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Sample records for reaktornyij ustanovtsyi wwer-1000

  1. WWER-1000 Burnup Credit Benchmark (CB5)

    International Nuclear Information System (INIS)

    Manolova, M.A.

    2002-01-01

    In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)

  2. Advanced fuel cycles for WWER-1000 reactors

    International Nuclear Information System (INIS)

    Semchenkov, Y. M.; Pavlovichev, A. M.; Pavlov, V. I.; Spirkin, E. I.; Styrin, Y. A.; Kosourov, E. K.

    2007-01-01

    Main stages of Russian uranium fuel development regarding improvement of safety and economics of fuel load operation are presented. Intervals of possible changes in fuel cycle duration have been demonstrated for the use of current and perspective fuel. Examples of equilibrium fuel load patterns have been demonstrated and main core neutronics parameters have been presented. Problems on the use of axial blankets with reduced enrichment in WWER-1000 fuel assemblies are considered. Some results are presented regarding core neutronic characteristics of WWER-1000 at the use of regenerated uranium and uranium-plutonium fuel. Examples of equilibrium fuel cycles for the core partially loaded with MOX fuel from weapon-grade plutonium are also considered (Authors)

  3. Thermohydraulic model of WWER-1000 core

    International Nuclear Information System (INIS)

    Maroti, L.; Szabados, L.

    1987-11-01

    Safe and economic operation of the WWER-1000 type reactor requires more accurate calculation of the thermohydraulic processes than the one satisfactory for the 440 type cores. The high degree of accuracy is needed both for reactor physics calculations and for the determination of the operational safety limits of the core. The paper illustrates the most important differences between the 1000 and 440 type reactors and presents the main fields of the development work necessary to reach the required accuracy. A prediction for the capability of the computer programs after the proposed development is also given and some suggestions for the further improvement is outlined. (author) 7 refs

  4. Modernization of WWER-1000 radiation monitoring systems

    International Nuclear Information System (INIS)

    Smith, T.

    1995-01-01

    A modernization scheme of the radiation monitoring system for WWER-1000 is proposed. It has a purpose to comply with international standards and to reduce operational and maintenance cost by deleting obsolete components and reducing the number of detector channels. Detailed layouts of I/C system architecture, digital radiation monitoring system (DRAMS) architecture and LRP block diagram are presented. If planned and implemented properly, this program can provide cost savings by reducing time required to access and display data and maintenance cost by deleting obsolete parts and decreasing the number of detector channels. 3 figs

  5. Modernization of WWER-1000 radiation monitoring systems

    Energy Technology Data Exchange (ETDEWEB)

    Smith, T [General Atomics, San Diego, CA (United States)

    1996-12-31

    A modernization scheme of the radiation monitoring system for WWER-1000 is proposed. It has a purpose to comply with international standards and to reduce operational and maintenance cost by deleting obsolete components and reducing the number of detector channels. Detailed layouts of I/C system architecture, digital radiation monitoring system (DRAMS) architecture and LRP block diagram are presented. If planned and implemented properly, this program can provide cost savings by reducing time required to access and display data and maintenance cost by deleting obsolete parts and decreasing the number of detector channels. 3 figs.

  6. Advanced fuel cycles of WWER-1000 reactors

    International Nuclear Information System (INIS)

    Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, A.

    2003-01-01

    The present paper considers characteristics of fuel cycles for the WWER-1000 reactor satisfying the following conditions: duration of the campaign at the nominal power is extended from 250 EFPD up to 470 and more ones; fuel enrichment does not exceed 5 wt.%; fuel assemblies maximum burnup does not exceed 55 MWd/kgHM. Along with uranium fuel, the use of mixed Uranium-Plutonium fuel is considered. Calculations were conducted by codes TVS-M, BIPR-7A and PERMAK-A developed in the RRC Kurchatov Institute, verified for the calculations of uranium fuel and certified by GAN RF

  7. Radiation embrittlement of WWER-1000 reactor vessel steels

    International Nuclear Information System (INIS)

    Nikolaeva, A.V.; Nikolaev, Yu.A.; Kevorkyan, Yu.R.

    2001-01-01

    Results obtained on the blank samples of materials of the WWER-1000 vessels irradiated by low density neutron flux are discussed. Chemical composition of the materials is characterized by the low content of the impurities (copper and phosphorus) and high content of nickel. Dependence of the radiation embrittlement of the WWER-1000 vessel materials on metallurgic variables and damage dose is treated. The research showed that nickel largely enhanced the radiation embrittlement. New dependences for determination of the radiation embrittlement real rate of the WWER-1000 vessel materials and its conservative estimation were developed [ru

  8. WWER-1000 reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series 12, 'Reactor Simulator Development' (2001). Course material for workshops using a pressurized water reactor (PWR) Simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003) and Training Course Series No. 23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation. N. V. Tikhonov and S. B. Vygovsky of the Moscow Engineering and Physics Institute prepared this report for the IAEA

  9. Fuel improvement and WWER-1000 FA main operational results

    International Nuclear Information System (INIS)

    Rozhkov, V.; Enin, A.; Bezborodov, Y.; Petrov, V.

    2003-01-01

    The JSC NCCP experience of WWER-1000 Fuel Assemblies (FAs) fabrication and operation confirms the adequate feasibility and efficiency of fuel operation in 3-4-x fuel cycles, high operating reliability and competitive capacity as compared with foreign analogues. The work on fuel improvement is aimed at an improvement of the operating reliability and an enhancement of the fuel use efficiency in WWER-1000 advanced FAs

  10. Concentration of WWER-1000 unit power on one site

    International Nuclear Information System (INIS)

    Rousek, J.; Kysel, J.; Sladek, V.

    1987-01-01

    The problem of a suitable number of nuclear power plant units built on one site is discussed. Using an example of three sites being prepared now in Czechoslovakia, two alternatives - one with two WWER-1000 units, the other with four WWER-1000 units on one site - are evaluated from the viewpoint of long-range nuclear power development program in Czechoslovakia, costs, transmission of electric power and heat supply. (author). 10 tabs., 13 refs

  11. Analysis of operating reliability of WWER-1000 unit

    International Nuclear Information System (INIS)

    Bortlik, J.

    1985-01-01

    The nuclear power unit was divided into 33 technological units. Input data for reliability analysis were surveys of operating results obtained from the IAEA information system and certain indexes of the reliability of technological equipment determined using the Bayes formula. The missing reliability data for technological equipment were used from the basic variant. The fault tree of the WWER-1000 unit was determined for the peak event defined as the impossibility of reaching 100%, 75% and 50% of rated power. The period was observed of the nuclear power plant operation with reduced output owing to defect and the respective time needed for a repair of the equipment. The calculation of the availability of the WWER-1000 unit was made for different variant situations. Certain indexes of the operating reliability of the WWER-1000 unit which are the result of a detailed reliability analysis are tabulated for selected variants. (E.S.)

  12. Extended analysis of WWER-1000 Charpy test data

    International Nuclear Information System (INIS)

    Vodenicharov, St.; Kamenova, Tz.

    2001-01-01

    The aim of this work is to study the embrittlement rate of WWER-1000 RPV weld metal with high Ni content and to determine influence of neutron irradiation on partial energies of ductile crack initiation, stable and unstable crack propagation and post crack arrest. (author)

  13. WWER-1000 nuclear fuel manufacturing process at PJSC MSZ

    International Nuclear Information System (INIS)

    Morylev, A.; Bagdatyeva, E.; Aksenov, P.

    2015-01-01

    In this report a brief description of WWER-1000 fuel manufacturing process steps at PJSC MSZ as: uranium dioxide powder fabrication; fuel pellet manufacture fuel rod manufacture working assembly and fuel assembly manufacture is given. The implemented innovations are also presented

  14. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Hoglund, J.; Riznychenko, O.; Latorre, R.; Lashevych, P.

    2011-01-01

    In 2005 six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in the South Ukraine Unit 3. This design has demonstrated full compatibility with resident fuel designs and all associated fuel handling and reactor components. Operations have further demonstrated adequacy of performance margins and the reliability requirements for multiple cycles of operation. The LTA's have now been discharged after completing the planned four cycles of operation and having reached an average assembly burnup in excess of 43 MWd/kgU. Post Irradiation Examinations were performed after completion of each cycle. The final LTA inspection program at end of Cycle 20 in 2010 yielded satisfactory results on all counts, and it was concluded that the 6 Westinghouse LTA's performed as expected during their operational regimes. Very good performance was demonstrated in the WWER-1000 reactor environment for the Zr-1%Nb as grid material, and ZIRLO fuel cladding and structural components. Control Rod Assemblies drop times and drag forces were all within the accepted values. The LTA program demonstrated that this fuel design is suitable for full core applications. However, the topic of fuel assembly distortion resistance was re-visited and Westinghouse therefore considered operational experience and design features from multiple development programs to enhance the basic Westinghouse WWER-1000 fuel design for Ukrainian reactors. The design now includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. This paper describes briefly the development of the Westinghouse WWER-1000 fuel design and how test results and operational experiences from multiple sources have been utilized to produce a most suitable fuel design. Early in 2011 a full region of the Westinghouse WWER-1000 design completed another full cycle of operation at South Ukraine Unit 3, all with excellent results. All 42 fuel assemblies were examined for visible damage or non

  15. Determination of fast neutron fluence at WWER-1000 pressure vessel

    International Nuclear Information System (INIS)

    Valenta, V. et al.

    1989-01-01

    The influence function method is an effective tool making it possible, by means of tabulated values to rapidly perform three-dimensional calculations of fast neutron fluences for various reactor core loadings and for various nuclear power plant units. The procedure for determining the spatial dependence of the fast neutron fluences in a WWER-1000 pressure vessel is described. For this, the reactor core is divided into sufficiently fine volume elements within which the neutron source can be regarded as coordinate-independent. The influence functions point to a substantial role of sources lying at the reactor core periphery. In WWER-1000 reactors, only 1 or 2 rows of peripheral assemblies are important. The influence function method makes possible a rapid and easy determination of preconditions for the assessment of the residual lifetime of the pressure vessel based on the actual reactor core loadings. (Z.M.). 7 figs., 8 refs

  16. Modelling of WWER-1000 fuel: state and prospects

    International Nuclear Information System (INIS)

    Medvedev, A.; Bibilashvili, Yu.; Bogatyr, S.; Khvostov, G.

