactor at Beloyarsk power station. The decision provided, firstly, the fullest use of the country's engineering facilities and, secondly, a more flexible design for the fuel cycle, since in particular the RBMK reactors can utilize fuel from the VVER. The VVER-440 and RBMK-1000 units were adopted as the standard ones. The first two units at Novyi Voronezh power station were prototypes for nuclear power station units containing VVER-440 [2]. The units are highly reliable and the load factors TABLE 1. Working Parameters of Nuclear Power Stations with VVER Units in Individ- ual Year Power station Year Installed power, MW Energy produc- tion, billion kW?h Load coeffi- cient, Novyi Voronezh 1979 1409 9,92 78,9 1980 1409 11,35 82,7 1981 2409 15,56 73,7 Kola 1979 880 5,90 76,5 1980 880 7,22 93,5 1981 880/1320 * 7,39 69/83 * Armenian 1979 407,5 2,39 66,8 1980 815 4,74 66,3 1981 815 5,50 77,0 *The denominator gives the value without al- lowance for the introduction of the new unit. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 are high (about 80%, and appreciably higher in some years), which provided a basis for the design of routine units. In the development of the VVER-440, the equipment was substantially upgraded, and major changes were made in the reactor design. The first two units containing the VVER-440 were also installed at Novyi Voronezh power station. Minor changes were made to the designs of the subsequent standard units. Subsequently, similar units were introduced at other nuclear power stations in the country (Kola, Armenian, and Rovensk), as well as in the German Democratic Republic, Bulgaria, Czechoslovakia, and Finland. At the Armenian power station, the equipment in the first loop was designed to withstand seismic shocks. Table 1 gives some parameters of nuclear power stations containing VVER-440. The high load coefficients are due to the successful design of the main unit. The length of a fuel cycle in the VVER-440 is 3 yrs, with intervals of 1 yr between par- tial reloads. The annual partial reload along with planned prophylactic maintenance occupies about 30 days. Periodic checks are made on the state of the main equipment in the first loop, and, in particular, the pressure-vessel metal is checked once every 4 yrs with the core and devices within the pressure vessel unloaded. The length of the shutdown necessary for checking the pressure-vessel metal is about 60 days. Some VVER vessels have been made of heat-re- sisting steel without anticorrosion stainless-steel coating. The first such vessel was installed in the second unit at Novyi Voronezh power station. The ammonia-potassium water treatment provides a satisfactory corrosion state in the pressure-vessel metal. The work on upgrading the VVER-440 led to the creation of the VVER-1000 reactor system, of which the first unit was commissioned at Novyi Voronezh power station. The following problems were overcome in the design of this: The equipment must be transportable by rail- road, the economic parameters should be improved, and the units should meet the latest safety requirements. Transportability for the pressure vessel restricted the diameter to 450-460 cm, with the diameter of the core zone, correspondingly, 310-320 cm. The energy production density in the core is 110 kW/liter or 30% higher than that in the VVER-440. This requires special measures to equalize the distribution. The main circulation loop was enclosed in a protec- tive shield of prestressed reinforced concrete designed for a pressure of 0.55 MPa. Popula- tion safety was also provided for instantaneous failure in DU-850 pipeline coinciding in time with the completely idle state of the power station. Commissioning operations and operating experience with a fifth unit at Novyi Voronezh power station confirmed that the basic designs were correct. On the other hand, certain changes were made in the design of the VVER-1000 to be located at the South Ukraine, Kalinin, and Rovensk power stations. Projects for power stations containing VVER-1000 were devised for regions with seismicity 5-6 points on the MSK 1964 scale. No substantial changes are proposed in the VVER-1000 reactor system for nuclear combined heat and power stations. Table 2 [3] gives some parameters of nuclear power stations containing RBMK-1000 units. At such power stations, the fuel is changed on load by a charging and discharging machine. Experience has shown that about 1000-1500 such operations can be performed at such a power TABLE 2. Working Parameters of Nuclear Power Stations with RBMK-1000 Units in In- dividual Years Energy Load Installed produc- coeffi- Power station Year power, lion, cient, MW billion a/b kW ?h Leningrad 1979 2000 13.1 711 1980 3000 18.82 71,4 1981 4000 24,1 73,8 Kursk 1979 2000 10,35 64.1 1980 2000 13,89 79,1 1981 2000 13,54 77,3 Chernoby 1979 2000 12,23 69,8 1980 2000 14,21 80,9 1981 2000 13,44 75,2 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 station during a calendar year. The condition of continuous reloading enables one to approxi- mately double the burnup by comparison with the state of simultaneous reloading for the entire core. The experience with the RBMK-1000 confirmed preliminary theoretical conclusions that the void and temperature coefficients of reactivity would increase as the fuel is burned and ab- sorbers are extracted from the core, which reduces the stability of the energy production distributions. Efforts to stabilize these distributions were concentrated on improving the level of automation by means of a branched controlled system for the reactor and changing the fuel isotope composition. A new local automatic-control system was devised and imple- mented, which showed high reliability and performance. This provided characteristic deforma- tion times for the energy-production patterns of not less than 6-8 h, which do not cause any difficulty in managing the reactor. In accordance with these studies, the initial fuel enrichment was raised to 2%. This not only improved the dynamic performance of the reactor, but also raised the economic pa- rameters by increasing the extent of burnup and reducing the specific fuel consumption. On account of elevated safety specifications, special systems were designed to provide acceptable temperature conditions in the fuel pins and to localize the escape of coolant on failure of any pipeline (including a maximum diameter one of 900 mm). Such safety systems have already been implemented at the Leningrad power station and are envisaged for all power sta- tions that are being constructed with RBMK-1000 units. The successful operation of the RBMK-1000 at its nominal power revealed considerable reserve margins in the design, so it was possible to incorporate heat-transfer intensifiers in the fuel-pin assemblies to increase the power in each channel by a factor 1.5 without changing the dimensions or numbers of the fuel channels. The design of the new fuel-pin assembly for the reactor, namely RBMK-1500, enables one to increase the thermal loads while maintaining a high level of unification with the RBMK-1000 fuel-pin assemblies. The first section of the Ignala power station with two RBMK-1500 units is now being built. The commissioning of the head unit will lay the basis for the new generation of channel reactors, which are more economical and should replace the well-recommended RBMK-1000. The construction of power stations containing the RBMK-1500 will reduce the specific capital cost by comparison with the RBMK-1000 and will reduce the costs assigned to the electricity. In the next stage of development for channel power reactors, one may go to the design of sectional block reactors with nuclear steam superheating (RBMKP) with unit powers of 1200 and 2400 MW. The gross efficiency of a nuclear power station containing RBMKP-1200 and-2400 units is expected to be about 37%. The building of nuclear power stations containing thermal reactors has been accompanied by the construction of two commercial units containing fast reactors: the BN-350 and the BN-600 [4]. Over nine years have elapsed since the power commissioning of the BN-350. The only major defect in the equipment has been repeated failure of the sealing between loops in the steam generators. After the completion of repair in the damaged steam generators, the reac- tor power was raised to 520 MW (thermal) in 1975, 650 in March, 1976, and 700 in September, 1980, which provides an electrical power of 125 MW and a daily production of 85 x 103 tons of distillate. The design burnup of 5% of heavy atoms was attained in 1976. At present, the burnup is being maintained at the level of 5.8% of the heavy atoms, which is related to the acceptable shape changes in the six-faced jackets of the fuel-pin assemblies. During the operation, the fuel pins were unified on outside diameter and sheath thickness with the pins in the BN-600. The resulting increase in the gas compensating volume reduced the pressure in the sheath, which reduced the number of cases of sheath failure by an order of magnitude. The BN-600 power unit differs from the BN-350 in having and integral (tank) style for the equipment in the first cooling loop. Steam generator of modular type also improve the reliability. The run-up to power began in April, 1980, and by December 18, 1981, the reactor was brought up to the nominal power of 1470 MW (thermal). On January 1, 1982, the unit con- taining the BN-600 had produced 3.7 billion kW-h, and it had operated at power for over 104 h. The reactor is readily controlled. Four fuel changes have been made during the operation. The maximum fuel burnup attained 7% of the heavy atoms. On the whole, the operating experi- ence has confirmed that the actual parameters correspond to the design ones. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 TABLE 3. Major Operating Nuclear Power Stations in the USSR Dates of major stages Power Electrical power of units, Reac- tor physi ener- gy re- ache station MW type cal com- nom- coxn- mis- inal mis- Sion- power pion- ing ing Novyi 1st unit - 210 VVER 12.63 09.64 12.64 Voronezh 2nd ? - 365 ? 12.69 12.69 04.70 3rd ? - 417 ? 12.71 12.71 06.72 4th ? - 417 ? 12.72 12.72 03.73 5th -1000 ? 04.80 05.80 02.81 Beloyarsk 1st ? - 100 UGR 09.63 04.64 09.67 2nd - 200 10.67 12.67 12.69 3rd ? - 600 13N 02.80 04.80 12.81 Kola 1st ? -- 440 VVER 06.73 06.73 12.73 2nd ? - 440 ? 11.74 12.74 02.75 3rd ? - 440 02.81 03.81 Leningrad 1st 1000 RBMK 09.73 12.73 11.74 2nd ? - 1.000 ? 05.75 07.75 01.76 3rd ? --1000 09.79 12.79 06.80 4th ? -- 1000 12.80 02.81 08.81 Armenian 1st ,> - 407,5 VVER 12.76 12.76 10.79 2nd ? - 407,5 > 01.80 01.80 05.80 Kursk 1st ? -.1000 RBMK 09.76 12.76 10.77 2nd ,> - 1000 ? 12.78 01.79 08.79 Chernobyl' 1st ? -1000 RBMK 08.77 09.77 05.78 2nd 1000 09.78 92.78 05.79 3rd 1000 06.81 12.81 06.82 Rovensk 1st ? -- 440 VVER 12.80 12.80 2nd ? - 440 12.81 12.81 The building and commissioning of nuclear power station units, containing BN-350 and BN-600 is an important stage in solving the problem of fuel breeding, whose final purpose will include the design of a standard fast-reactor unit to be built on a large scale. As of December 31, 1981, the installed power of nuclear power stations in the USSR was about 16 GW (Table 3). The production of electrical energy at nuclear power stations in 1980 was 73 billion kW-h, as against 86 in 1981.* The safety of nuclear power stations in use and under construction in the USSR [5] is supported by a wide range of measures, of which the following are the main ones: 1) high equality in equipment manufacture and installation; 2) state monitoring for the equipment at all stages in use; 3) the definition and implementation of efficient protective measures to prevent emer- gencies, compensate for any faults arising, and reduce the consequences of possible emer- gency situations; 4) the definition and implementation of facilities for localizing radioactive materials in the case of an emergency; 5) logical execution of all engineering and organizational measures to provide for safe- ty at all stages in the building and operation of nuclear power stations; 6) standardization of engineering and organizational aspects in the provision of safety; 7) the Government supervision system. Throughout the period of nuclear power station operation in this country, there have been no instances where an emergency has led to the need to take measures to protect the pop- ulation, although. much attention has always been given to preparing such measures. For ex- *At the start of 1983, the installed nuclear power station capacity in the USSR was 18 GW (electrical), and the electrical energy production was about 100 billion kW?h - Editor. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 TABLE 4. Releases of Gases and Aerosols Containing 131, from Two Units of Nuclear Power Stations Containing VVER-440 and RBMK-1000 Units Release Year Kola Novyi I Voronezh I Kursk Chernob C , V Gas, Ci/yr 1979 2.103 2.4.103 67.9.103 133.103 1980 2-10.3 2,13103 88,9.103 280.103 ,3il mCi/yr 1979 1 4 66 290 1980 . 