    1994-01-01

    The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs

  17. Operational indices of WWER-1000 fuel assemblies and their improvements

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Demin, E.

    1994-01-01

    The most general design features of WWER-1000 fuel assembly are discussed. The following advantages of design are stated as well as their operational confirmation and occurrences: 1) 'packing' density (tight-lattice) of fuel rods within the fuel assemblies; 2) simple handling of fuel assemblies and its small vulnerability; 3) good conditions for coolant mixing; 4) protection of the absorber rods against coolant effect; 5) adaptability to manufacture that provides stable quality. The main operational indices gathered during a ten-year period (1982-1992) at 17 WWER-1000 units in Russia and Ukraine are outlined. Provisions for emergency protection reliability are described. Future directions to improve fuel economy and control rod operability are discussed. 1 fig

  18. Modelling of WWER-1000 fuel: state and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Medvedev, A; Bibilashvili, Yu; Bogatyr, S; Khvostov, G [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    The role of START-3 code in studying and computerized modelling of post-irradiation behaviour of standard fuel rods in real operation conditions of WWER-1000 reactors is described. The models used in the code are based on experimental study of material properties, processes and post irradiation research on standard and experimental fuel pins. The code capability is verified by comparison with data from experiments on WWER test rods performed in MR reactor, the Russia-Finland tests SOFIT and the international program FUMEX. The comparison performed and the results thus obtained demonstrate the satisfactory ability of START-3 code to simulate fuel rod behaviour in normal operation condition. The calculations confirm the experimentally observed evidence of an essential margin on serviceability of WWER-1000 fuel pin with three year operation cycle permitting an increase in design fuel burnup. 2 tabs., 18 figs.

  19. Operational indices of WWER-1000 fuel assemblies and their improvements

    Energy Technology Data Exchange (ETDEWEB)

    Vasilchenko, I; Demin, E [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1994-12-31

    The most general design features of WWER-1000 fuel assembly are discussed. The following advantages of design are stated as well as their operational confirmation and occurrences: (1) `packing` density (tight-lattice) of fuel rods within the fuel assemblies; (2) simple handling of fuel assemblies and its small vulnerability; (3) good conditions for coolant mixing; (4) protection of the absorber rods against coolant effect; (5) adaptability to manufacture that provides stable quality. The main operational indices gathered during a ten-year period (1982-1992) at 17 WWER-1000 units in Russia and Ukraine are outlined. Provisions for emergency protection reliability are described. Future directions to improve fuel economy and control rod operability are discussed. 1 fig.

  20. Safety of NPP with WWER-440 and WWER-1000 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E [Energoproekt, Sofia (Bulgaria); Gledachev, J; Angelov, D [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab.

  1. Safety of NPP with WWER-440 and WWER-1000 reactors

    International Nuclear Information System (INIS)

    Balabanov, E.; Gledachev, J.; Angelov, D.

    1995-01-01

    The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab

  2. Experience and prospects of WWER-1000 reactor spent fuel transport

    International Nuclear Information System (INIS)

    Kondratyev, A.N.; Yershov, V.N.; Kozlov, Yu.V.; Kosarev, Yu.A.; Ilyin, Yu.V.; Pavlov, M.S.

    1989-01-01

    The paper deals with the USSR experience in shipping the commercial WWER-1000 reactor spent fuel in TK-10 and TK-13 casks. The cask designs, their basic characteristics and the WWER-1000 spent fuel features are described. An example of calculational/experimental approach in the design of a basket (one of the most important components) for spent fuel assembly (SFA) accommodation in a cask is given. The main problems of future development works are presented in brief. A concept of development of nuclear power industry with the closed fuel cycle is assumed in the Soviet Union, hence the spent nuclear fuel is to be transported from NPPs to reprocessing plants. To transport WWER-1000 spent fuel, the casks of two types were developed. These are: a pilot TK-10 cask of 3t capacity in fuel; a commercial TK-13 cask of ∼6t capacity in fuel. The pilot TK-10 cask is thick-walled (360mm) cylindrical vessel manufactured of steel shells and a bottom welded to each other. The material of the body is carbon steel. There is a steel jacket on the outer side of the cask body and at 120 mm distance off the bottom. On its cylindrical part between the jacket and the body there are T-shaped circular ribs acting as shock-absorbers. The space between the jacket and the body is filled with ethylene glycol solution of 65 degree C crystallization temperature, which functions as a neutron shielding. The TK-10 cask coolant is water or air (nitrogen) at minor excess pressure resulted from FA heatup after the cask sealing

  3. WWER-1000 fuel cycles: current situation and outlook

    International Nuclear Information System (INIS)

    Kosourov, E.; Pavlov, V.; Pavlovichev, A.; Spirkin, E.; Shcherenko, A.

    2013-01-01

    Usage mode of nuclear fuel in WWER type reactor has been changed significantly till the moment of the first WWER-1000 commissioning. There are a lot of improvements, having an impact on the fuel cycle, have been implemented for units with WWER-1000. FA design and its constructional materials, FA fuel weight, burnable poison, usage mode of units and etc have been modified. As the result of development it has been designed a modern FA with rigid skeleton. As a whole it allows to use more efficient configurations of the core, to extend range of fuel cycle lengths and to provide good flexibility in the operation. In recent years there were in progress works on increasing FA uranium capacity. As the result there were developed two designs of the fuel rod: 1) the fuel column height of 3680 mm, diameters of the fuel pellet and its central hole of 7.6 and 1.2 mm respectively and 2) the fuel column height of 3530 mm, the fuel pellet diameter of 7.8 mm without the central hole. Such fuel rods have operating experience as a part of different FA designs. Positive operating experience was a base of new FA (TVS-4) development with the fuel column height of 3680 mm and the fuel pellet diameter of 7.8 mm without the central hole. The paper presents the overview of WWER-1000, AES-2006 and WWER-TOI fuel cycles based on FAs with fuel rod designs described above. There are demonstrated fuel cycle possibilities and its technical and economic characteristics. There are discussed problems of further fuel cycle improvements (fuel enrichment increase above 5 %, use of erbium as alternative burnable poison) and their impact on neutronics characteristics. (authors)

  4. The summary of WWER-1000 fuel utilization in Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Afanasyev, A [Ukrainian State Committee on Nuclear Power Utilization, Kiev (Ukraine)

    1997-12-01

    The report discusses the status of the fuel and fuel cycles of WWER-1000 reactors in Ukraine. The major reasons that caused the Ukrainian utilities to overcome the conservative design solutions in order to improve fuel utilization and extend fuel burnup are shown. At the same time the sufficient fuel reliability and fuel cycle flexibility are ensured. The burnup distribution in the unloaded fuel assemblies and average fuel rod failure rate are presented. The questions of reactor core operation safety and the economical problems of the front end of the fuel cycle are also considered. (author). 2 refs, 3 figs, 4 tabs.

  5. Control rod drive WWER 1000 – tuning of input parameters

    Directory of Open Access Journals (Sweden)

    Markov P.

    2007-10-01

    Full Text Available The article picks up on the contributions presented at the conferences Computational Mechanics 2005 and 2006, in which a calculational model of an upgraded control rod linear stepping drive for the reactors WWER 1000 (LKP-M/3 was described and results of analysis of dynamical response of its individual parts when moving up- and downwards were included. The contribution deals with the tuning of input parameters of the 3rd generation drive with the objective of reaching its running as smooth as possible so as to get a minimum wear of its parts as a result and hence to achieve maximum life-time.

  6. Neutron physics calculation for WWER-1000 absorber element lifetime determination

    International Nuclear Information System (INIS)

    Kurakin, K.Yu.; Kushmanov, S.A.

    2009-01-01

    Absorber element with compound absorber has been operating in WWER-1000 power units since 1995. AE design meets operating organizations requirements for reliability, service life (to 10 years) and safety functions. Extension of AE service life up to 20 - 30 years by the complex of calculation and experimental work is an important problem of WWER new designs development. The paper deals with the issues related to calculation determination of main factors that influence AE service life limitation - neutron flux and fluence onto absorbing and structural materials during extended service life. (Authors)

  7. Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

  8. Evaluation of WWER-1000 vessel materials fracture toughness

    International Nuclear Information System (INIS)

    Grinik, Eh.U.; Revka, V.N.; Chirko, L.I.; Chajkovskij, Yu.V.

    2007-01-01

    The lifetime of WWER-1000-type reactor vessels is finally conditioned by the fracture toughness (crack growth resistance) of RPV materials. Up to now in line with the regulations the fracture toughness is characterized by the critical temperature of brittleness determined by the results of the Charpy specimen impact testing. Such approach is typical for all countries operating the water pressure reactors. However, regulatory approach is known from the western specialists not always to characterize adequately the crack growth resistance of the vessel materials and in some cases to underestimate their characteristics in the reference state that leads to unreasonably high conservatism. Excessive conservatism may lead to the invalid restrictions in the operating modes and the service life of the reactor vessel. Therefore there appeared the necessity to apply another approaches based on the state-of-the-art experimental methods of the fracture mechanics and allowing evaluating the fracture toughness parameters sufficiently. The paper presents the results of the comparison of the regulatory approach and the Master curve approach from the point of view of the adequate determination of the vessel material crack growth resistance parameters. Analysis of the experimental data of the surveillance specimens illustrated the potential possibility of applying the new statistical method for the WWER-1000- type reactor vessel lifetime extension

  9. Nuclear power plants with reactors WWER-1000 type: today and tomorrow; AEhS s WWER-1000: nastoyashchee i budushchee

    Energy Technology Data Exchange (ETDEWEB)

    Molchanov, V; Biryukov, G; Novak, K [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1996-12-31

    There are currently 19 NPP units based on WWER-1000 reactors working in Russia, Ukraine and Bulgaria. They are of four types: V-187, V-302, V-338, V-320. The design principles of these reactors comply with regulations of the eighties, and it is necessary to introduce improvements according to the new regulations and to the operation experience gained. Two approaches for safety and efficiency enhancement are described: AS-91 and AS-92. AS-91 implies gradual improvement of the base WWER-1000/V-320 design by incorporation of new design solutions avoiding the need of building large scale models. AS-92 refers to entirely new design which require experimental research by building a full scale models or by using natural stands. The latter approach will be used for NPP projects to be built after year 2000. The main new feature of AS-92 is the addition of passive safety systems to the active ones in order to protect the fuel from damage.

  10. Consistent Code Qualification Process and Application to WWER-1000 NPP

    International Nuclear Information System (INIS)

    Berthon, A.; Petruzzi, A.; Giannotti, W.; D'Auria, F.; Reventos, F.

    2006-01-01

    Calculation analysis by application of the system codes are performed to evaluate the NPP or the facility behavior during a postulated transient or to evaluate the code capability. The calculation analysis constitutes a process that involves the code itself, the data of the reference plant, the data about the transient, the nodalization, and the user. All these elements affect one each other and affect the results. A major issue in the use of mathematical model is constituted by the model capability to reproduce the plant or facility behavior under steady state and transient conditions. These aspects constitute two main checks that must be satisfied during the qualification process. The first of them is related to the realization of a scheme of the reference plant; the second one is related to the capability to reproduce the transient behavior. The aim of this paper is to describe the UMAE (Uncertainty Method based on Accuracy Extrapolation) methodology developed at University of Pisa for qualifying a nodalization and analysing the calculated results and to perform the uncertainty evaluation of the system code by the CIAU code (Code with the capability of Internal Assessment of Uncertainty). The activity consists with the re-analysis of the Experiment BL-44 (SBLOCA) performed in the LOBI facility and the analysis of a Kv-scaling calculation of the WWER-1000 NPP nodalization taking as reference the test BL-44. Relap5/Mod3.3 has been used as thermal-hydraulic system code and the standard procedure adopted at University of Pisa has been applied to show the capability of the code to predict the significant aspects of the transient and to obtain a qualified nodalization of the WWER-1000 through a systematic qualitative and quantitative accuracy evaluation. The qualitative accuracy evaluation is based on the selection of Relevant Thermal-hydraulic Aspects (RTAs) and is a prerequisite to the application of the Fast Fourier Transform Based Method (FFTBM) which quantifies

  11. Methodology of thermalhydraulic tests of fuel assemblies for WWER-1000

    International Nuclear Information System (INIS)

    Archipov, A.; Kolochko, V.N.

    2001-01-01

    At present 11 units with WWER-1000 are in operation in Ukraine. The NPPs are provided with nuclear fuel from Russia. The fuel assemblies are fabricated and delivered to Ukrainian NPPs from Russia. However the contemporary tendencies of nuclear energy development in the world assume a diversification of nuclear fuel vendors. Therefore the creation of the own nuclear fuel cycle of Ukraine is in mind in the strategy of nuclear energy development of Ukraine. As a part of the fuel assemblies fabrication process complex of the thermalhydraulic tests should be carried out to confirm design characteristics of the fuel assemblies before they are loaded in the reactor facility. The experimental basis and scientific infrastructure for the thermalhydraulic tests arrangement and realization of the programs and procedures for the core equipment examination are under consideration. (author)

  12. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Höglund, J.; Jansson, A.; Latorre, R.; Davis, D.

    2015-01-01

    In 2005, six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in South Ukraine Unit 3 (SU3). The LTAs completed the planned four cycles of operation and reached an average assembly burnup in excess of 43 MWd/ kgU. Post Irradiation Examination (PIE) inspections were performed after completion of each cycle and it was concluded that the 6 Westinghouse LTAs performed as expected during their operational regimes. In 2010, a full region of 42 assemblies of an enhanced WWER-1000 fuel design for Ukrainian reactors, designated WFA, was loaded in SU3. The WFA includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. In 2015, 26 WFAs completed their planned four cycles of operation reaching an average assembly burnup in excess of 42 MWd/ kgU. Currently 36 WFAs continue operating their fourth cycle in SU3. In addition, South Ukraine Unit 2 (SU2) has been loaded with WFAs and 27 assemblies have completed two cycles of operation reaching an average assembly burnup above 24 MWd/kgU. PIE for the WFAs has been completed after each cycle of operation. All assemblies have been examined for visible damage or non-standard position of fuel assembly components during unloading and reloading. All WFAs have also been subject to the standard leak testing process, with all fuel rods found to be hermetically sealed and non-leaking. Each outage, six WFAs have been subject to a more extensive inspection program. In 2012, 2013, and 2015, the Westinghouse Fuel Inspection and Repair Equipment (FIRE) workstation were used for the SU3 inspections. Excellent irradiation fuel performance has been observed and measured on all WFAs. The fuel assembly growth, rod cluster control assembly (RCCA) drag forces, oxide thickness, total fuel rod-to-nozzle gap channel closure, and fuel assembly bow data were within the bounds of the Westinghouse experience database. Results and concluding remarks from the PIEs are provided in this paper. In

  13. Analysis of WWER 1000 SG cold collector cracking

    International Nuclear Information System (INIS)

    Matocha, K.; Wozniak, J.

    2000-01-01

    Following the recommendations of the 1993 consultants' meeting on 'Steam Generator Collector Integrity of WWER 1000 Reactors', an extensive experimental program was started with the aim of finding the dominant damage mechanism responsible for cold collector cracking in steam generators, and of determining whether proper operating conditions can make the operation of VITKOVICE-produced steam generators safe throughout their lifetime. The experiments consisted of: a study of the effect of strain and thermal ageing and dissolved oxygen content on subcritical crack growth in 10GN2MFA steel; a study of the effect of high temperature water and tube expansion technology on the fracture behaviour of ligaments between holes for heat exchange tubes; a study of the effect of drilling, tube expansion technology and heat treatment on residual stresses on the surface of holes for heat exchange tubes. Details of the experimental techniques used are given as well as a discussion of the results obtained and presented in tables and graphs. (A.K.)

  14. Upgrading of WWER-1000 NPP safety on spent fuel transportation

    International Nuclear Information System (INIS)

    Kostarev, V.; Shchukin, A.; Petrenya, Yu.; Nikitin, V.; Romanovskij-Romanko, A.; Shevchenko, V.

    2003-01-01

    Transportation process for the WWER-1000 spent fuel assemblies consists of three main steps: (i) lifting of unloaded cask on the elevation of +38.05 m; (ii) loading of spent fuel assemblies into the cask; (iii) loaded cask lowering to the conveyer located in the transport corridor on the elevation 0.00 m. The most hazardous situation within described process for the cask itself and reactor building structures is an accidental drop of the cask from the height of 38.05 m to the transport corridor floor due to failure of traverse or crane's cable break. According to international practice and standards' requirements the cask shall be designed for the drop from 9 meters height to a rigid plate. However, preliminary analyses have shown that in case of 38 m drop the value of g-loads are several times larger than allowable limits. Additionally, strength capacity of the foundation slab of the reactor building is not guaranteed. Using of special damping device that is capable to bring dynamic loads to allowable limits could mitigate the catastrophic consequences of cask's 38.05 meters drop. The paper presents a basic design of the special damping platform and discusses results of analyses of different modes of cask drops and efficiency of the proposed solution. (author)

  15. Fuel assembly leak tightness control on WWER-1000 reactor

    International Nuclear Information System (INIS)

    Ivanova, R.; Gerchev, N.; Mateev, A.

    2001-01-01

    The main index for integrity of the fuel rods cladding is the specific activity value of the primary coolant. This value determines the safe operation of the reactor. The limit for safe operation of WWER-1000 reactor is the value of the total activity of Iodine isotopes in the primary coolant 5.0x10 -3 Ci/l. The paper briefly describes the methodology for performing a fuel tightness test (sipping test) and shows the results from these tests performed during the period 1987 -1999 in units 5 and 6 at the Kozloduy NPP. An additional index related to the safe operation is defined to characterize the fuel cladding integrity Fuel Reliability Index (FRI). The FRI is defined as value of the average activity of 131 I in the primary coolant, corrected with a part of precipitated 235 U migration and fixed to the general permanent purification frequency. Two criteria (quantitative and statistic) are determined to qualify the fuel cladding integrity. The results from sipping tests show good reliability of the fuel irradiated in unit 5 and 6 at the Kozloduy NPP

  16. Dynamic analysis of WWER-1000 nuclear power plants

    International Nuclear Information System (INIS)

    Asfura, A.P.; Jordanov, M.J.

    1995-01-01

    As part of the effort to assess the seismic vulnerability of nuclear power plants in Eastern Europe, a series of dynamic analyses have been carried out for several plants. These analyses were performed using modern analysis techniques, current local seismic parameters, and local soil profiles. This paper presents a compilation of some of the seismic analyses performed for the WWER-1000 reactor buildings at the nuclear power plants of Belene and Kozloduy in Bulgaria, and Temelin in the Czech Republic. The reactor buildings at these three plants are practically identical and correspond to the standard building design for this type of reactors. The series of analyses performed for these buildings encompasses various soil profiles, seismic ground motions, and different soil-structure interaction analysis techniques and modelling. The analysis of a common structure under different conditions gives the opportunity to assess the relative importance that each of the analysis elements has in the structural responses. The use of different SSI computer programs and foundation modeling was studied for Kozloduy, and the effects of different soil conditions and site-specific seismicity were studied by comparing the responses for the three plants. In-structure acceleration response spectra were selected as the structural responses for comparison purposes

  17. Modal analysis of spent fuel cask for WWER-1000 reactors

    International Nuclear Information System (INIS)

    Azimfar, S. A.; Kazemi, A.