1 11 458 5000 ample, for each nuclear power station there is a plan of measures to protect the staff and population in the case of hypothetical emergencies going outside the framework of the design ones. A comparison of the health risks for the staff and population in the production of elec- trical power by nuclear power stations and thermal ones indicates that the former are pre- ferable. When organic fuel is used, there is considerable environmental pollution from release of ash and gases. Current coal-fired stations consume over 2x 106 tons of coal per GW (elec.). This results in about 4 x 10' tons of ash, of which about 8 x 103 tons is released into the atmosphere. The sulfurous gases are particularly harmful, and these constitute tens of thousands of tons per GW (elec.). A difference of a nuclear power station from a coal one is that there is no such release. Also, a nuclear power station does not consume oxygen from the atmosphere. Observations have been performed for many years on nuclear power stations containing VVER-440 units, not only in this country, but also in the German Democratic Republic, Bulgaria, Czechoslovakia, and Finland, as well as on stations with RBMK-1000 units, and these indicate that radiation safety for the staff and population is reliably provided. The levels of pene- trating radiation at permanently staffed and largely unstaffed locations do not exceed 1.4 and 2.8 mrem/h, which have been set as the permissible levels. The average individual dose of external radiation to the staff does not exceed 15-18% of the maximum permissible for a year, which is 5 rem/yr. The release listed in Table 4 are substantially below the level permitted by the health rules of 1979 and are 183 Ci/MW (elec.). yr on gases and 3.65 mCi/ MW (elec.). yr on 131I. With such small releases to the environment, the radiation back- ground in the locality is determined by natural sources of ionizing radiation as well as by the artificial radionuclides formed by nuclear weapons tests. It is impracticable to distin- guish the radionuclides derived from the station against the background of the global radio- nuclides. The y-ray dose outside the nuclear power station does not increase with operating time and does not vary with distance from the station. . In the Soviet Union, nuclear power is considered as a very important means of solving major problems in the fuel and energy balance over a long period. The safe and reliable operation of existing nuclear power stations goes with their minimal effects on the environ- ment and their high economic performance, and at the 26th Congress of the CPSU a decision was therefore taken to provide for the increase in production of electricity in the European part of the country mainly by the construction of nuclear power stations and hydroelectric ones. Correspondingly, nuclear power stations are being constructed at over 20 sites and will gradually displace base-load stations employing organic fuel in the northwest, west, center, and south of the European part of the country. Nuclear power stations are being constructed along the Volga and in the Ural. The installed power of the individual stations is 4-6 GW. The rise in installed nuclear power station output in 1981-1985 is planned to be provided mainly by the introduction of RBMK-1000 and VVER-1000 units. This decision substantially eases the problems in the fuel and energy balance. On the other hand, less than 25% of the energy resources go to the production of electricity, and during the next five-year period nuclear power stations can provide electrical energy only to base-load users in the European part of the country, so the contribution of nuclear power to the fuel and energy balance can hardly exceed 10-15%. This means that nuclear power sta- tions, while substantially alleviating the problem of the power and fuel balance, cannot provide a radial solution. Solutions to the problems can come only from substantial extension of nuclear-power applications. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 In the USSR, about 20% of the organic fuel used is employed in centralized heat supply, and this is mainly the scarcest form, namely gas and oil. The main users of centralized heat supply are located in the European part of the country, i.e., in regions furthest from the sources providing for an increment in organic fuel production. Therefore, the extension of nuclear power to centralized heat supply is a major task in solving the fuel and energy prob- lems. The first steps have already been taken. The Bilibinsk combined heat and power station has been operated since 1973 and provides heat to the population, while the Shevchenko nuclear power station provides fresh water, and the heat from the Beloyarsk, Leningrad, Kursk, and Chernobyl' nuclear power stations is also utilized. To reduce the consumption of organic fuel in centralized heat supply, two major stations have been built for domestic heat supply close to the cities of Gor'kii and Voronezh, which will supply users with hot water. These heat stations are reliably protected from accidents such as explosions, aircraft crashes, etc. It is impossible for radioactivity to reach the users, because there is an intermediate circuit in which the pressure in the coolant is less than that in the heat-user circuit. These features of the heat-supply stations make them a reasonably powerful (300-500 MW) and safe source of heat supply, which can be located in major inhabited areas. Under these conditions there is no need to lay long and expensive heat-carrying pipes. The first major combined heat and power station is being constructed near Odessa, in which the production of heat will be accompanied by the production of electrical power. The energy source is provided by a VVER-1000 reactor. Studies are being made on the scope for building nuclear power stations to supply steam for industrial purposes. The rates of introduction of nuclear heat sources are planned to increase substantially in subsequent five-year periods. Over 15% of the organic fuel is used directly in industry, including chemistry, metal- lurgy, etc. The introduction of high-temperature reactors will further extend the use of nuclear power, including the production of synthetic fuel. Of course, it is essential that the general use of nuclear power in branches of the econ- omy using substantial amounts of energy must be reliably supported with nuclear fuel. With the existing thermal reactors, the energy yield from a ton of natural uranium is not more than 7.5 X 103 MW-day/ton, i.e., the extent of use of the natural uranium is not more than 1%. A solution to the fuel problem requires a substantial increase, namely by about an order of magnitude by comparison with the existing level. This is possible, in particular, by the introduction of fast reactors, in which the use of natural uranium can be increased by almost two orders of magnitude. Therefore, considerable attention is being given to fast-reactor development in the USSR. As developments proceed, the proportion of these in the structure of nuclear power will increase. One of the future problems is to prepare various branches of the industry for the routine introduction of fast reactors. Unfortunately, such reactors cannot take on the role of basic energy sources in many areas. It is undesirable to operate them with a variable load graph. Also, they can hardly provide the basis for centralized heat supply, since, firstly, the unit power is too large (in order to increase the economic performance and improve the neutron balance, fast reactors of unit power 800-1600 MW (elec.) are devised, while to provide heat one requires sources mainly of 300-500 MW (thermal) and less), and, secondly, for reasons of safety they have to be located at considerable distance from heat users. They also cannot produce high-potential heat (about 1000?C) efficiently. At present, it is difficult to establish with certainty the optimum proportion of fast reactors in the structure of future nuclear power. This will be dependent on the structure of energy use, and, in particular, on the level of electrification. However, one assumes that the reasonable proportion of such reactors can hardly exceed about 0.5. The problem is that when the proportion is less than 0.5, it is still necessary to provide developing nuclear power with artificial fuel in the necessary amount. Table 5 gives theoretical values for the rate of accumulation of excesss plutonium in fast reactors with sodium cooling for external fuel-cycle durations of 1 and 3 yrs. The less the duration of the external cycle, the higher the rate of plutonium accumulation, but also the more complicated the shipping and reprocessing of the spent fuel. Table 5 shows that with an external fuel-cycle duration of 1 yr, the BN-800 and -1600 (fuel reproduction coeffi- cient RC = 1.3) can provide a rate of accumulation of plutonium of about 0.05 yr-3, while the improved BU units (RC = 1.55) can provide about 0.08 y-' [6]. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 TABLE 5. Rates of Accumulation of Excess Plutonium External fuel cycl Reactor duration, yrs Rate of accumula tion of excess plutonium, yr BN 3 0,028 BN 1 0,051 BU 1 0,077 TABLE 6. Equilibrium Proportion of Fast Reactors Nuclear power Devel- opment Length of ex- ternal Input of natural uranium, ton/ GW (elec.) ? yr structure rate, yr-1 fuel cy- cle, yrs 0 I 10 20 BN + LVR 0,01 3 0,83 0,71 0,60 BN + LVRU 0,01 3 0,73 0,54 11,38 BN + LVR 0,01 1 0,79 0,68 0.57 BN + LVRU 0,01 1 0,67 0,48 0,30 BU + LVR 0,01 1 0,64 0,55 0,46 BU + LVRU 0,01 1 0,49 0,35 0,22 BU + LVR 0,03 3 * 0,90 0,77 BN + LVRU 0,03 3 * 0,84 0,63 BU + LVR 0,03 1 0,90 0,78 0,66 BN + LVRU 0,03 1 0,83 0,64 0,45 BU + LVR 0,03 1 0,75 0,65 0,55 BU + LVRU 0,03 1 0,63 0,49 0,34 *Fast reactors are not capable of provid- ing the required rate of accumulation in nuclear fuel. Table 6 gives the equilibrium proportion of fast reactors in relation to the rate of development of nuclear power, the duration of the external fuel cycle, and supplies of natural uranium. By LVR is meant a water-cooled reactor with the characteristics of the VVER-1000 (the RBMK has similar characteristics), while by the improved LVRU is meant a reactor with characteristics analogous to those of the LVR but with the oxide fuel replaced by a uranium of elevated density, while the improved fast BU reactors are ones with heterogenous struc- tures for the core and sodium cooling [6]. Table 6 shows that the minimum proportion of fast reactors is 0.55 (if the BU is used) if these work together with the LVR, or 0.34 with the LVRU even for the comparatively low rate of development in nuclear power of 0.03 yr-1, and with a consumption of natural uranium of 20 tons/GW (elec.). yr, which corresponds to an average consumption of the uranium of about 5%. The equilibrium proportion is appreciably higher if there is a lower RC and the length of the external fuel cycle is 3 yrs. Therefore, these preliminary calculations show that a solution to the problem of reli- able nuclear fuel supplies requires fast reactors with RC > 1.5, and external fuel cycles of 1 yr, and also improved thermal reactors with RC = 0.7-0.75. In principle, all three conditions can be met. Research show that fast reactors with RC > 1.5 are possible, e.q., upon using carbide fuel and a heterogenous core construction with sodium cooling. A possible competitive form may be fast reactor with helium cooling. There are no theoretical obstacles to the creation of thermal reactors with RC = 0.7-0.75. If for some reasons it does not prove possible to attain such RC in LVR with uranium fuel, then upon using thorium one can attain RC= 1 even in reactors with light-water coolant. For ex- ample, the VVER-1000 can provide RC= 0.7 while maintaining the diameter of the fuel pins, the lattice pitch, and the reactor power if uranium dioxide is replaced by a mixture of thorium dioxide with enriched uranium. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 Therefore, there is every reason to assume that the fuel problems of nuclear power will be solved, and therefore solutions will become available to the basic problems of the fuel and power balance over a long time-scale. 1. A. P. Aleksandrov, in: Nuclear Power in the 20th Century [in Russian], Atomizdat, Mos- (1974) cow . 2. F. Ya. Ovchinnikov et al., (this issue). 3. N. A. Dollezhal' et al., (this issue). 4. 0. D. Kazachkovskii et al., (this issue). 5. _V. A. Sidorenko and 0. M. Kovalevich, Papers at the IAEA International Nuclear Power and Its Fuel Cycle, Salzburg, May 2-13, 1977. Conference on 6. S. B. Bobrov et al., (this issue). EXPERIENCE WITH THE CREATION, OPERATION, AND MEANS OF IMPROVEMENT OF NUCLEAR POWER PLANTS WITH WATER- COOLED-WATER-MODERATED REACTORS (VVER) F. Ya. Ovchinnikov, Yu. V. Markov, V. A. Sidorenko, V. A. Voznesenskii, G. L. Lunin, V. V. Stekol'nikov, G. G. Bessalov, and Yu. V. Vikhorev After the start-up in 1964 of the first unit of the Novovoronezh Nuclear Power Plant (NNPP) with a gross electric capacity of 210 MW, 10 VVER reactors intended to produce 365- 440 MW of electrical power were subsequently placed into operation at the Novovoronezh, Kol'skaya, Armyansk, and Rovensk Nuclear Power Plants. During this same period the design power has been utilized in 12 units with VVER in the German Democratic Republic, Bulgaria, Czechoslovakia, and Finland (Table 1). Further construction of nuclear power plants in the USSR has continued mainly with the use of a new series of water-water reactors,(VVER-1000) with a thermal capacity of 3000 MW and a nominal value of 1000 MW for the electric capacity of the unit. The first unit with a VVER-1000 at NNPP was connected to the supply system on May 30, 1980 and reached its nominal power on February 20, 1981. The range of their application in power engineering has expanded simultaneously with the increase in the individual capacity of the reactors: Designs have been created and construction has proceeded of nuclear power plants with VVER in seismic regions, and the operation of VVER has been proposed under conditions of regulation of the frequency and capacity in power systems as well as for the combined generation of electrical power and heat. IMPROVEMENT OF THE BASIC ENGINEERING SOLUTIONS FOR NUCLEAR POWER PLANTS WITH VVER The basic engineering characteristics of nuclear power plants with VVER are given in [1-4] and in Table 2. Nineteen years after the start-up.of the first unit of the Novovoronezh Nuclear Power Plant, the individual electric capacity of the units has increased from 210 to 1000 MW, the specific intensity of the active zone from 47 to 111 kW/liter, the pressure in the reactor (absolute) from 100 to 160 kgf/cm2 (1 kgf/cm2 = 9.8 x 10? Pa), and the steam pressure in the steam generators from 32 to 64 kgf/cm2. The following solutions have remained unchanged: Six-sided heat-generating assemblies (HGA) with cylindrical fuel elements containing U02 and covered by an alloy of Zr +1% Nb were used in the active zone; high-strength chrome-molyb- denum steel was used for the reactor housing; steam generators of the horizontal type [5] Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 249-262, April, 1983. 0038-531X/83/5404-0251$07.50 ? 1983 Plenum Publishing Corporation 251 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 TABLE 1. Sequence of Entry into Service of the Power Units of Nuclear Power Plants with VVER Power plant, Attain- Connec- utiliza- power unit Reactor m~qt of tion to tion of crrucalr- supply 100% of y system capacity Novovoronezh, VVER-210 17.12.63 30.09.64 .31.12.64 I Rheinsberg VVIr R-70 11.03.66 06.05.66 10.10.66 Novovoronezh, VVE R-365 23.12.69 27.12.69 14.04.70 Novovoronezh, VVE R-440 22.12.71 27.12.71 29.06.72 III ovoronezh, N 25.12.72 28.12.72 24.03.73 ov Kol'skaya, I ? 26.06.73 29.06.73 28.12.73 Nord, I ? 02.12.73 13.12.73 11.07.74 Kozlodui, I 30.06.74 17.06.74 28.10.74 Kol'skaya, II 30.11.74 09.12.74 21.02.75 Nord, II ? ? 02.12.74 23.12.74 16.04.75 Kozlodui, II ? 28.08.75 26.08.75 05.11.75 Armyansk, I > 22.12.76 28.12.76 06.10.79 Lovisa, 1 20.01.77 08.02.77 09.05.77 Nord, m 06.10.77 03.11.77 03.05.78 Y'aslovske-Bo- ? 27.11.78 17.12.78 30.03.79 gunitse, I Nord, IV 22.07.79 02.08.79 31.10.79 Armyansk, II > 04.01.80 06.01.80 31.05.80 Yaslovske-Bo- 15.03.80 20.03.80 25.05.80 gunitse, II Novovoronezh, VE R-1000 30.04.80 30.05.80 20.02.81. V Lovisa, II MR-440 17.10.80 04.11.80 22.12.80 Kozlodui, 111 ? 04.12.80 16.12.80 27.01-81 Rovensk, I ? 17.12.80 22.12.80 Kol'skaya LII 07.02.81 24.03.81 Rovensk, II ? 19.12.81 27.12.81 *The nominal electric capacity of the unit is utilized at a thermal capacity of 92%. were used for production of saturated steam; and transportability of the reactor housing over the railroads of the USSR was ensured. The layouts of nuclear power plants with VVER and the designs of the absorbers and the actuators of the control elements, the intrahousing mechanisms and the main circulating pumps, steam generators, and turbines have differed in their appreciable variety. For ex- ample, along with the single-unit (one reactor + one turbine) Rheinsberg nuclear power plant in the German Democratic Republic, the second unit of the Novovoronezh Nuclear Power Plant, in which there are eight circulation loops and five turbogenerators, is successfully oper- ating with a high usage coefficient of the installed capacity (Tables 3, 4). The main changes in the equipment and systems of nuclear power plants with VVER are discussed below. Reactor and Intrahousing Mechanisms. Questions of vibration stability under the dynamic action of the coolant flow have exerted a dominant effect on selection of the design of the intrahousing mechanisms. A shift of the thermal shield in the first unit of the Novovoronezh Nuclear Power Plant in 1969 led to a reconsideration of the streamline flow conditions and the securing of all elements of the intrahousing mechanisms. The thermal shield for VVER-365 and the first WER-440 units was mounted on the hollow shaft of the reactor with welding in the upper part around the entire perimeter (Fig. 1). For VVER-1000 and later modifications of the VVER-440, the thermal shield was eliminated as a structural element, due to thickening of the walls of the other intrahousing mechanisms. It was repeatedly necessary in 1974-1975, when a defect in the zirconium covers of the fuel parts of individual control elements was discovered on some VVER-440 at the places at which they contact the steel spacer networks, to return to questions of the interaction of the coolant flow with the structural elements in the housing of VVER. The cause of the de- fects is fretting corrosion, which had arisen due to increased vibrations of the control HGA (Fig. 2) and thickening the wall of the zirconium cover from 1.5 to 2.1 mm. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 TABLE 2. Basic Engineering Characteris- tics of Reactor Installations with VVfR Parameters r a w ? c~7 w co _ ? w ow 00 a.. w[w > ,z Thermal capacity 265 760 1320 '1375 3000 of reactor, MW Number of circu- 3 6 8 6 4 latibn loops Pressure, 2 kgf/ cm : in reactor 100 100 105 125 160 in steam gener 32 32 33 47 li4 ator Temperature, C at reactor en- 250 245 248 268 288 trance at reactor exit 266 266 274 296 317 Flow rate of coal- 16 000 33 000 50 000 ` 45 000 88 000 ant through re- 3- / h actor, m Inner diam, of re- 2640 3560 3560 3560 4139 actor housing, mm Equivalent diam. 190 288 288 288 311 of active zone, cm Height of active 250 250 246 246 356 zone in opera- tional state, cm Power intensity of 38 47 83 86 111 active zone, kW/liter Number of HGA 148 343 349 349 151 in active zone Number of fuel 90 90 126 126 317 elements in a HGA Outer diam. of 1(1,2 10,2 9,1 9,1 9,1 fuel element, m Thickness of fuel- 0,6 0,6 0,65 0,65 0,67 dement c4ver- rng made from loy of Zr+1o mm Av. 1'n ar capaci 80 ty oaf fuel ele- 99 122 127 176 ments, W / cm Fuel charge into 17,0 40,0 41,5 41,5 66,0 reactor, tons of metal Specific capacity 15,5 of fuel, kW./ kg 19 32 33 45,5 of U Enrichment of 2,0 2,0 3,0 3,5 4,4 makeup fuel upon relace- lI 0At of 1/3 Fuel depletion 13 14 28 30 40 depth, MW/ day Number of SCR 19 37 73 73 or 3 1()9 assemblies *With the operation of seven loops (one is a reserve loop). Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 TABLE 3. Expenditures for Construction and the Average Technico-economic In- dices of the Operation of Nuclear Power Plants with VVER in 1977-1981 Indices Novovoronezh, units Kol' skaya, units Arm- yansk, units I I II IIII11V 1 V I, 1, I, II Installed capaci- 210 365 880 * 1000 880 815 ty, MW (elec.) gross Specific capital b 326 256 200 308 263 327 costs, ru res/ kW Electrical power 7,4 13,9 30,1 6,0 1 31 ,0 15,4 output, billions k Average usage 0,80 0,87 0,81 - 0,80 0,62 coefficient of installed capac- ity *Since 1979 the total installed capacity of units III and IV has decreased to 834 MW in connection with the high average temperature of the water which cools the turbine condensers. tData for 1980-1981. TABLE 4. Technicoeconomic Characteristics of Nuclear Power Plants with VVER during the 1977-1981 Period Novovoronezh, units Kol'skaya, units r , yansk units ts Characteristics Year I II I III IV V I II I III I II Usage coefficient of in- 1977 0,76 0,82 0,78 0,79 - 0,59 0,73 - 0,208 - stalled capacity 1978 0,86 0,90 0,82 0,75 - 0,83 0,83 - 0,535 - 1979 0,818 0,909 0,758 0,707 - 0168 0,85 - 0,668 - 1980 0,714 0,856 0,843 0,843 0,22 0,962 0,907 - 0,774 0,551 1981 0,847 0,859 0,838 0,898 0,562 0,802 0,855 0,338 0,778 0,762 Electrical power output, 1977 1397 2614 3013 3057 - 2288 2823 - 834 - millions of kWh 1978 1584 2892 3150 2891 - 3205 3199 - 1909 - 1979 1505 2905 2850 2656 - 2616 3285 2386 - 1980 1317 2745 3079 3088 1112 3717 3507 - 2772 1974 1981 1558 2745 3061 3280 4918 3091 3297 1001 2779 2720 For further improvement of the hydrodynamic conditions of operation of the active zone on all VVER, a special perforated bottom which equalizes the flow distribution through the HGA has been installed in the lower space of the reactor, starting from the first unit of the Armyansk Nuclear Power Plant. When the new units are started up, expanded operating programs are performed for measurement of vibration and stresses in the structural elements of the reactor and the intrahousing mechanisms. At present, a great deal of experience has been accumulated on the operation of 10 VVER- 365 and VVER- 440 housings made out of steel 15Kh2MFA without noncorroding planting. A sat- isfactory corrosion state of the inner surface of the housings is provided by the obser- vance of an ammonia-potassium aqueous chemistry regime in the course of operation and by the creation of an increased ammonia concentration and the execution of measures for reduc- tion of the nitrate concentration during recharging periods. Moreover, noncorring plating has again been introduced in accordance with universal practice on the VVER-1000 and Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 Fig. 1. Construction of the thermal shield in: a) VVER-210; b) MR-365; S) MR-440 of the third and fourth units of the NNPP; d) in the commercial VVER- 440; 1) thermal shield; 2) shaft. and VVER-440, starting from the first unit of the Lovisa Nuclear Power Plant. Maintenance of the required aqueous chemistry regime has been simplified somewhat. It has become necessary for recent modifications of VVER in connection with the increase in capacity of the systems for emergency cooling of the active zone, which supply a relatively cold solution of boric acid directly to the reactor, to consider, in addition, the question of the calculated reserve of the reactor housings from the standpoint of resistance to brittle fractures. A detailed analysis of the dependence of the radiation resistance of the housing materials on the flux of fast neutrons has been performed for all VVER housings on the basis of an investigation of the properties of irradiated test samples. Additional mea- sures, which provide a guaranteed calculated reserve of housing operation, were applied for several VVER-440 housings in connection with an increase in the amount of phosphorus and copper impurities in a welded seam in the region of the active zone. Model HGA which permit reducing the maximum irradiation of housing sections by a factor of three were mounted on the periphery of the active zones of these reactors in place of 36 HGA with fuel. The Active Zone and the Control and Protection System. Fuel HGA with a covering of Zr + 2.5% Nb alloy were used for assembly of the active zone in all VVER operating on July 1, 1982. Spacing of the fuel elements in the beam was accomplished by 12-16 steel grids. The size of the "turned on" HGA for all VVER-1000 is 144 mm. In order to increase the specific capacity of the active zones, the outer diameter of a fuel element was reduced from 10.2 (MR-70 and MR-210) to 9.1 mm with a simultaneous increase in the number of fuel elements in a HGA. The ratio of the number of hydrogen atoms to the number of uranium atoms in the opera- tional state is 4.2-4.7, which provides a negative reactivity coefficient with respect to coolant temperature during operation of the reactors. The temperature coefficient of reac- Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 Fig. 2. Alteration of the design of the intrahousing mecha- nisms of the reactor to decrease the dynamic action of the coolant on the control HGA of the active zone in: a) VVER- 365; b) VVER-440 of the third and fourth units of the NNPP; c) the commercial VVER-440; 1) shaft; 2) bottom of the shaft; 3) openings for passage of coolant into the control assemblies; 4) perforated bottom. tivity is slightly positive only at the start of operation of the first loadings of VVER-440 and VVER-1000, which operate with boron regulation. The depletion depth of the fuel in a VV9R-440 is ti30 MW?days/kg of U for an average enrichment of the makeup of 3.5%. One-third of the HGA which have reached maximum depletion are unloaded each year. Fresh fuel is in- stalled on the periphery of the active zone. Regulating safety-and-control-rod (SCR) assemblies equipped with actuators are provided on VVER-70 and VVER-210 to act on the reactivity. Part of these assemblies, which are in- tended mainly for compensation of the total reactivity reserve, contain absorbers made out of boron steel in the upper part and an assembly with nuclear fuel which is similar in con- struction to a fuel HGA in the lower part. In addition, there are assemblies for emergency protection which are intended for rapid shutdown of the reactor. They do not contain fuel and have their own actuator construction. All the SCR assemblies and their actuators have become identical in the subsequent de- velopment of VVER. The