    2011-01-01

    The Spent Fuel Assemblies of WWER-1000 reactors are planned to be transported by special containers which are supposed to be designed in a manner to stand against vibrations and impacts in order to protect the spent fuel from any possible damage. The vibration opposition of these containers shall be far beyond the critical resonance, because the resonances about the natural frequency of the structure will cause the enhancement of its oscillation range and may end with its disintegration. Determination of the amounts of natural frequencies and their mode shape can be achieved by vibration analyzing methods. The amount of the natural frequency of any structure crucially depends on its shape, material and lean points as well as the amount of the loads and the type of these loads. Due to the fact that the Spent Fuel Casks used for transportation in nuclear power plants in Russian Federation are TK-13 type and the pieces of information released are negligible, the scientists in Russia are working on the design and analysis of a new type made up of composite Material. In the presented paper the cask of spent fuel of TK-13 is modeled by ANSYS at 10.0 and ten natural frequency modes have been calculated, followed by the comparison of this result with the composite cask.

  18. Accidents during shutdown conditions for NPP with WWER-1000/428

    International Nuclear Information System (INIS)

    Antropov, H.A.

    1996-01-01

    This paper considers the effect of such internal initiating events upon safety of the NPP with WWER-1000/428 as boron dilution and loss of residual heat removal at the reduced coolant inventory in the primary circuit. 2 refs

  19. Solution of the 'MIDICORE' WWER-1000 core periphery power distribution benchmark by KARATE and MCNP

    International Nuclear Information System (INIS)

    Temesvari, E.; Hegyi, G.; Hordosy, G.; Maraczy, C.

    2011-01-01

    The 'MIDICORE' WWER-1000 core periphery power distribution benchmark was proposed by Mr. Mikolas on the twentieth Symposium of AER in Finland in 2010. This MIDICORE benchmark is a two-dimensional calculation benchmark based on the WWER-1000 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of the benchmark is to test the pin by pin power distribution in selected fuel assemblies at the periphery of the WWER-1000 core. In this paper we present our results (k eff , integral fission power) calculated by MCNP and the KARATE code system in KFKI-AEKI and the comparison to the preliminary reference Monte Carlo calculation results made by NRI, Rez. (Authors)

  20. An investigation of axial xenon stability in WWER-1000 reactor designs

    International Nuclear Information System (INIS)

    Doshi, P.K.; Miller, R.W.

    1993-01-01

    The nuclear power plants of the WWER-1000 design have experienced frequent xenon oscillation control problems. In most PWRs, xenon oscillations are largely a problem in the axial direction. An one dimensional core model representative of the WWER-1000 design was set up to examine the controllability of the current design. An investigation of possible improvements to this design was made. There was no indication that xenon oscillations were an inherent problem in WWER-1000 core design. Simple changes to the control rod system coupled with a sound power distribution control strategy that has been proven to be an effective but simple procedure to follow, eliminate xenon control problems. The changes proposed can be implemented in a very cost effective manner. There are no equipment changes needed, existing control rods can be used. Only software changes are required. (Z.S.) 1 tab., 2 figs., 7 refs

  1. Improvement of operational performance and increase of safety of WWER-1000/V-392

    International Nuclear Information System (INIS)

    Kurakov, Y.A.; Dragunov, Y.G.; Podshibiakin, A.K.; Fil, N.S.; Krushelnitsky, V.N.; Berkovich, V.M.

    2001-01-01

    The national programme of nuclear power development approved by the Russian Federation Government in 1998 considers the design of WWER-1000/V-392 power unit as a priority project of the new generation NPP with improved operational performances and increased safety. The pilot unit of this design (NVAES-2) is licensed for construction at the Novovoronezh NPP site. The NVAES-2 design is developed on the basis of standard power unit with reactor plant V-320. Twenty units of this type are in operation at the nuclear power plants in Russia, Ukraine and Bulgaria having totally about 270 reactor-years of operation. Two more V-320 units are being commissioned this year at Rostov NPP and Temelin NPP. So, the WWER-1000/V-392 design is as a whole an evolutionary development of the operating standard unit WWER-1000/V-320. Many technical solutions aimed at increase of safety and improvement of operational performance of the plant are implemented in the NVAES-2 design, such as advanced reactor WWER-1000, passive system of residual power removal, passive system of the core flooding under loss-of-coolant accidents, and others. NVAES-2 design refers to a class of advanced light water reactors and corresponds to the international requirements imposed to the nuclear power plants to be put into operation after the year 2000. New V-392 power unit has a good perspective from the view point of extensive implementation in the framework of the nuclear electricity production in Russia. Design decisions on NVAES-2 power unit with WWER-1000/V-392 reactor plant which assure significantly higher safety level and improve economical performance as compared to the operating WWER-1000 units are briefly considered in the present paper. (author)

  2. Systems required during and after an earthquake. Summary report. WWER-1000 nuclear power plants

    International Nuclear Information System (INIS)

    Monette, P.

    1995-01-01

    The scope of this document is to list the mechanical, instrumentation and electrical components required during and after earthquake, in order to achieve and maintain safe shutdown conditions of a WWER-1000 type nuclear power plant. The main objective pursued in establishing the systems and equipment list is to provide guidance for the design and implementation of the backfits which are necessary to increase seismic resistance of the components required after earthquake. The presented list is established on generic basis, i.e. it is applicable to any specific WWER-1000

  3. ACT-1000. Group activation cross-section library for WWER-1000 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zolotarev, K I; Pashchenko, A B [National Research Centre - A.I. Leipunsky Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-10-01

    The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10{sup -9}-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (author)

  4. Electroerosion cutting of low-sized templets from WWER-1000 type reactor vessel

    International Nuclear Information System (INIS)

    Neklyudov, I.M.; Ozhigov, L.S.; Gozhenko, S.V.

    2012-01-01

    The article presents the results of developed method of electroerosion cutting of low-sized templets for the reactor vessel metal composition and structure control in laboratory environment. The article describes the equipment for the remote electroerosive cutting of templets from WWER-1000 type reactor vessel by rigid electrode. The testing results are also shown.

  5. Steam generator and condenser design of WWER-1000 type of nuclear power plant

    International Nuclear Information System (INIS)

    Zare Shahneh, Abolghasem.

    1995-03-01

    Design process of steam generator and condenser at Russian nuclear power plant type WWER-1000 is identified. The four chapter of the books are organized as nuclear power plant, types of steam generators specially horizontal steam generator, process of steam generator design and the description of condenser and its process design

  6. The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code

    International Nuclear Information System (INIS)

    Hordosy, G.; Maraczy, Cs.

    2000-01-01

    Preparations for disposition of weapons grade plutonium in WWER-1000 reactors are in progress. Benchmark: Defined by the Kurchatov Institute (S. Bychkov, M. Kalugin, A. Lazarenko) to assess the applicability of computer codes for weapons grade MOX assembly calculations. Framework: 'Task force on reactor-based plutonium disposition' of OECD Nuclear Energy Agency. (Authors)

  7. Technological problems connected with execution of the protection sheets for nuclear power sets WWER-1000

    International Nuclear Information System (INIS)

    Hajutin, J.G.; Kriczewskij, A.Z.

    1977-01-01

    The choice of the structure and the prestressing system of the R.C. protection sheet for nuclear power sets WWER-1000 is motivated. The technological problems arised during the execution stage, as well as the technological line producing the tendons to prestress the structure by up winding are presented. (author)

  8. Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-07

    At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

  9. Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

    International Nuclear Information System (INIS)

    1994-01-01

    At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

  10. Substantiation of operation limits of reactivity insertion during WWER-1000 reactors start-up; Obosnovanie ehkspluatatsionnykh predelov vvoda reaktivnosti pri puske reaktorov WWER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Boev, I; Sabitov, A; Sal` kov, V; Sudarev, O; Yakovlev, A [ATOMTECHENERGO RF, Novovoronezh (Russian Federation)

    1996-12-31

    The methods and programmes used to define the tolerable rate of reactivity insertion during WWER-1000 start-up are presented. They include calculation of the neutron source power in the core during the sub-critical stage and calculation of the relative neutron density and reactor period during the critical stage. The need for optimisation and regulation of tolerable rates is discussed along with the tool parameters affecting the reactivity during start-up. The possibility of increasing the feed rate of pure condensate into the first loop during the time needed to reach critical stage is justified. 4 refs., 3 tabs.

  11. Methodology for the evaluation of tolerability of defects in WWER-1000/V-320 reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Horacek, L.; Ruscak, M.

    1996-05-01

    The methodology provides guidelines for the assessment of tolerability of defects found during in-service inspection of the base material and overlay of WWER-1000/V-320 type reactor pressure vessels. With regard to the method of calculating the tolerability of defects and rules for the preparation and implementation of repairs, this methodology can also find use in the assessment of tolerability of defects in selected facilities of WWER-1000/V-320 type nuclear power plants provided that adequate input data concerning the materials, manufacturing technology, and operating load regime are available and that the facilities are made of ferrite/bainite type steels. This methodology should serve as a binding document underlying the development of a technical approach to provisions for a further operation of facilities in which intolerable defects have been found by nondestructive testing. (author)

  12. Stainless steel corrosion in conditions simulating WWER-1000 primary coolant. Corrosion behaviour in mixed core

    International Nuclear Information System (INIS)

    Krasnorutskij, V.S.; Petel'guzov, I.A.; Gritsina, V.M.; Zuek, V.A.; Tret'yakov, M.V.; Rud', R.A.; Svichkar', N.V.; Slabospitskaya, E.A.; Ishchenko, N.I.

    2011-01-01

    Research into corrosion kinetics of austenitic stainless steels (06Cr18Ni10Ti, 08Cr18Ni10Ti, 12Cr18Ni10Ti) in medium which corresponds to composition and parameters of WWER-1000 primary coolant with different pH values in autoclave out-pile conditions during 14000 hours is given. Surface of oxide films on stainless steels is investigated. Visual inspection of Westinghouse and TVEL fuel was carried out after 4 cycles in WWER-1000 primary water chemistry conditions at South Ukraine NPP. Westinghouse and TVEL fuel cladding materials possess high corrosion resistance. Blushing of weldments was observed. No visual corrosion defects or deposits were observed on fuel rods.