problem of increasing the efficiency of the SCR system was initially solved by increasing the number of assemblies with absorbers (the second unit of the NNPP). Starting from the third unit of the NNPP, the reactivity reserve against fuel depletion and slow changes in the reactivity have been compensated by the introduction of a boric acid so- lution into the coolant of the first loop. The application of boron regulation has permitted reducing the number of regulating SCR assemblies on VVER-440 from 73 to 37. Control elements in the form of bunches of 12-18 fuel elements containing europium dioxide or boron carbide are used in the reactor of the fifth NNPP unit and other VVER-1000. The transition to a new construction of the absorbers has per- mitted reducing the height of the reactor housing by virtue of a decrease in the volume under the active zone in which the fuel portion of the SCR assemblies is positioned in VVER of smaller capacity. The larger number of control elements, including elements with absorbers one-half the length of those in the first MR-1000 reactor in the USSR, permits effectively influencing the energy distribution throughout the active zone if necessary. lifliniff Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 TABLE 5. Design Engineering Characteris- tics of the Main Circulation Pumps (MCP). Ty pes of reactors and pu mps VVER VVER VVER-440 VVER- Characteristic 210 365 100.0 MCP- MCP- MCP MCP- MCP 309 310 317 195 195 Witho ut a st uffing box with .pressur ized with pac shaft an king of d in- elec tric m otor creased moment of inerti a Flow rate, m3/ h 5250 5600 6500 7100 20 000 Coolant tempera- 250 250 270 270 300 ture, ?C Intake pressure, 100 '105 925 125 156 kgf / om2 4 5,5 5.3 43 6,75 Number of rpm nchrono s1) (s 1500 1500 1500 1500 1000 y Consumablecapac- ity (no more than), kW in cold water -- - -- 1600 7000 in hot water 11150 1500 2000 1400 5300 Mass of MCP with 42 36 51 55 150 auxiliary equip- ment, tons including elec - - - 15 48 tric motors `Replacement of MCP-138 by MCP-309A was performed in 1972-1975. TABLE 6. Design Engineering Characteris- tics of SG Characteristics VVER- 210 VVER- 365 VVER- 440 VVER- 1000 Steam capacity, 230 325 452 1468 tons/h.' Steam pressure, 32 33 47 64 kgf / cm2 Temperature, ?C; of coolant of first loop at entrance 273 280 301 320 at exit of sup- 252 252 268 290 ply water 189 195 . 225 220 Humidity of steam 0,2 .0,2 . 0,2 . 0,2 at exit no more than, 1 H atin surface m n outer di ( 1300 1810 2510 6115 . e o of pipes), m outer diam, and 21x1,5 16X1,4 16X1,4 161,5 wall thickness of pipes, mm Mass of dry steam 104 112 155 292 generator, tons Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 TABLE 7. Design Characteristics of Volume Compensators Characteristic VVER- 210 VVER- 365 VVER- 440 VV@R- 1000 Type Gas Steam Total internal vol. 4x17,5 4x10,7 38 or 44 79 VOL, m Inner diam. of cy- 1800 1500 2400 3000 lindrical part, mm Volume of nitro g- t 52 18 16 or 18 24 no or steam a inal capacity, m3 O perating pressure 100 105 125 160 kgf/cm Operating tem- 260 313 325 346 perature, ?C Total capacity of - 1680 1620 2520 heaters, kW Actuators of the "screw-nut" (VVER-70, VVER-210, and VVER-365) and the "rack-and-pinion," (VVER-440) types and several kinds of electromagnetic actuators (VVER-1000) were developed for movement of the SCR absorbers. Intrareactor Control. In order to control the energy distribution in the active zone on VVER, systems of intrareactor control are provided. The temperature is measured on VVER- 210 and VVER-365 at the exit of about two-thirds of all the HGA with fuel, and, in addition, 12 or 36 measuring channels containing 5-7 neutron-flux detectors each .are mounted in the HGA on VVER-440. The temperature at the exit of all the HGA is controlled on the VVER-1000 of the fifth NNPP unit, and neutron-flux detectors are mounted in 31 HGA. Analysis of the data of the intrareactor measurements and presentation of the results to the operator is presently accomplished by a special system linked to the information-computation complex com- mon to the unit as a whole. In the event of a breakdown of this complex, the system switches to an automatic operating mode in which information is processed by simplified algorithms. Main Circulation Pumps (Table 5). Low-inertia pressurized pumps with a synchronous ro- tation frequency of the rotor of 1500 rpm are used on 17 of the 24 operating VVER units. In order to cool the active zone in the case of disconnection from the, power system, the opera- tion of the MCP for 100 sec after shutdown of the reactor is provided on these units by means of the energy of electromechanical coasting of the main generators or special internal-dis- charge generators located on the same shaft with the turbines. Pumps with the rotor of the electric motor extended beyond the confines of the first loop are used for the VVER-1000 and also in the new designs of nuclear power plants with VVER-440. A special flywheel provides for a slow decline of the flow rate when the MCP are disconnected A test of the operation of the MCP at nuclear power plants in the USSR has shown that they are one of the most reliable pieces of equipment of a reactor facility. Steam Generators (Table 6). Steam generators (SG) with a horizontally positioned housing and a pipe bunch are used at nuclear power plants with VVER, which provides for moderate loads on the surface of the evaporation mirror. Cylindrical collectors of the primary coolant are located in the surroundings of the second loop. The pipe bunch is fabricated out of OKhl8NlOT steel, and the housing material is carbon steel. The principal structural change in the course of the improvement of SG is the realiza- tion for VVER-440 and VVER-1000 of access from above to the collectors of the first loop. In this case it proved to be possible to reject special subshaft access spots for disposition of equipment servicing the SG. Maintainance and a surveying of the sites where pipes are sealed into the collectors are accomplished from above directly from the central room of the nuclear power plant. Since a flanged joint is situated in the surroundings of the second loop, special attention to the behavior of the metal of the collectors of the first loop is necessary in the phase partition zone. Pressure Compensators. A nitrogen pressure compensator is used in the first unit of the NNPP and the Rheinsberg Nuclear Power Plant. Steam pressure compensators having better Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 weight-size characteristics (Table 7) are used in the equipment complement of the remaining VVER reactor facilities. The housings of the pressure compensators are fabricated out of carbon steel. On units in which the reactor housing has no noncorroding plating, it is ab- sent on the pressure compensator. Main Circulation Pipelines and Slide Valves. For all the reactors except VVER-1000, the main circulation pipelines (MCP) with an inner diameter of 500 mm (Du 500) are fabricated out of lKhl8N9T and 1Kh18N12T steels. For VVER-100 the MCP are made two-layered (on the outside - carbon steel; inside - noncorroding steel), and their straight-through cross section is Du 850. The possibility of cutting off loops with the help of slide valves having an electric actuator which closes in a time no greater than 90 sec is provided for. Rapid-acting slide valves with a hydroactuator, which have been used on the MCP of the first unit of the NNPP and the Rheinsberg Nuclear Power Plant, were excluded in the subsequent designs. Safety Systems. As VVER have developed, safety mechanisms intended to limit the con- sequences of accidents and to localize radioactivity which has leaked from the main circula- tion loop (MCL) have steadily been improved [6]. The increased reliability of the emergency protection systems of the reactor, emergency cooling of the active zone, diversion of heat from the steam generators, and localization of fission products has been attained both by providing the necessary emergency arrangement of these systems and by applying better en- gineering solutions. If an instantaneous break in a pipeline with a diameter of about 100 mm with one-way outflow were the maximum design emergency for the first VVER, then the protective and local- izing mechanisms for contemporary VVER-440 (Rovensk Nuclear Power Plant) and VVER-1000 prov- ide for safety in the event of accidents right up to instantaneous fracture of the MCP coin- ciding in time with conditions of complete deactivation of the nuclear power plant. An in- crease in the temperature of the fuel element casings above 1200?C is prevented with the help of the provided systems of emeregency cooling of the active zone (water tanks connected in pairs to the entrance and exit space of the reactor; high- and low-pressure pumps). Localiza- tion of fission products which escape from the MCL is accomplished for new nuclear power plants with VVER-440 by the traditional method for VVER - with the help of a system of pres- surized rooms; the reactor room remains accessible for servicing. A condenser-bubbler pro- vides for steam condensation during the first period of a maximum design emergency. The maximum pressure in the chambers in the course of eliminating the emergency does not exceed 2.5 kgf/cm2. For nuclear power plants with VVER-1000, construction of a shell is provided which en- closes all the rooms of the MCL and the reactor room and is calculated for the total pressure arising upon outflow of all the coolant (5 kgf/cm2) with subsequent reduction in the pressure of the sprinkler system. In order to prevent the escape of activity during an accident, the installation in sequence of three rapid-acting pneumatic valves is provided on the pipelines which connect the shell to the external systems; each valve closes from its own high-pres- sure air system. A high degree of independence of the redundant protective and localizing systems is provided in the designs by means of placing them in different rooms, a separate electrical supply, etc. POWER PLANTS WITH VVER The cumulative duration of the operation of nuclear power plants with VVER from the time they were included in the grid to July 1, 1982 amounts to 150 reactor-yrs. Analysis of the operation of VVER during this period permits asserting that nuclear power plants with such reactors are capable of providing a reliable supply of electrical power to consumers, with high technicoeconomic indicators . As follows from Table 3, the specific capital expenditures for the construction of nuclear power plants decreases at first and reached a minimum sum for the third and fourth units - 200 rubles per 1 kW of installed capacity. The increase in the cost of construction for the subsequent units is explained both by factors of local significance (construction of nuclear power plants at Zapolyar'e - the Kol'skaya Nuclear Power Plant; provision of seismic-resistant buildings and equipment - the Armyansk Nuclear Power Plant) and by general tendencies: the increase in costs to provide for the safety of nuclear power plants and the rise in prices of energy resources. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 In the USSR 106.6 billion kWh of electrical power was produced at nuclear power plants with VVER during 1977-1981, which amounts to about 2% of the total amount of electrical power produced during this period in the entire country. The average cost of the production of electrical power in 1981 is significantly lower than at thermal power plants. For units which have operated longer than a year at the design capacity, the usage coefficient of installed capacity exceeds, as a rule, the design value of 0.8 (see Table 3, 4). A continuous effort is being made at nuclear power plants to increase the reliability and safety of the equipment. Within the framework of the system for collection of informa- tion on equipment failures which has been operating since January 1, 1977, the causes for the failures are investigated and classified, and equipment with reduced reliability is revealed. The information is directed to the factories-manufacturers and to the design organizations for adoption of the necessary measures. Quantitative reliability indices of the first and second units of the Kol'skaya Nuclear Power Plant during the period from the start of its operation to June 30, 1980 without taking account of the failures during the shakedown period are given below in hours: Unit as a whole Operating time to failure 1980 1902 Average recovery time 24 8.5 Shakedown period 3406 1490 Reactor with control and protection system Operating time to failure 10,000 4692 Average recovery time 2.6 4.8 Shakedown period 2923 2000 Main circulation pump Operating time to failure 57,230 20,020 i 388 250 me Average recovery t Shakedown period 0 0 Steam generator Operating time to failure 6804 11,190 Average recovery time 149 294 Shakedown period 3000 5000 These data indicate the stable operation both of nuclear plants as a whole and of the basic equipment. One should note that the failure of a unit as a whole is an event leading to complete degradation of the charge, but failure of the MCP or an SG is an event which results in the disruption of their work capacity, which leads to disconnection of the loop. Among the problems which must be solved for operating units, one should note the devel- opment of measures for constant upgrading of the safety of the units in connection with the change in the operating norms and rules which regulate safety questions. In connection with the expiration in 1984 of the design term of service of the reactor housing of the first unit of the NNPP, the question of an operation extension by means of a possible annealing of the housing, with simultaneous replacement of the reactor cover, the actuators of the con- trol elements, and part of the intrahousing mechanisms, is being considered. METHODS OF FURTHER IMPROVEMENT OF VVER The subsequent development of nuclear power plants with VVER will be accomplished by the application of more refined equipment, simplification of the layout of the MCL and nuclear power plants as a whole, optimization of the thermal engineering parameters and the fuel cy- cle, and improvement of the reliability of the systems for provision of safety. The range of possible usage of VVER in power engineering will be simultaneously expanded. A modified VVER-1000 reactor assembly which is distinguished by an improved layout of the MCL and the absence of the main shutoff slide valves is being used in the majority of the units which will be placed into operation up to 1990. The extent of the MCL and the protec- tive shell is reduced by more than 20%. The rejection of the use of covers for the HGA per- mits placing 163 HGA instead of 151 (fifth unit of the NNPP) in the active zone. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 The number of control elements is decreased from 109 to 61 with a simultaneous increase in the number of fuel elements in a bunch from 12 to 18. The design characteristics of the MR-1000 (the fifth unit of the NNPP and the modern- ized unit, respectively) are given below: Thermal capacity, MW Coolant pressure, kgf/cm2 Average coolant temperature, ?C Coolant flow rate, m3/h Outer diameter of the reactor housing, mm Height of assembled reactor, mm Equivalent diameter of the active zone, cm Height of the active zone in the operating state, cm Power intensity of the active zone, kW/liter Number of HGA Shape and type of HGA Size of an HGA with cover when "turned on", mm Fuel (U02) charge in the active zone, tons Outer diameter and spacing of the fuel elements, mm Average thermal flux, W/cm2 Operating period of the fuel, yrs Number of rechargings per operating period Enrichment of fresh fuel in the steady-state recharging regime, % Average depletion depth of the fuel, MW-days/kg of U Number of control elements Number of fuel elements in a con- trol element Number of MCP Number of revolutions of the MCP, rpm Presence of shutoff slide valves in the MCP loops Number of SG Type of SG Steam capacity of a single SG, tons/h Steam pressure at the exit from the SG, kgf/cm2 Steam temperature, ?C 3000 160 306 80,000 4535 22,592 311 356 111 151 Six-sided 3000-3200 160 306-307 80,000 4535 19,137 316 356 107-115 163 Six-sided with cover without cover 238 75.5 9.1/12.75 176 2 or 3 2 or 3 234 80 9.1/12.75 166-177 3 3 3.3 or 4.4 4.4 27 or 40 40 109 61 12 18 4 4 1000 1000 Yes No 4 Horizontal 4' 1469 1469-1575 64 64 278.5 278.5 As the development is accomplished, new more refined equipment will be used in the VVER- 1000. More compact SG constructions are being designed: vertical with natural circulation of the evaporator, and direct-flow. Steam generators with higher steam parameters (305-310?C, 70-74 kgf/cm2) are being considered in a number of possible alternatives. The development of a pump assembly with a shaft rotation of 3000 rpm and with mass characteristics approximate- ly twice as good as those of individual subassemblies. the MCP-195 is in the stage of experimental checking of the The use of a nuclear fuel other than uranium dioxide is not anticipated for VVER reac- tors up to 1990. In order to improve the characteristics of the fuel cycle, it has been Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 proposed to increase the average fuel depletion depth to 40 MW-days/kg of U by increasing the enrichment of the makeup fuel with simultaneous optimization of the recharging regime and a restriction in the structure of the active zone on the number of construction materials with large neutron absorption cross section (steel). The efficiency of the fuel elements of the VVER-1000 at such a depletion has been confirmed in an experimental microroentgenometer loop There has also been a successful test of the attainment of a depletion of ti 50 MW?days/kg of U in the regular HGA of the VVER-440 with them located in the active zone for 5 yrs. The gradual conversion of VVER to fuel of increased density or the incorporation of thorium into the fuel cycle is advisable in the future in the interests of the economy of inexpensive natural uranium. The scientific-engineering bases for such a change in the fuel cycle should be worked out in the forthcoming decade. An increase in the safety level of VVER reactor installations is mainly guaranteed by means of the further development of the system of intrareactor measurements and the incorpora- tion of operational systems for control of the state of the equipment and the metal of the MCP. At present, the USSR COMECON member-nations are working on a cooperative program which includes tests on operating reactors for working out methods of recording variations in the noise spectrum during disruptions of the normal operation of a reactor facility, including the onset of boiling in the active zone. Scientific principles and instrumentation for the detection and classification of defects in the materials of MCL.equipment using the acoustic emission method are being developed. As ordinary fuel becomes more expensive, the number of regions in the USSR in which it is economically advisable to construct nuclear power plants will increase during the suc- ceeding decades. In this connection, as well as with possible deliveries for export taken into account, VVER-1000 reactor assemblies will be calculated to withstand an earthquake with a force up to a reading of 9 (maximum acceleration at ground level of 0.4 g). The pos- sibility of using seawater for cooling of the auxiliary equipment and the placement of nu- clear power plants at sites with a humid tropical climate is foreseen. It has been proposed that nuclear power plants with VVER can participate in providing for the variable loading diagram of power systems. The requirements on such nuclear power plants at present provide for the possibility of daily disconnections from the grid for 5-8 h and weekly ones for 24-55 h, enhanced rates of change of the loading from 1-4% of the nominal capacity per min, and keeping the units in operation in the event of short-term de- creases in the frequency right down to 46 Hz. It has become necessary to solve a number of problems and first of all to produce a design for the fuel element which is efficient under long-term cyclic loads and to verify it experimental'. It is possible that VVER intended for control of the power and frequency in a system will operate at a lower power intensity than baseline reactors but with higher coolant parameters. An important aspect of the use of VVER is their application for a centralized heat sup- ply. The technicoeconomic discussion carried out up to the present of the alternatives for the heat supply of a number of large cities in the European part of the USSR from sources based on nuclear and organic fuel indicates the advisability of the application of nuclear heat supply plants (NHSP) for the generation of heat and of nuclear heat-power plants (NHPP) for the combined generation of heat and electrical power in comparison with boilers and heat and electric power plants operating on organic fuel. The choice of NHPP or NHSP depends on the conditions of a given city. No significant variations of any kind are assumed in the VVER-1000 reactor assembly for NHPP. The heat supply for the consumers is accomplished from diversions of the steam of the TK-450/500-60 central-heating turbines, which provide a maxi- mum heat production of 450 Gcal/h at an electrical load of about 450 MW each. 1. A. Ya. Kramerov et al., Third Geneva Conference, USSR Lecture R/304 [in Russian] (1964). 2. V. P. Denisov et al., At. Energ., 31, No. 4, 323 (1971). 3. V. A. Voznesenskii, At. Energ., 44, No. 4, 299 (1978). 4. Yu. V. Vikhorev et al., At. Energ., 50, No. 2, 87 (1981). 5. V. F. Titov et al., The main trends in the development of steam generators for nuclear power plants with VVER in the Soviet Union," Lecture at the Soviet-Italian Seminar "Con- temporary Problems of Power Engineering," November 21-24, 1977, Moscow. 6. V. A. Sidorenko et al., At. Energ., 43, No. 6, 449 (1977). Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 SOME CHARACTERISTICS OF AND EXPERIENCE WITH THE OPERATION OF NUCLEAR POWER PLANTS WITH RBMK-1000 HIGH-POWERED WATER-COOLED CHANNEL REACTORS (RBMK) N. A. Dollezhal', I. Ya. Emel'yanov, UDC 621.039.56:621.039.524.2. Yu. M. Cherkashov, V. P. Vasilevskii, 034.44 L. A. N. P. Podlazov, V. Sirotkin, V. V. P. Postnikov, Kevrolev, and A. Ya. Kramerov During the period from the end of 1973, when the first power unit with an RBMK-1000 en- tered into service at the Leningrad Nuclear Power Plant, to January 1, 1981, the installed capacity of nuclear power of the Soviet Union grew from 3.2 to 14.6 GW. During this period 8 GW of the 11.4-GW growth in nuclear capacities, i.e., about 70%, was due to the fraction of nuclear power plants with RBMK-1000. Now nine power units with this reactor are being operated: four at the Leningrad Nuclear Power Plant (LNPP), three at the Chernobyl Nuclear Power Plant (ChNPP), and two at the Kursk Nuclear Power Plant (KNPP). In all, nuclear power plants with RBMK-1000 have generated about 200 billion kWh of electrical power. The signi- ficant increase in the capacities of the nuclear power plants of the country based on the RBMK-1000 which has been achieved in a comparatively short period of time, the successful utilization of their nominal capacity, and the reliability and safety of their operation testify to the promising outlook for channel uranium-graphite reactors, on which the develop- ment of nuclear power in the USSR will be based in the succeeding decades. One can add the following to the list of factors in favor of channel boiling reactors of the RBMK type, which have been taken into account in the process of the design and construction development and which completely support the practice of their construction and operation: The RBMK-1000 is manufactured at operating factories and does not require the construc- tion of new industrial enterprises with unique equipment; there do not exist limiting values of the individual capacity associated with the manu- facture, transporting, and maintenance of the equipment used; shattering of the main loops increases the overall safety of the reactor, since there is not complete dehydration of the active zone; due to the good physical characteristics of the reactors and the continuous fuel re- charging, the prerequisities are created for highly efficient utilization of weakly enriched fuel, the attainment of a small content of fissionable uranium isotopes in the exhausted fuel, and the production of a sufficiently large increase in the depletion due to consumption of the plutonium made in passing; and a high thermal engineering reliability of the power units is provided by a broad range of regulation of parameters by in-channel control. Approximately 1660 fuel channels and more than 200 special channels of the monitoring, control, and protection system are positioned in vertical openings of the graphite reactor stack in a square array with a 250-mm spacing. Two heat-generating assemblies (HGA) with 18 fuel elements in each are mounted inside the zirconium tube of the fuel channel. Pellets made out of uranium dioxide with an enrichment of 2% in 235U are used as the fuel. The casings of the fuel elements with an outer diameter of 13.5 mm and a thickness of 0.9 mm are fabricated out of a zirconium-niobium alloy. Water from the distributing collectors under-. heated to boiling is individually supplied to each channel. The necessary flow rate is established with the help of a channel flowmeter and a regulating valve. In the active zone %15% of the water is converted into steam. The steam-water mixture from each channel is also diverted into separators through individual pipelines. Saturated steam at a pressure Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 257-262, April, 1983. 0038-531X/83/5404-0263$07.50 ? 1983 Plenum Publishing Corporation 263 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 //4\ \Y H S t //~\\\~,/ s~~ 90 v~ o a b c d a b c d Fig. 1 Fig. 2 Fig. 1. The duration of construction of a nuclear power plant until power-up of the second units: a, b) first and second turns of the LNPP, respectively; c, d) first turn of the KNPP and ChNPP, respectively. Fig. 2. Duration of reactor assembly until the start of the flushings of the reactor systems (the notation is the same as in Fig. 1). 7.7 9 0 ~~ss~, s9 6 T ;r n r '~\r n H 2 a b c d a b c Fig. 3 Fig. 4 Fig. 3. Duration of the start-up adjustment operations from the start of flushings of the systems until power-up of the unit (the notation is the same as in Fig. 1). Fig. 4. Duration of utilization of the capacity from power-up of the unit attainment of nominal capacity (the notation is the same as in Fig. 1). of 7 MPa is directed into two turbines with a capacity of 500 MW each. The separated water, mixing with the supply water, is again fed by the main circulation pumps to the entrance of the channels. The reactor has two circulation loops whose equipment is mounted symmetrically with respect to the vertical plane passing through the reactor axis in the direction of the machine room [1]. Paired layout is used in the design of nuclear power plants with RBMK-1000. i.e., two power units are located in the main building of the nuclear power plant. Each reactor, mounted in its own shaft, has an independent circulation loop. The four turbines of the two units are arranged in series on the same axis in the common machine room which is adjacent to the main building. The units do not depend on each other, but they have a series of aux- iliary interchangeable systems, which creates definite advantages in the course of operation of the nuclear power plants and especially in the maintenance of the equipment. The adopted layout of a nuclear power plant provides for start of the construction of the main buildings of the first (I) and second (II) units and maintenance of the equipment practically simulta- neously (Figs. 1-4). As follows from the diagrams given in Figs. 1-4, the even units are activated appreciably more rapidly. If assembly of the odd units is accomplished in 1.5-2 yrs, it has proven possible to shorten this time to 8-10 months for the even units. On the average, the time to carry out the start-up adjustments on the even units is shortened by a factor of two. The nominal capacity of the leading units was utilized in 8-10 months, and in 5-6 months on the succeeding even units. This is explained to a significant extent by the fact that readiness of a large number of auxiliary systems which are common to both units is required for start-up of the leading units. In addition, the experience of performing the start-up adjustment on the first power units has shown that their extent can be reduced on the subsequent nuclear power plants. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 TABLE 1. Some Operational Indices of Nu- clear Power Plants with RBMK-1000 in 1980 Installed Production of electrical Nuclear power capacity, power, kWh CUC, plant MW Leningrad 3000 18,82 73,0 Kursk 2000 13,89 79,2 Chernobyl 2000 14,2t 80,9 Total 7000 46,92 - The start-up adjustment operations should be determined by tests of the installed equipment and complex tests of the systems, but it is advisable to perform only those of the investi- gative efforts whose results obtained on the preceding units turn out for this or the other reason to be unsuitable for the subsequent ones. This has served as the basis for the crea- tion of optimal standard procedures and programs of performing the adjustment of the systems. In the standard plot which has been developed for utilization of the capacity of a commercial unit with an RBMK-l000, six months is provided. Further reduction of the periods for utiliza- tion of the capacity of commercial units has been acknowledged as being inadvisable for the following reasons: In the first place, a specified time is required for checking the fitness of the equipment in intermediate stages which permits predicting a safe elevation of the ca- pacity at the next stage; secondly, experience in control of the unit is acquired by the operating personnel. In 1980, 73 billion kWh of electrical power was produced at the nuclear power plants of the country, and of this amount 47 billion kWh or 64.5% was produced at nuclear power plants with RBMK-1000. In comparison with the previous year, the production of electrical power at nuclear power plants increased by more than 35%. This has been achieved not only due to the introduction of new capacities, but also owing to the high value of the capacity utilization coefficient (CUC). The average value of the CUC for nuclear power plants with RBMK-1000 ex- ceeds 76% (Table 1). The data on the operational readiness of the reactor equipment of the power units is of interest. In 1979 the operational readiness coefficient (ORC) on the first turn of the Lenin- grad Nuclear Power Plant reached 85%, at the KNPP 84.5%, and at the ChNPP 88.7%, with CUC values of 74.4, 73.1, and 74.6%, respectively. These same high values of the ORC also charac- terize the operation of RBMK-1000 in 1980. For example, at the ChNPP the number of hours of operation of the reactors of the first and second units was 7522 and 7622 h, and out of these the reactors operated for 6899 and 7313 h at nominal capacity. At the KNPP the number of hours of operation of the reactors of the first and second units was 7677 and 7642 h, and the reactors operated for 6999 and 6890 h at nominal capacity. The cited operational indices of nuclear power plants with channel reactors are not inferior, judging from the published data, to the best operational indices of foreign nuclear power plants with reactor housings of equal capacity, both boiling and with water under pressure. Deep depletion of the nuclear fuel with a low initial enrichment is characteristic of RBMK-type reactors, which is provided for by continuous fuel recharging at the operating facil- ity. Fuel recharging at capacity is constantly accomplished at all nuclear power plants with RBMK-1000 with the help of an unloading-loadirig machine. A regime of continuous re- chargings permits increasing by approximately a factor of two the fuel depletion depth in comparison with the regime of one-time complete recharging of the active zone. The 235U con- centration decreases from 18-20 to ,, 3.7 kg/ton of U, and the amount of fissionable plutonium reaches ti 2.8 kg/ton of U. With such a change in the isotopic composition of the fuel, the neutron-physical characteristics of the cell are significantly altered. If in the steady- state regime of fuel recharging only the local characteristics (e.g., the power) of the chan- nels are altered but the characteristics of the reactor as a whole remain practically con- stant, then the most important changes in its physical characteristics, in particular, the reactivity coefficients (steam, thermal of the graphite, thermal from heating up) occur during the initial period of operation of a reactor loaded with fresh fuel and additional absorbers. The values of these coefficients depend not only on the isotopic composition of the fuel, but also on the presence of absorbers in the active zone. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 Experience with the operation of the RBMK-1000 has confirmed the theoretical conclusions that as the fuel is depleted and the absorbers are withdrawn, the reactivity coefficients in- crease and the stability of the energy distribution decreases. A radial-azimuthal energy distribution, for which the form of the nonsteady deformations is determined by several of the lowest harmonics, turned out to be the least stable. Measures related to stabilization of the energy distribution have been carried out in two directions: an increase in the automation by virtue of the creation of a branched system for regula- tion of the reactor; and a purposeful change in the composition of the nuclear fuel. As a result of the development of measures of the first direction, a qualitatively new system of local automatic regulation of the energy distribution (LAR) and local emergency protection (LEP) which operates from intrazonal detectors [2] has been created and introduced into operational practice. The LAR system fulfills the function of automatic stabilization of the lowest harmonics of the radial-azimuthal energy distribution. Maintaining a specified capacity of the reactor, this system can, by virtue of auxiliary elements operating in the individual mode, automatically regulate the capacity in individual regions of the active zone. The LEP system accomplishes emergency power reduction in the case of local bursts of it, which can arise due to the failure of LAR elements or for other reasons. A structural peculiarity of the LAR and LEP consists of the use, for regulation of the capacity and protection of the reactor, of groups of (from 7 to 12) slave mechanisms with a regulating rod uniformly posi- tioned in the active zone and surrounded by two LEP detectors and four LAR detectors. The average correction signal of the LAR detectors is used to control the rods. Triaxial chambers located in the central hermetic sleeves of the HGA serve as the detectors of the LAR-LEP sys- tem. As follows from operating experience the LAR-LEP system has exhibited high reliability and effectiveness. Computational investigations of the effectiveness of the measures of the second direc- tion have shown that when the initial enrichment of the fuel in 235U is increased, not only do the dynamic properties of the reactor improve, but its technicoeconomic indices also in- crease due to an increase in the depletion depth and a decrease in the specific consumption of nuclear fuel. An important dependence of the variation of the time constant of the first azimuthal harmonic of the deformation of the energy distribution (To,) on the steam reactivi- ty coefficient has been established. The smaller the value of the positive steam reactivity coefficient, the higher the stability of the energy distribution and the simpler the moni- toring of the reactor. The most rational method for decreasing the steam coefficient is an increase of the ratio of the concentration of 235U nuclei and the moderator nuclei in the active zone. A decrease in the steam coefficient due to a change to a fuel of 2% enrichement is estimated to be approximately 1.3 8, where 8 is the effective fraction of delayed neu- trons. These conclusions have served as the basis for the adoption of the solution of in- creasing the enrichment of the RBMK-1000 fuel to 2% (Table 2). The 8-yr operation of systems which provide for the control and regulation of the energy distribution in RBMK-1000 has confirmed the correctness of the engineering solutions which have been taken as the basis for their development. The combined and consistent functioning of the three systems the monitoring and protection system (MPS), which operates off lateral ionization chambers; the system for physical control of the energy distribution (SPCED) with respect to radius and height of the active zone, which uses 8-emission neutron detectors of the cable type; and the SKALA system for centralized control (SCC) - has facilitated the reliable control and regulation of the energy distribution in all operating modes of the re- actor. The accumulated experience of the assimilation and subsequent operation of the moni- toring and control systems has permitted developing and incorporating measures directed at a further increase in the reliability of their operation. Among these measures one can count the conversion of the logic portion of the MPS to more reliable integrated circuits, which have permitted appreciably developing its functional possibilities with a reduction by se- veral times in the dimensions of the electronic equipment, the replacement of the cable link in the slave mechanisms of the MPS by a belt link to increase their operational reserve, and the introduction of noncontact thyristor circuits for strong control of the MPS servomecha- nisms. The service term of the detectors for control of the energy distribution with respect to the radius of the active zone exceeds the operating time of the HGA in which they are mounted. In order to increase the reliability of operation of the detectors, soldered connec- tions have been replaced by welded ones. The detector assemblies for control of the energy Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 TABLE 2. Basic Characteristics of the RBMK-1000 Fuel Cycle Characteristic Initial 235U enrichment, % 1,8 2.0 Uranium depletion, MW-days/kg 18.5 22.3 Final 235U content in unloaded fuel, kg/ton 3.9 3.5 Decrease in steam reactivity coefficient, $ - -1.3 Ann. cnsmptn. of enriched uranium per 50.5 41.5 priming, tons/GW?yr Ann. cnsmptn, of fuel elements per priming 16.3 13.3 of reactor, thousands/GW?yr Ann. cnsmptn. of natural uranium per 169 159 priming, tons/GW?yr Oper. period of fuel, eff. days 1000 1350 distribution with respect to the height of the active zone preserve their effectiveness for 4 yrs. A great deal of attention has been devoted to the perfection of thermal automation and emergency protection systems in the interests of increasing the reliability and safety of the operation of nuclear power plants with RBMK-1000. The equation of kinetics, hydrodynam- ics, and heat transfer and algorithms of the operation of the equipment and systems for automatic regulation of the parameters of a nuclear power plant are used in a mathematical model which has been developed for the investigation of transition and emergency conditions [3]. Upon comparison of the results of calculations with the data of the dynamic processes on operating units with RBMK-1000, it has been established that the model