  13. Vibrodynamical tests of RP equipment with application of imitation area of WWER-1000 reactor

    International Nuclear Information System (INIS)

    Khajretdinov, V.U.; Tarkhanov, V.V.; Rodionova, I.N.

    2015-01-01

    Performance of preoperational tests and measurements with application of imitation area of the reactor is a distinctive characteristic of putting into operation of NPP Units with WWER-1000/1200. The imitation area consists of 163 full-scale FA models, where fuel matrixes made of nuclear-fissionable material, are replaced by leaden simulators. Vibrodynamic tests involve inspection of hydrodynamic disturbances in the primary circuit (dynamic impact on the inspected elements), characteristics of vibration response of the main equipment stress-deformed state of bearing structure, and also parameters of moving and geometry of the inspected objects (boundary conditions at process simulation). Preoperational tests and measurements on the simulated area of WWER-1000/1200 are obligatory and performed at every unit of NPP of this type [ru

  14. Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kiselyov, V.A.; Sokov, L.M.

    1997-04-01

    The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.

  15. Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

    International Nuclear Information System (INIS)

    Kiselyov, V.A.; Sokov, L.M.

    1997-01-01

    The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor

  16. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E.; Ivanova, A. [Energoproekt, Sofia (Bulgaria)

    1995-12-31

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment.

  17. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E; Ivanova, A [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment.

  18. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    International Nuclear Information System (INIS)

    Balabanov, E.; Ivanova, A.

    1995-01-01

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment

  19. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L. [Nuclear Research Institute Rez (Czech Republic)

    1997-04-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors.

  20. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    International Nuclear Information System (INIS)

    Zdarek, J.; Pecinka, L.

    1997-01-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors

  1. Best estimate modeling of fuel thermomechanical behaviour in WWER 1000 LB LOCA

    International Nuclear Information System (INIS)

    Valach, M.; Klouzal, J.; Zymak, J.; Dostal, M.

    2009-01-01

    The paper summarizes our calculations of the performance of the WWER 1000 NPP fuel rods during postulated LB LOCA. The thermomechanical modeling was performed by FRAPTRAN using the FRACAS-I mechanical model using the boundary conditions calculated by the ATHLET code. The results and their statistical evaluation are presented, the process of the generalization of gained insight into the best-estimate thermal-hydraulic analyses (BE TM) predictions in order to define a generic BE TM methodology is outlined (authors)

  2. Experimental study of the passive flooding system in the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Malyshev, A.B.; Efanov, A.D.; Kalyakin, S.G.

    2002-01-01

    The design solution of the passive flooding system in the WWER-1000 reactor core with the V-392 reactor facility and the scheme of the GE-2 large-scale thermohydraulic stand for substantiation of its functions are presented. The proposals, improving the efficiency of the system are developed on the basis of the experimental studies on the equipment input-output operational characteristics and the recommendations on the substantiation of the function of the reactor core flooding system are given [ru

  3. Status and prospects of activities on algorithms and methods in WWER-1000 core control

    International Nuclear Information System (INIS)

    Filimonov, P.; Krainov, Y.; Proselkov, V.

    1994-01-01

    On the basis of long-term operational experience and investigations the problems of WWER-1000 reactor control are discussed. Such control is needed for WWER-1000, as well as for its Western analog PWR, for suppressing the axially instable power density field resulted from non-equilibrium redistribution of Xe-135 nuclei in the reactor core. It has been found that an adequate assessment of the reactor state and the prediction of its response to various control actions is essential for the control of power density distribution. For this purpose a computerized operator's adviser with a reactor simulator realizing a physical reactor model based on BIPR-7 code is used. The operation experience of WWER-1000 shows that the available control algorithms allow, with a fair degree of assurance, the prevention of intensive xenon oscillations and the stabilization of the axial offset. But in connection with the renunciation of half-length control rods a new algorithm is under development which makes use of full-length control rods for suppressing the intensive xenon oscillations in the descending phase. A new method based on BIPR-7 and PERMAK codes is also being developed for estimating the value and rate of linear power rating change of the fuel elements in power cycling. 12 figs., 7 refs

  4. Status and prospects of activities on algorithms and methods in WWER-1000 core control

    Energy Technology Data Exchange (ETDEWEB)

    Filimonov, P; Krainov, Y; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    On the basis of long-term operational experience and investigations the problems of WWER-1000 reactor control are discussed. Such control is needed for WWER-1000, as well as for its Western analog PWR, for suppressing the axially instable power density field resulted from non-equilibrium redistribution of Xe-135 nuclei in the reactor core. It has been found that an adequate assessment of the reactor state and the prediction of its response to various control actions is essential for the control of power density distribution. For this purpose a computerized operator`s adviser with a reactor simulator realizing a physical reactor model based on BIPR-7 code is used. The operation experience of WWER-1000 shows that the available control algorithms allow, with a fair degree of assurance, the prevention of intensive xenon oscillations and the stabilization of the axial offset. But in connection with the renunciation of half-length control rods a new algorithm is under development which makes use of full-length control rods for suppressing the intensive xenon oscillations in the descending phase. A new method based on BIPR-7 and PERMAK codes is also being developed for estimating the value and rate of linear power rating change of the fuel elements in power cycling. 12 figs., 7 refs.

  5. A procedure for temperature-stress fields calculation of WWER-1000 primary circuit in PTS event

    Energy Technology Data Exchange (ETDEWEB)

    Petkov, G [Technical Univ., Dept. Thermal and Nuclear Power Engineering, Sofia (Bulgaria); Groudev, P; Argirov, J [Bulgarian Academy of Science, Inst. for Nuclear Research and Nuclear Energy, Sofia (Bulgaria)

    1997-09-01

    The paper presents the procedure of an investigation of WWER-1000 primary circuit temperature-stress field by the use of thermohydraulic computation data for a pressurized thermal shock event ``Core overcooling``. The procedure is based on a model of the plane stress state with ideal contact between wall and medium for the calculation. The computation data are calculated on the base of WWER-1000 thermohydraulic model by the RELAP5/MOD3 codes. This model was developed jointly by the Bulgarian and BNL/USA staff to provide an analytical tool for performing safety analysis. As a result of calculations by codes the computation data for temperature field law (linear laws of a few distinguished parts) and pressure of coolant at points on inner surface of WWER-1000 primary circuit equipment are received. Such calculations can be used as a base for determination of all-important load-carrying sections of the primary circuit pipes and vessels, which need further consideration. (author). 7 refs, 2 figs, 2 tabs.

  6. Steam generator collector integrity of WWER-1000 reactors. IAEA extrabudgetary programme on the safety of WWER NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C.; Strupczewski, A. [International Atomic Energy Agency, Vienna (Austria)

    1995-12-31

    At the Consultants` Meeting on `The Safety of WWER-1000 Model 320 Nuclear Power Plants` organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of WWER-1000 steam generator integrity was identified as an important issue of safety concern. Considering the safety importance of this issue, a Consultants` Meeting on `The Steam Generator Integrity of WWER-1000 Nuclear Power Plants` was convened in Vienna in May 1993, attended by 15 international experts in the area to compile information on the steam generator operating experience, deficiencies and corrective measures implemented and planned. In order to also include information from the main designer OKB Gidropress and to finalize the meeting report the IAEA convened a second meeting on the issue on 23-27 November 1993. The present paper summarizes the information and conclusions from those meetings.

  7. Steam generator collector integrity of WWER-1000 reactors. IAEA extrabudgetary programme on the safety of WWER NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C; Strupczewski, A [International Atomic Energy Agency, Vienna (Austria)

    1996-12-31

    At the Consultants` Meeting on `The Safety of WWER-1000 Model 320 Nuclear Power Plants` organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of WWER-1000 steam generator integrity was identified as an important issue of safety concern. Considering the safety importance of this issue, a Consultants` Meeting on `The Steam Generator Integrity of WWER-1000 Nuclear Power Plants` was convened in Vienna in May 1993, attended by 15 international experts in the area to compile information on the steam generator operating experience, deficiencies and corrective measures implemented and planned. In order to also include information from the main designer OKB Gidropress and to finalize the meeting report the IAEA convened a second meeting on the issue on 23-27 November 1993. The present paper summarizes the information and conclusions from those meetings.

  8. Experience of CR and RCCA operation in Ukrainian WWER-1000: Aspects of reliability, safety and economic efficiency

    International Nuclear Information System (INIS)

    Afanasyev, A.

    2000-01-01

    The next topics are represented in the paper: A brief history of WWER-1000 control rod (CR) and WWER-1000 rod cluster control assembly (RCCA) design; Evolution of WWER-1000 CR manufacturing technology and design; Experience of RCCA operation; Lifetime extension of WWER-1000 boron carbide CR; WWER-1000 reactor core operation problems due to partial RCCA insertion; Designing and licensing procedures and first operational experience of WWER-1000 RCCA (CR) with a combined absorber 'boron carbide-hafnium' and a chromium-nickel alloy cladding. The main conclusions are: Fuel assembly (FA) bow is the main reason of partial RCCA insertion during reactor core operation. However, the use of the RCCA and its driver bar with increased dead load, alongside with other measures, allow to reduce the probability of incomplete RCCA insertion; The materials used in CRs of RCCA in existing reactor operating modes have been working reliably; The use of hafnium under an appropriate price policy can give certain economic advantages for the Ukrainian NPPs, however, additional research is needed in order to confirm the specific CR physical characteristics and reliability. (author)

  9. Structural capacity assessment of WWER-1000 MW reactor containment. Progress report

    International Nuclear Information System (INIS)

    Jordanov, M.

    1999-01-01

    The objective of the project is to provide assessment of the structural behaviour and safety capacity of the WWER-1000 MW Reactor Building Containment at Kozloduy NPP under critical combination of loads according to the current international requirements. The analysis is focused on a realistic assessment of the Containment taking into account the non-linear shell behaviour of the pre-stressed reinforced concrete structure. Previous assessments of the status of pre stressing cables pointed out that the efficiency of the Containment as a final defence barrier for internal and external events depends on their reliability. Due to this, the experimental data obtained from embedded sensors (gauges) at pre-stressed shell structure is to be compared with the results from analytical investigations. The reliability of the WWER-1000 MW accident prevention system is under evaluation in the project. The Soviet standard design WWER-1000 MW type units installed in Kozloduy NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1g. The new site seismicity studies revealed that the seismic hazard for the site significantly exceeds the originally estimated and a Review Level Earthquake (RLE) anchored to PGA=0.20g was proposed for re-assessment of the structures and equipment at Kozloduy NPP. The scope of the study is a re-assessment of the Containment structure under critical combination of loads according to the current safety and reliability requirements, including comparison between the Russian design requirements and the international regulations. Additionally, an investigation of the pre-stressing technology and the annual control of the cables' pre-stressing of the Containment is to be made. The crane influence on the dynamic behaviour of the Containment will be done as well as a study of the integrity of the Containment as a final defence barrier

  10. Experience running in and improvement of secondary circuit water chemistry of Kalinin NPP WWER-1000

    International Nuclear Information System (INIS)

    Noev, V.V.; Kukharev, N.D.; Otchenashev, G.D.; Guzeeva, G.I.; Kochetova, G.G.

    1991-01-01

    Basic characteristics of the secondary circuit water-chemical conditions at the Kalinin 1 and 2 reactors are presented. These are the WWER-1000 reactors with K-1000-1500 turbines. The analysis conducted makes it possible to conclude that all indicated values can meet the standards by introducing the hydrazine-ammonium regime in the secondary circuit with feedwater pH value equal to 9±0.2. Realization of design scheme for condensate-feeding circuit washing is necessary for acceleration of the water-chemical mode stabilization. Moreover the units should be equipped with automated chemical control instrumentation of a new generation

  11. Approach to normalization of the secondary circuit water chemistry of NPP with WWER-1000

    International Nuclear Information System (INIS)

    Mamet, V.A.; Erpyleva, S.F.; Banyuk, G.F.

    1998-01-01

    The approach to normalization if indices of water-chemical regime of the secondary circuit of the NPP with WWER-1000 reactor, based on pH calculational values at the coolant working temperature in dependence on the normalized admixtures concentration is considered. The possibility for conducting the water regime of steam generators by the ratio of sodium concentration and electrical conductivity of H-cation sample of blow-through water is shown. The limitations (os action level) by deviation of normalized indices from recommended ones for normal operational conditions are described

  12. Results of developing and problems of further improvements for WWER-1000 FA of alternative design

    International Nuclear Information System (INIS)

    Molchanov, V.L.; Panyushkin, A.K.; Zheleznyak, V.M.; Samojlov, O.B.; Kuul', V.S.; Kurylev, V.I.

    2001-01-01

    A new fuel assembly of alternative design (FAA) is created for a WWER-1000 reactor which is characteristic of dimensional stability on operation. The FAA includes a frame of 15 alloy Eh-110 spacer grids spot welded to 6 alloy Eh-635 angle sections. In-core tests show that the FAA is stable to form changing. It is concluded that the FAA can serve as a basis for development of promising fuel cycle with high burnups, specifically, for a 4 year cycle with the use of uranium-gadolinium fuel [ru

  13. Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

    International Nuclear Information System (INIS)

    Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

    1987-01-01

    Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

  14. Recommended reactor coolant water chemistry requirements for WWER-1000 units with 235U higher enriched fuel

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

    The last decade worldwide experience of PWRs and WWERs confirms the trends for the improvement of the nuclear power industry electricity production through the implementation of high burn-up or high fuel duty, which are usually accompanied with the usage of UO 2 fuel with higher content of 235 U - 4.0% - 4.5% (5.0%). It was concluded that the onset of sub-cooled nucleate boiling (SNB) on the fuel cladding surfaces and the initial excess reactivity of the core are the primary and basic factors accompanying the implementation of uranium fuel with higher 235 U content, aiming extended fuel cycles and higher burn-up of the fuel in Pressurized Water Reactors. As main consequences of the presence of these factors the modifications of chemical / electrochemical environments of nuclear fuel cladding- and reactor coolant system- surfaces are evaluated. These conclusions are the reason for: 1) The determination of the choices of the type of fuel cladding materials in respect with their enough corrosion resistance to the specific fuel cladding environment, created by the presence of SNB; 2) The development and implementation of primary circuit water chemistry guidelines ensuring the necessary low corrosion rates of primary circuit materials and limitation of cladding deposition and out-of-core radioactivity buildup; 3) Implementation of additional neutron absorbers which allow enough decrease of the initial concentration of H 3 BO 3 in coolant, so that its neutralization will be possible with the permitted alkalising agent concentrations. In this paper the specific features of WWER-1000 units in Bulgarian Nuclear Power Plant; use of 235 U higher enriched fuel in the WWER-1000 reactors in the Kozloduy NPP; coolant water chemistry and radiochemistry plant data during the power operation period of the Kozloduy NPP Unit 5, 15 th fuel cycle; evaluation of the approaches and results by the conversion of the WWER-1000 Units at the Kozloduy NPP to the uranium fuel with 4.3% 235 U as

  15. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    International Nuclear Information System (INIS)

    Beghini, M.; D'Auria, F.; Galassi, G.M.; Vitale, E.

    1997-01-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs

  16. Electromagnetism Mechanism for Enhancing the Refueling Cycle Length of a WWER-1000

    Directory of Open Access Journals (Sweden)

    Navid Poursalehi

    2017-02-01

    Full Text Available Increasing the operation cycle length can be an important goal in the fuel reload design of a nuclear reactor core. In this research paper, a new optimization approach, electromagnetism mechanism (EM, is applied to the fuel arrangement design of the Bushehr WWER-1000 core. For this purpose, a neutronic solver has been developed for calculating the required parameters during the reload cycle of the reactor. In this package, two modules have been linked, including PARCS v2.7 and WIMS-5B codes, integrated in a solver for using in the fuel arrangement optimization operation. The first results of the prepared package, along with the cycle for the original pattern of Bushehr WWER-1000, are compared and verified according to the Final Safety Analysis Report and then the results of exploited EM linked with Purdue Advanced Reactor Core Simulator (PARCS and Winfrith Improved Multigroup Scheme (WIMS codes are reported for the loading pattern optimization. Totally, the numerical results of our loading pattern optimization indicate the power of the EM for this problem and also show the effective improvement of desired parameters for the gained semi-optimized core pattern in comparison to the designer scheme.

  17. Electromagnetism mechanism for enhancing the refueling cycle length of a WWER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Poursalehi, Navid; Nejati-Zadeh, Mostafa; Minuchehr, Abdolhamid [Dept. of Nuclear Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2017-02-15

    Increasing the operation cycle length can be an important goal in the fuel reload design of a nuclear reactor core. In this research paper, a new optimization approach, electromagnetism mechanism (EM), is applied to the fuel arrangement design of the Bushehr WWER-1000 core. For this purpose, a neutronic solver has been developed for calculating the required parameters during the reload cycle of the reactor. In this package, two modules have been linked, including PARCS v2.7 and WIMS-5B codes, integrated in a solver for using in the fuel arrangement optimization operation. The first results of the prepared package, along with the cycle for the original pattern of Bushehr WWER-1000, are compared and verified according to the Final Safety Analysis Report and then the results of exploited EM linked with Purdue Advanced Reactor Core Simulator (PARCS) and Winfrith Improved Multigroup Scheme (WIMS) codes are reported for the loading pattern optimization. Totally, the numerical results of our loading pattern optimization indicate the power of the EM for this problem and also show the effective improvement of desired parameters for the gained semi-optimized core pattern in comparison to the designer scheme.

  18. Analysis of an accident with the main circulation tube rupture at the WWER-1000

    International Nuclear Information System (INIS)

    Boyadzhiev, A.I.; Stefanova, S.J.

    1984-01-01

    In connection with the forthcoming construction of a npp with the wwer-1000 reactor the loss of coolant accident associated with the main circulation tube rupture at the inlet near the reactor is analyzed. The relap4/mod6 program is used for the analysis. The data obtained show that the coolant outflow stage continues for about 25s. On the average the pressure in the circuits varies from 16 to 10 mpa per 0.1s and then it continues to decrease slowly. The pressure in the steam generator at the secondary circuits end increases approximately up to 6.9 MPa as a result of steam generator blocking and remaining coolant heating and then somewhat decreases owing to the primary circuit cooling. By the end of the fuel and can temperatures are equalized and the heat transfer coefficient is stabilized at the level of 100 w/1 (m 2 xK). It is concluded that during a loss of coolant accident at the wwer-1000 reactor in procesess of coolant blowdown in the medium power fuel elemets neither the fuel, melting temperature (3000 k), nor the critical temperature (1000 k) of plastic deformation zirconiu can initiation are attained

  19. Main examination results of WWER-1000 fuel after its irradiation in power reactors

    International Nuclear Information System (INIS)

    Bibiliashvili, Yu.; Dubrovin, K.; Vasilchenko, I.; Yenin, A.; Kushmanov, A.; Smirnov, A.; Smirnov, V.