satisfactorily de- scribes the dynamics of the power unit. Some emergency conditions associated mainly with the transition to natural circulation of the coolant have been studied on special test stands. In order to justify the reliability of the cooling of the active zone under conditions of natural circulation, three series of experiments have been performed under natural conditions on the first and third units of the LNPP and the second unit of the KNPP in steady-state and transitional regimes. As a result of the computational-experimental investigations, a set of measures have been developed which raise the reliability and safety of operation of the unit. One should point out the main ones: Automatic reduction of the reactor power right down to its complete shutdown is intro- duced with emergency reduction of the flow rate of the supply water; the modes of operation of the automatic steam-discharge devices and the number of main steam safety valves are optimized; supplementary emergency protection of the reactor for a number of engineering parameters (reduction of the flow rate in the circulation loop, an increase in the pressure in the re- actor space, and dehydration of the MPS channels) is introduced; and a system of automatic regulation of the level and pressure in the separators is con- verted into a new element base of the KASKAD type which possesses the best characteristics, and the structure of the regulation system has been improved. Based on the results of the start-up adjustment operations, experimental investigations, and operating experience, some changes in the construction of the individual reactor subas- semblies and the equipment of the circulation loop have been introduced. A large part of further structural improvement has been accomplished not only on the nuclear power plants being designed with RBMK-type reactors, but also on the nuclear power plants which are operating and being constructed. For example, a redesign of the pipelines of the steam water communications is being performed, a rearrangement of the steam pipes in the space of the separator rooms is being carried out, and optimal shimming of the steam-discharge fittings of the separators has been introduced for equalization of the steam loads and elimination of misalignments of the levels lengthwise and between adjacent separators. Experience with systematic preventive and capital maintenance of the equipment of oper- ating power units with RBMK-1000 has shown that in order to shorten the periods for carrying out the maintenance operations and to increase their quality, it is necessary to improve the Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 1000 N P 0 PD ffi Ig 1980 Fig. 5. Plot of the operation of the second unit of the ChNPP in 1980: 1) disconnection of the turbogenerator (TG) for the elimina- tion of flaws on the T-junction of the se- parator-steam generator (SSG); 2) disconnec- tion of the TG for replacement of a section of pipeline of warming steam condensate of the SSG; 3) disconnection of the TG due to failures of the automatic equipment of the pumps of the machine room; 4) shutdown of the unit for the elimination of leaks in the condensers; 5) planned shutdown of the unit for average maintenance; 6) shutdown of the unit for elimination of a leak in the MPS cooling loop; 7) planned shutdown of the unit for current maintenance. maintenance technology and the methods of cooling a shutdown reactor. A method of prelimina- ry formation of icy stoppers in the underwater communications has been developed for masssive replacement of the pressure-regulating valves and the detectors of flowmeters of the fuel channels after they have exhausted their reserve capacity. With this method of refriger- ating, the necessary operations for maintenance of the indicated subassemblies is accomplished 4-5 times faster than when the design technology is used. A special system for diverting the residual heat generation with forced circulation of the coolant has been developed for the maintenance of the pipelines of the circulation loop without unloading the fuel from the active zone. The intensive development of nuclear power has not only raised a series of immediate problems concerning providing for the safety of nuclear power plants, but has also required a tightening up of the requirements which are presented to the technical means for safety provision. First of all, this refers to emergencies associated with depressurization of the pipelines of the circulation loop. An instantaneous complete break in the pipeline with a maximum diameter of " 900 mm is adopted as the maximum design emergency. The technical means of safety provision, the principal ones of which are the emergency reactor cooling system and the accident localization system [4], are calculated for this emergency. In order to deter- mine the parameters and characteristics of these systems, a set of scientific research and experimental-structural operations is performed. As a result, systems have been designed which provide for an acceptable thermal regime of the fuel elements upon a fracture of any pipeline of the circulation loop and for localization of coolant ejections. It is well known that the fuel temperature, the temperature of the graphite stack and the metal structures, and the margin of heat exchange until a crisis are the determining parameters which limit the power of channel uranium-graphite reactors with a boiling coolant. These parameters in an operating RBMK-1000 do not reach the limiting permissible values. Thus, the maximum power of a fuel channel at the nominal reactor power is about 2600 kW with a permissible value of 3000 kW, the maximum temperature of the graphite stack is 550?C with a permissible value of 750?C, the maximum temperature of the metal structures is 300?C with a permissible value of 350?C, and the margin of heat exchange until a crisis is no lower than 1.05-1.06. A plot of the load of the second unit of the ChNPP by months is presented in Fig. 5. Experience with the successful operation of power units with RBMK-1000 at nominal capa- city and the presence of reserves in the operation of the reactor equipment indicate that one can increase the reactor power without changing the dimensions and number of fuel channels Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 by increasing the critical power of the channels at which a heat-exchange crisis arises [5]. This problem has been solved by means of applying heat-exchange intensifiers in the HGA. Test-stand experiments have shown that the power of an RBMK channel with intensifiers in- creases by approximately a factor of 1.5. The construction of a new HGA, with special me- chanisms permitting an increase in the thermal loads, which has been developed for the RBMK- 1500, has a high level of unification of the individual subassemblies with the HGA of the RBMK-1000. At present, the first turn of the Ignalina Nuclear Power Plant with two RBMK-1500 units having an electric capacity of 1500 MW each is being constructed.. Power-up of the leading unit will mark the start of the creation of a new generation of channel reactors, which, being more economical, should be a replacement for the well-recommended RBMK-1000. The con- struction of nuclear power plants with RBMK-1500 will permit reducing by 20-30% the specific capital expenditures in comparison with nuclear power plants with the RBMK-1000 and reducing the cited costs for electrical power. Experience with the operation of channel uranium-graphite RBMK-1000 with boiling coolant confirms the validity of the adopted solution of creating in the USSR a large series of nu- clear power plants with reactors of a given type. The accumulated experience in creating powerful channel power reactors is a good basis for their further refinement and development. 1. N. A. Dollezhal' and I. Ya. Emel'yanov, A Atomizdat, Moscow (1980). Channel Nuclear Power Reactor [in Russian], 2. I. Ya. Emel'yanov et al., At. Energ., 49, No. 6, 357 (1980). 3. I. Ya. Emel'yanov, S. P. Kuznetsov, and Yu. M. Cherkashov, At. Energ., 50, No. 4, 251 (1981). 4. I. Ya. Emel'yanov et al., At. Energ., 43, No. 6, 458 (1977). 5. N. A. Dollezhal' and I. Ya. Emel'yanov, At. Energ., 40, No. 2, 117 (1976). Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 0. D. Kazachkovskii, A. G. Meshkov, F. M. Milenkov, V. P. Nevskii, L. A. Kochetov, V. I. Kupnyi, B. I. Lukasevich, V. M. Malyshev, V. V. Pakhomov, F. G. Reshetnikov, A. A. Samarkin, M. F. Troyanov, V. I. Shiryaev, V. A. Tsykanov, and D. S. Yurchenko The concept of the construction of fast reactors in the Soviet Union was promoted at the end of the 1940s by A. I. Leipunskii. It was based on the supposition about the more advan- tageous neutron balance, from the physics point of view, in a reactor with a "hard" spectrum. More than 20 yrs of intensive scientific-research and experimental-constructional work have been required in order to traverse the path from physical guesses to the first BN-350 reac- tor prototype. In order to consolidate the new trend in nuclear power generation, weighty arguments concerning the advantages of fast reactors by comparison with the simpler reactors were necessary, their success already having been demonstrated at this time, but such success was also due to the large efforts ground of nuclear power generation based on thermal reactors. A number of fundamental problems had to be solved in the fields of physics, heat-mass ex- change, material behavior, chemistry, economics, and also time, in order to realize more clearly the acute necessity for the development of fast reactors. Among the most important questions requiring resolution at that time, the following may be mentioned: What realistic value of the breeding factor can be obtained in future large-sized power reactors? It is possible to ensure the necessary safety of fast reactors and is control (automatic or manual) at all possible with such a reactor? What specific power intensity must be ensured and what coolant most completely meets the demands of a fast reactor? What structural and fuel materials can satisfy the requirements of fast reactors? How practicable is the industrial chemical reprocessing of spent fuel and the commercial manufacture of fuel elements based on plutonium, and what are the fuel losses in this manu- facture? What are the reserves of nuclear natural fuel and how much time will be available to the community before the mass construction of fast reactors? In order to answer these questions, the painstaking efforts of theoreticians and experi- menters were necessary - it was necessary to construct a powerful experimental base. Thermo- hydraulic and material testing rigs were constructed, upon which research on thermophysics, hydraulics, and material behavior of liquid metal coolants (sodium, sodium-potassium, and mercury) were conducted; chemicotechnological rigs were constructed for the development of monitoring and purification of coolants; physics rigs (BR-1, BFS-1, BFS-2, and "Kobra") were constructed for the study of physics problems; and experiments on the fuel and the whole fuel cycle were developed. The joint work of the theoreticians and experimenters allowed the ini- What grouping (design) of the reactor will ensure the optimum value of the breeding fac- Translated from Atomnaya Energiya, Vol. 54, No. 4, pp. 262-273, April, 1983. 0038-531X/83/5404-0270$07.50 Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6 tial concepts to be refined significantly and allowed convincing proofs of the long-term outlook of the trend to be obtained. It was essential to proceed to the next stage of the work -the construction of experimental facilities and design work on prototype plants. The construction of the research reactors BR-2 in 1956 (capacity 100 kW, coolant mercu- ry), BR-5 in 1968 (5 MW, sodium), and the experimental power reactor BOR-60 in 1969 (60 MW, sodium) was an extremely important stage in the history of fast reactors. The development and operation of these reactors allowed the choice of the principal decisions to be con- firmed finally for the BN-350 and BN-600 demonstration reactors: oxide core, sodium coolant, dense packing of the fuel elements in the fuel element assembly. Very valuable experience was obtained in the operation of sodium radioactive, circuits. The feasibility emerged of proceeding to bulk radiation tests and investigations of different structural materials and types of fuel (mixed oxide, carbide, and carbide-nitride), and to tests of steam generators. The feasibility was shown of extended nuclear fuel breeding, using already tested decisions of the core with a breeding factor of 1.30-1.45. A great deal of experience was built up on transition regimes and safety; the high degree of safety and the excellent controllability of fast reactors with sodium coolant were confirmed (BN reactors). To summarize, all this allowed, with the necessary degree of justification, complete conversion to design work on demonstration facilities, and the gradual solution of problems of reliability and efficiency and of those scientific-technical questions which most of all are associated with large-scale facilities (fuel cycle, verification of different concepts of plant grouping and different plant designs, especially of the steam generators). The current stage of development of fast reactors in the Soviet Union if characterized by the buildup of experience in the operation of three successfully active power reactors: the experimental BOR-60 and two large commercial reactors, the BN-350 and the BN-600. Opera- tion of the BR-10 research reactor is continuing. In the present paper it is proposed to discuss only the results of recent years, as earlier information has been given repeatedly [1, 2]. BR-10 Reactor. From 1973 to 1979, the reactor has been operated at a power of up to 7.5 MW; plutonium dioxide was used as the fuel. On October 1, 1979 the reactor was shut down for overhaul. At this time, the maximum fuel burnup attained 14.2%.