    1994-01-01

    WWER-1000 fuel examination has been undertaken to specify the properties of fuel assembly members by defining the parameters of their materials and their interconnection in power reactor operation conditions. Nine fuel assemblies are examined. The examination program includes: visual inspection, measurement of overall dimensions, eddy-current test, gamma-scanning, X-ray and neutron radiography, analysis of gas pressure and composition inside fuel rods, ceramography/metallography, mass spectrometry, microanalysis and electron microscopy of fuel and fuel claddings. The examination results suggest that WWER-1000 fuel spent at steady-state operation conditions up to 50 Mwd/kg U of burnup is in satisfactory condition. The examination of all types of fuel cladding failures indicates that the reason lies in the interaction of cladding with coolant solid impurities. The nodular cladding corrosion of fuel assembly discharged from the South-Ukrainian NPP is caused by the graphite compounds deposited on the fuel rod. Those deposits are a result of the circulating pump damage and had accidental, non-typical character. Some of the rods were found to have a small cladding 'fretting' of the spacer grid cell material. The values of the majority of parameters determining the fuel efficiency allow to assume that there is a potential for further extension of fuel burnup and operation length. 1 tab., 11 figs

  20. Main examination results of WWER-1000 fuel after its irradiation in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bibiliashvili, Yu [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation); Dubrovin, K [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Vasilchenko, I [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation); Yenin, A; Kushmanov, A [AO Novosibirskij Zavod Khimcontsentratov, Novosibirsk (Russian Federation); Smirnov, A; Smirnov, V [Nauchno-Issledovatel` skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation)

    1994-12-31

    WWER-1000 fuel examination has been undertaken to specify the properties of fuel assembly members by defining the parameters of their materials and their interconnection in power reactor operation conditions. Nine fuel assemblies are examined. The examination program includes: visual inspection, measurement of overall dimensions, eddy-current test, gamma-scanning, X-ray and neutron radiography, analysis of gas pressure and composition inside fuel rods, ceramography/metallography, mass spectrometry, microanalysis and electron microscopy of fuel and fuel claddings. The examination results suggest that WWER-1000 fuel spent at steady-state operation conditions up to 50 Mwd/kg U of burnup is in satisfactory condition. The examination of all types of fuel cladding failures indicates that the reason lies in the interaction of cladding with coolant solid impurities. The nodular cladding corrosion of fuel assembly discharged from the South-Ukrainian NPP is caused by the graphite compounds deposited on the fuel rod. Those deposits are a result of the circulating pump damage and had accidental, non-typical character. Some of the rods were found to have a small cladding `fretting` of the spacer grid cell material. The values of the majority of parameters determining the fuel efficiency allow to assume that there is a potential for further extension of fuel burnup and operation length. 1 tab., 11 figs.

  1. Evaluation of the PTS potential in a WWER-1000 following a steam line break

    Energy Technology Data Exchange (ETDEWEB)

    Beghini, M; D` Auria, F; Galassi, G M; Vitale, E [Universita degli Studi di Pisa, Dipt. di Costruzioni Meccaniche e Nucleari, Pisa (Italy)

    1997-09-01

    A qualified nodalization for WWER-1000 is available at DCMN (Dipartimento di Costruzioni Meccaniche e Nucleari) of University of Pisa that is suitable for running with the thermohydraulic system code Relap5/mod3.2. The nodalization consists of about 1400 hydraulic nodes and more than 5000 mesh points for conduction heat transfer. The four loops of the NPP are separately modelled. Detailed information about the plant hardware has been gotten from contacts with Eastern Organizations in Bulgaria, Russia and Ukraine. The qualification of the nodalization has been achieved at a steady state level utilizing a procedure available at DCMN and at a transient level on the basis of operational (planned) transients performed in the Bulgarian Kozloduy-5 NPP and of the unplanned transient occurred at the Ukrainian Zaporosche NPP (April 1995). Data measured in steam generators have also been utilized. The nodalization has been widely applied to the analysis of accident scenarios in WWER-1000, including Large Break LOCA, Small Break LOCA, ATWS, Loss of Feedwater and Station Blackout. The present activity aims at evaluating the potential for PTS (Pressurized Thermal Shock) following a steam line break accident. The thermalhydraulic results were employed as input for a parametric Fracture Mechanics analysis based on conservative hypothesis of the shape and localization of a pre-existing defect. Stress analysis evidenced the effect of partial cooling of the vessel and gave some general indications of the risk for unstable crack propagation under the simulated PTS conditions. (author). 30 refs, 17 figs, 4 tabs.

  2. On steady-state concentrations of ammonia and molecular hydrogen in the primary circuit of the WWER-1000 reactors

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kamakchi, S.A.

    1997-01-01

    It is shown that the MORAVA-N2 software package describes well the coolant state in the primary circuit of an actual reactor facility with the WWER-1000 during on-load operation. It permits using the package for analysis of process perturbation effect on the coolant composition. Specific feature of ammonia radiation chemistry in the primary circuit of a reactor facility with the WWER-1000, assuring the rates hydrogen concentration in the coolant with ammonia concentration variation in the coolant within wide limits, when reactor operates on power, can be mentioned by way of example, the fact being ascertained in this study

  3. Fuel element cladding state change mathematical model for a WWER-1000 plant operated in the mode of varying loading

    Directory of Open Access Journals (Sweden)

    S. N. Pelykh

    2010-09-01

    Full Text Available Main features of a fuel element cladding state change mathematical model for a WWER-1000 reactor plant operated in the mode of varying loading are listed. The integrated model is based on the energy creep theory, uses the finite element method for imultaneous solution of the fuel element heat conduction and mechanical deformation equa-tions. Proposed mathematical model allows us to determine the influence of the WWER-1000 regime parameters and fuel assembly design characteristics on the change of cladding properties under different loading conditions of normal operation, as well as the cladding limiting state at variable loading depending on the length, depth and number of cycles.

  4. Determination and elimination of the reasons for increased control rod insertion time of the Kozloduy NPP WWER-1000

    International Nuclear Information System (INIS)

    Nikolov, K.

    1996-01-01

    The emergency insertion speed of the control rod upon reactor shutdown is of crucial importance for reactor safety. The designed insertion time for WWER-1000 reactors should be in the limits 1,5 to 4 s. Having in mind some data about increased insertion times of WWER-1000 type reactors in Russia and Ukraine, a practice of measuring this parameters during each planned outage of the Kozloduy NPP Unit 6 is introduced. Some technical improvements of the fuel assembly are made in order to reach the nominal parameters of the unit

  5. Determination of in-service change in the geometry of WWER-1000 core baffle: Calculations and measurements

    International Nuclear Information System (INIS)

    Margolin, B.Z.; Varovin, A.Y.; Minkin, A.J.; Sorokin, A.A.; Piminov, V.A.; Evdokimenko, V.V.; Fedosovsky, M.E.; Sherstobitov, A.E.; Ovchinnikov, A.G.; Pikulik, S.S.; Erak, D.Y.; Bobkov, A.V.; Timofeev, A.M.; Timokhin, V.I.; Yakushev, S.V.; Vasiliev, V.G.

    2015-01-01

    The paper gives the basic constitutive equations describing radiation swelling and creep depending on neutron dose, irradiation temperature and triaxial stress state, and justifies these equations experimentally. The WWER-1000 core baffle change in geometry was calculated by different models describing the effect of stresses on radiation swelling. The calculated results are compared with the measured ones for the operating WWER-1000 core baffle at the Balakovo NPP, Unit 1. A method of individual prediction of core baffle geometry change on the basis of the measurement results has been proposed. (authors)

  6. Examples for cost reduction in the design of a WWER-1000 nuclear power plant

    International Nuclear Information System (INIS)

    Kukkola, T.

    1991-01-01

    In a design project during recent years, a version for Finnish conditions has been and is being developed based on the Soviet WWER-1000 PWR plant with four horizontal steam generators. The plant will have a double containment. The inner containment will be a dry full pressure prestressed concrete containment with liner and the secondary containment will be made of ordinary concrete. Four train safety approach is adopted. It is supposed that the plant is to be designed according to the present Finnish safety requirements, e.g. severe reactor accidents are considered. When striving at an economic plant no compromises are made as far as safety is concerned. This paper describes possible cost reduction by redesigning the main technical equipment. (author). 1 ref

  7. Calculated and experimental research of WWER-1000 assembly vibration and fretting damage

    International Nuclear Information System (INIS)

    Drozdov, Y.; Afanasyev, A.; Makarov, V.; Tutnov, A.; Tutnov, A.; Alekseev, E.

    2008-01-01

    The report covers the methods and results of the latest analytical and experimental studies of fretting corrosion and natural vibrations of a WWER-1000 reactor fuel assemblies (FA). The process of fretting-corrosion was investigated using a multi-specimen facility that simulated fragments of fuel rod-to-spacer grid and lower support grid mating units. A computational model was developed for vibrations in the mechanical system of a fuel rod fragment and a spacer grid fragment. A calculational and experimental modal analysis of a FA was performed. Natural frequencies, modes and decrements of FA vibrations were determined and a satisfactory coincidence of analytical and experimental results was obtained. The assessment of fretting-corrosion process dynamics was made and its dependences on operational factors were obtained. (authors)

  8. Neutron flux uncertainty and covariances for spectrum adjustment and estimation of WWER-1000 pressure vessel fluences

    International Nuclear Information System (INIS)

    Boehmer, Bertram

    2000-01-01

    Results of estimation of the covariance matrix of the neutron spectrum in the WWER-1000 reactor cavity and pressure vessel positions are presented. Two-dimensional calculations with the discrete ordinates transport code DORT in r-theta and r-z-geometry used to determine the neutron group spectrum covariances including gross-correlations between interesting positions. The new Russian ABBN-93 data set and CONSYST code used to supply all transport calculations with group neutron data. All possible sources of uncertainties namely caused by the neutron gross sections, fission sources, geometrical dimensions and material densities considered, whereas the uncertainty of the calculation method was considered negligible in view of the available precision of Monte Carlo simulation used for more precise evaluation of the neutron fluence. (Authors)

  9. Experimental verification of FA for WWER-1000. Investigations of TVS - 2M

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Seleznev, A.; Kobelev, S.; Makarov, V.; Afanasjev, A.; Matvienko, I.; Enin, A.; Ustimenko, A.; Volkov, S.

    2008-01-01

    During development of TVS-2M design for WWER-1000 and within the scope of its pre-reactor verification there was performed a complex of bench tests of FA for operational impacts. These tests were carried out with the use of the full-scale non-irradiated FA dummy and the models of units. The present report presents the methods and the results of tests of the full-scale dummy for static concentrated loads, impact of thermal cycling and vibration under separate and simultaneous impact of the listed loads. The results of tests carried out indicate that the solutions implemented in TVS-2M, first of all, a rigid welded skeleton, provide much higher resistance to distortion due to thermomechanical loads in comparison with zirconium AFA of the previous generation. This work was organized by JSC TVEL. (authors)

  10. Data base and postirradiation examination results of spent WWER-1000 fuel elements and assemblies

    International Nuclear Information System (INIS)

    Kanashov, B.A.; Polenok, V.S.; Smirnov, A.V.; Zhitelev, V.A.