* All standard and ex- perimental fuel element assemblies were removed from the reactor and the hermeticity of the fuel element cans was tested. According to preliminary data, the hermeticity of the cans was destroyed for "1% of the fuel elements. Nevertheless, this did not prevent the completion of the planned program and subsequent conducting of the repair work. After draining the so- dium from the primary circuit, it was washed out three times by the steam-gas method, with a final washing with distillate. Samples were cut out at various points of the primary circuit for investigations, which confirmed the satisfactory state of the material of the circuit (steel OKh18N9T). At the instant of shutdown, the main reactor vessel was irradiated with a fluence of 8.1022 neutrons/cm2, which corresponds to "-40 displacements/atom. Measurements showed that at the site of maximum fluence, the diameter of the central duct was increased by 3.10? 0.27 mm (swelling of the steel AV/V = 2.8%). The high fluence received by the material of the reactor vessel also was one of the principal reasons for shutting down the reactor for overhaul with a replacement vessel. By means of a specially developed tool and protective facilities, the vessel was cut off from the primary circuit and withdrawn from the reactor shaft. The overhaul also provided for the replacement of part of the main plant (cold trap, pump), reactor monitoring and control systems, and electrical heating and emergency cooling systems. At the end of 1981, the new vessel was installed in the shaft and completely joined to the primary circuit. The overhaul work is continuing. The next fuel charge is being pre- pared, based on uranium nitride. BOR-60 Reactor. Major work has been carried out recently on the BOR-60 reactor, namely: An extensive material testing program has been carried out, intensive investigations have been carried on astudy of the radiation effect on the behavior of austenitic and fer- rite-inartensitic steels and fuel and moderating materials; *Here and below, we have in mind the fraction of heavy atoms. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196ROO0300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 a micromodulular steam generator of Czechoslovakian construction has been tested (1973- 1981), and investigations of a large-scale model of the BN-600 steam generator have been con- ducted since 1978; in 1981, tests of the so-called reverse steam generator, also developed by specialists of Czechoslovakia, were started; and a study of the monitoring and safety sys- tems of the steam generators is continuing; a program and experimental equipment for carrying out work on a study of sodium boiling in the core* have been developed, a study of the behavior of radioactive corrosion products and fission products in the circuit has been continued, and methods of purifying the sodium of the primary circuit from the most dangerous radionuclides have been investigated. Recent work has been completed on the creation of a design for a trap based on graphite, which has undergone tests, in the BOR-60 and then in the BN-350. By pumping the sodium through the trap in the BOR-60, 1,20 TBq (540 Ci) of cesium were removed from the circuit, and the y background in the primary circuit compartments was reduced by a factor of ti2.6 [3]. Among the overall achievements in the field of material behavior investigations, the as- similation of vibration technology in the manufacture of the fuel elements should be mentioned, which is interesting from the point of view of setting up an automated process for the produc- tion of fuel elements of mixed uranium-plutonium oxide fuel. This required the carrying out of an extensive complex of technological investigations, involving the following: determination of the conditions for achieving the required density values of the mixed fuel and uniformity of distribution of plutonium oxide in the oxide mixture; a study of the dynamics of structural changes in the initially molded filling and vibro- compacted fuel column during raising of the reactor power; a study of the redistribution of the fuel components over the height and radius of the fuel elements during its lifetime tests; a study of the temperature conditions and swelling of the fuel elements; a study of gas release [4]. As a result of the investigations carried out, confirmation emerged that fuel elements prepared from a mixture of oxides by vibration technology are able to provide the same power intensity of the core and the same burnup as pelleted fuel elements. For the final verifica- tion of these preliminary conclusions an additional program was planned, according to which in 1981 a set of fuel element assemblies was prepared, based on mixed fuel, by vibration technology; these fuel element assemblies were loaded into the BOR-60 reactor. For the purpose of determining the prospects for increasing the breeding in fast reac- tors, the investigations of a metallic fuel in the BOR-60 are of important value. The in- vestigations of a metallic uranium and uranium-plutonium fuel were directed at the prevention of large swelling of the metallic fuel and its significant interaction with the cladding. As a result of many years of functioning in the BOR-60 reactor, a large number of tests of experimental fuel elements with metallic fuel have been conducted. A burnup of '6% was achieved in the experimental fuel elements with.uranium-plutonium fuel, in conditions similar to those characteristic for the present-day fast reactors with oxide fuel. BN-350. Since the time of the power generation startup of the BN-350, the first in the Soviet Union and the most powerful commercial fast reactor at that time, 9 yrs have elapsed. The only major plant defect which appeared during the whole process of assimilation of the power of the station was a defect of the steam generators: repeated breakdown of the inter- circuit sealing [5]. The principal cause of this was the poor quality of manufacture and welding of the lower end components of the heat transfer tubes. Because of the special fea- tures of the circulation from the direction of the tertiary circuit (natural circulation in the Field tubes), concern was caused by the primary pores and by the quality of the feed water, particularly the iron content in it (15-20 pg/kg). The overhaul of all the damaged (five of the six) steam generators was completed in 1975, and the power of the facility was raised to 520 MW (thermal); in March, 1976 it was raised to 650 MW (thermal), and in Septem- ber, 1980, to 700 MW (thermal). At 700 MW the reactor provides an electrical capacity of 125 MW (elec.) and additionally generates 85,000 tons per day of distillate. In May, 1980, `Specialists of the German Democratic Republic participated in these investigations. Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 assembly was completed of the first steam generator of Czechoslovakian design and it was brought to the condition of normal operation. The time-utilization factor over the period from the instant of startup to 1977 amounted to 86%, and when operating at a power of 650 MW (thermal) and above, it was 88%, which cor- responds to "7700 h of operation of the facility on power annually. Other important economic indexes are the attained fuel burnup and the operating lifetime of the main plant. For the first time, the planned fuel burnup of 5% in the central section of the fuel element assemblies was attained in 1976. At the present time, this index is equal to 5.8% and is dependent on the permissible dimensions of the change of shape of the hexagonal sheaths of the fuel element assemblies. The initial operating lifetime of the major part of the newly developed nonstandard plant was exceeded (with the exception of the steam-generator evaporators.). The operating life- time of the plant as of January 1, 1982 is shown below: Steam-generator evaporators (after overhaul) Steam-generator steam superheaters Control and safety equipment Sodium-sodium intermediate heat exchange Primary circuit slide valves with diameter 500 and 600 mm Cold traps of primary and second- ary circuits Lifetime increased from 20,000 to 50,000 h Lifetime increased from 20,000 to 50,000 h. Maximum operating period amounts to 57,000 h 45,000-55,000 h 51,000-57,000 h Operating up to now without replace- ment Ensure normal recharging cycle; ob- served jamming of plugs eliminated by increasing the sodium tempera- ture from 200 to 250?C. Operating without breakdown of seal- ing for more than 60,000 h Operating without faults; provide absolute sealing when closed Operating up to the present time without replacement It can be seen that all the main plant has operated for more than 9 yrs without replace- ment, including all the steam-generator steam superheaters. The steam-generator evaporators after overhaul also demonstrated the considerable lifetime of accident-free operation up to 55,000 h. The evaporators of one regular steam generator operated accident-free for 56,000 h.. At the present time the generator is dismantled and has been transferred-to research; a second generator of Czechoslovakian design has been installed in its place. Almost 10 years of operating experience also confirms the high degree of safety of the facility. Thus, during the whole of this time, there was not one case of sodium leakage from the primary circuit; in the secondary circuit during the same period, two leakages were re- corded (in the sampling and oxide indication systems). In each case the leakage did not ex- ceed 10 liters. The radioactivity of the discharges into the ventilation duct is determined by 41Ar and amounted-to not more than 7.4.1011 Bq/day (20 Ci/day), and the radioactivity of discharged aerosols was a factor of 106 less than the argon activity. One shutdown of the facility oc- curred during operation, in which all safety devices functioned normally. Recently, the following systems and plants have been modernized: The geometrical dimensions of the fuel elements have been unified with the fuel elements of the BN-600 reactor (diameter 6.9x 0.4 mm); at the same time, the gas compensation space of the fuel elements has been increased, which has led to a reduction of the pressure. under the cladding, and to a reduction by a factor of 10 of cases of depressurization of the fuel ele- ment cans; the control and reactivity compensation rods have been modernized, the efficiency of the rods has been increased, and the operating time of the reactor on power between two shutdowns for recharging has been increased from 55 to 73.5 days; Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6  Declassified and Approved For Release 2013/02/06: CIA-RDP10-02196R000300020004-6 the steam generator feed has been converted to water with total and extreme purifica- tion (desalination), and a cycle of complex feed water processing has been introduced; the return valves at the head of the primary circuit pumps have been redesigned. At the present time the reactor is also being used for experimental work in physics, material behavior and sodium technology. Among the most important tasks and achievements here the following should be mentioned: a cycle of experimental work on refining the breeding parameters (conversion). Measure- ments carried out have allowed the experimental value of the conversion factor 1.05? 0.05 to be established, which agrees quite well with that predicted by a computational method (1.03); a cycle of work to study the changes of shape and mechanical properties of materials in conditions of irradiation with high fast neutron fluences; in individual fuel element as- semblies, a burnup of 6.6% has been achieved and the maximum burnup in an experimental fuel element assembly attained 7.7%. Investigations of the change of shape of spent fuel element assemblies have shown that as a consequence of radiation swelling and radiation creep, the diameter of the hexagonal can is increased on the average from 96 to 97.2 mm, with a sag of 15-17 mm. These and other experimental studies have been conducted in the BN-350 and not to the detriment of the planned tasks on the generation of electric power and distillate, the ful- fillment of which is the main criterion for assessing the activity of the staff and the ef- ficiency of the facility. NB-600. In contrast to the BN-350, the grouping of the plant of this reactor is inte- gral; the diameter of the vessel is 12.8 m and the height 13.0 m. Moreover, in contrast to the facility with the BN-350, in the unit with the