    1995-01-01

    The report presents the results of the postirradiation shape change examination of standard fuel elements and fuel assemblies irradiated in standard conditions in Russian power reactors of the WWER-1000 type. The information is based on the results obtained at the Fuel Research Department of the Federal Scientific Centre Research Institute of Atomic Reactors (FSC RIAR, Dimitrovgrad, Russian Federation) within the period from 1987 to 1994. Emphasis is placed on such experimental and calculational data as: length, cross-section dimensions and shape of FAs with wrapper; change of standard FA skeleton members dimensions; fuel bundle elongation; change of the fuel cladding outer diameter; and elongation and change of the fuel stack outer diameter. (author)

  11. Status and trends in nuclear technology for power plants with WWER-1000 reactors. Review

    Energy Technology Data Exchange (ETDEWEB)

    Zorev, N N

    1977-04-01

    The problems of improving quality of nuclear equipment for WWER-1000 power plants and associated nuclear technology automation are surveyed. Examples of technological innovations are presented which significantly reduce labour intensity, time consumption and increase quality standards of the products. Some new automated equipments for materials welding, working, machining and quality control are described. The discussion is centering around heavy-section steel technologies. Some mechanical properties of new-developed nuclear grade steels designed for producing reactor vessels and steamgenerators, volume compensators and pipes, as well as steam separators and steamsuperheaters are also presented. Their properties (impact strength and radiation resistance) are pointed out to be superior to that of steels used abroad. The basic trend in nuclear structural material developments is towards integrated optimization of strength, performance and workability.

  12. Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    D`Auria, F.; Galassi, G.M. [Univ. of Pisa (Italy); Frogheri, M. [Univ. of Genova (Italy)

    1997-12-31

    The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

  13. Fuel cycles of WWER-1000 based on assemblies with increased fuel mass

    International Nuclear Information System (INIS)

    Kosourov, E.; Pavlovichev, A.; Shcherenko, A.

    2011-01-01

    Modern WWER-1000 fuel cycles are based on FAs with the fuel column height of 3680 mm, diameters of the fuel pellet and its central hole of 7.6 and 1.2 mm respectively. The highest possible fuel enrichment has reached its license limit that is 4.95 %. Research in the field of modernization, safety justification and licensing of equipment for fuel manufacture, storage and transportation are required for further fuel enrichment increase (above 5 %). So in the nearest future an improvement of technical and economic characteristics of fuel cycles is possible if assembly fuel mass is increased. The available technology of the cladding thinning makes it possible. If the fuel rod outer diameter is constant and the clad inner diameter is increased to 7.93 mm, the diameter of the fuel pellet can be increased to 7.8 mm. So the suppression of the pellet central hole allows increasing assembly fuel weight by about 8 %. In this paper we analyze how technical and economic characteristics of WWER-1000 fuel cycle change when an advanced FA is applied instead of standard one. Comparison is made between FAs with equal time interval between refueling. This method of comparison makes it possible to eliminate the parameters that constitute the operation component of electricity generation cost, taking into account only the following technical and economic characteristics: 1)cycle length; 2) average burnup of spent FAs; 3) specific natural uranium consumption; 4)specific quantity of separative work units; 5) specific enriched uranium consumption; 6) specific assembly consumption. Collected data allow estimating the efficiency of assembly fuel weight increase and verifying fuel cycle characteristics that may be obtained in the advanced FAs. (authors)

  14. Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    D` Auria, F; Galassi, G M [Univ. of Pisa (Italy); Frogheri, M [Univ. of Genova (Italy)

    1998-12-31

    The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

  15. Analysis of expediency to set regulators of high-pressure emergency core cooling system of WWER 1000 (B-320)

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Komarov, Yu.A.; Tikhonova, G.G.; Nikiforov, S.N.; Bogodist, V.V.; Fol'tov, I.M.; Khadzh Faradzhallakh Dabbakh, A.

    2011-01-01

    The work shows that setting regulative valves in high-pressure emergency core cooling system of WWER 1000/B-320 can be effective only involving the additional tuning to account traverse speed of operating elements of regulator and configuration of the systems providing cooling of primary loop.

  16. Structural optimization of static power control programs of nuclear power plants with WWER-1000

    International Nuclear Information System (INIS)

    Kokol, E.O.

    2015-01-01

    The question of possibility the power control programs switching for WWER-1000 is considered. The aim of this research is to determine the best program for the power control of nuclear reactor under cyclic diurnal behavior of electrical generation, as well as the switching implementation. The considered problem of finding the best control program refers to the multicriteria optimization class of problems. Operation of the nuclear power generation system simulated using the following power control programs: with constant average temperature of transfer fluid, with constant pressure in the reactor secondary circuit, with constant temperature in input of the nuclear reactor. The target function was proposed. It consists of three normalized criteria: the burn up fraction, the damage level of fuel rod array shells, as well as changes in the power values. When simulation of the nuclear power generation system operation within the life was done, the values of the selected criteria were obtained and inserted in the target function. The minimum of three values of the target function depending on the control program at current time defined the criterion of switching of considered static power control programs for nuclear power generation system

  17. Experimental assessment of welded joints brittle fracture on the crack arrest criterion for WWER-1000 RPV

    International Nuclear Information System (INIS)

    Blumin, A.A.; Timofeev, B.T.

    2000-01-01

    The crack arrest fracture toughness in a vessel steel used in WWER-1000 reactor, namely in steel 15Kh2NMFA and its submerged arc welded joints, produced with Sv-08KhGNMTA, Sv-12 Kh2NMFA welding wires and NF-18 M, FZ-16 A welding fluxes, is under study. Experimental studies are carried out using three heats with the chemical composition meeting the specifications. Weld specimens 100-200 mm thick are subjected to tempering according various regimes to induce the embrittlement and simulate mechanical properties (yield strength and ductile-brittle transition temperature) corresponding to those at the end of service life under neutron radiation effect. Base metal and weld properties are compared. The wide scatter is noted for experimental data on fracture toughness temperature dependences. A possibility to use the dependence of K Ia = f (T-T k ) for determining the crack arrest fracture toughness is discussed taking in account that K Ia is a stress intensity factor calculated within the frame of static fracture mechanics [ru

  18. Fuel Management of WWER-1000 Reactors of Kudankulam Nuclear Power Plant, India

    International Nuclear Information System (INIS)

    Pandey, Y.; Chauhan, A.

    2008-01-01

    Two units of WWER-1000 reactors of Russian design are under construction at Kudankulam site in India. These reactors are expected to be commissioned in 2008. The fuel management services for these reactors shall be carried out using Russian Computer codes. This paper includes a brief description of the core, fuel assembly lattice and physics modeling of the lattice and core for these reactors. Presented in this paper are the salient features of the core load pattern designs and fuel performance for 8 operating cycles of these reactors. The paper describes key improvements in the core load pattern designs to enhance the fuel utilization and its thermal behaviour. Presented in the paper are also the on site fuel management strategies with regard to fuel inventory and nuclear material accounting. A computer code for Fuel Inventory and Nuclear Material Accounting (FINMAC) has been developed for this purpose. The code FINMAC takes care of receipt of fresh fuel, flow between various accounting sub areas (ASAs), burnup or production of nuclear isotopes in the reactor cores and discharge from the reactor core. The code generates Material Balance Reports (MBRs) and Composition of Ending Inventory Reports (COEIs) as per the IAEA standards. (authors)

  19. WWER-1000 reactor simulator. Material for training courses and workshops. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No.12, Reactor Simulator Development (2001). Course material for workshops using a pressurized water reactor (PWR) simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication, Training Course Series No. 22, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

  20. Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

  1. Effect of burnup history by moderator density on neutron-physical characteristics of WWER-1000 core

    International Nuclear Information System (INIS)

    Ovdiienko, I.; Kuchin, A.; Khalimonchuk, V.; Ieremenko, M.

    2011-01-01

    Results of assessment of burnup history effect by moderator density on neutron physical characteristics of WWER-1000 core are presented on example of stationary fuel loading with Russian design fuel assembly TWSA and AER benchmark for Khmelnitsky NPP that was proposed by TUV and SSTC NRC at nineteenth symposium. Assessment was performed by DYN3D code and cross section library sets generated by HELIOS code. Burnup history was taken into account by preparing of numerous cross section sets with different isotopic composition each of which was obtained by burning under different moderator density. For analysis of history effect 20 cross section sets were prepared for each fuel assembly corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect. Main attention of study was paid to effect of burnup history by moderator density to axial power distribution. Results of study for AER benchmark were compared with experimental values of axial power distribution for fuel assemblies of first, second, third and fourth year operation. (Authors)

  2. Comparison of HRA methods based on WWER-1000 NPP real and simulated accident scenarios

    International Nuclear Information System (INIS)

    Petkov, Gueorgui

    2010-01-01

    Full text: Adequate treatment of human interactions in probabilistic safety analysis (PSA) studies is a key to the understanding of accident sequences and their relative importance in overall risk. Human interactions with machines have long been recognized as important contributors to the safe operation of nuclear power plants (NPP). Human interactions affect the ordering of dominant accident sequences and hence have a significant effect on the risk of NPP. By virtue of the ability to combine the treatment of both human and hardware reliability in real accidents, NPP fullscope, multifunctional and computer-based simulators provide a unique way of developing an understanding of the importance of specific human actions for overall plant safety. Context dependent human reliability assessment (HRA) models, such as the holistic decision tree (HDT) and performance evaluation of teamwork (PET) methods, are the so-called second generation HRA techniques. The HDT model has been used for a number of PSA studies. The PET method reflects promising prospects for dealing with dynamic aspects of human performance. The paper presents a comparison of the two HRA techniques for calculation of post-accident human error probability in the PSA. The real and simulated event training scenario 'turbine's stop after loss of feedwater' based on standard PSA model assumptions is designed for WWER-1000 computer simulator and their detailed boundary conditions are described and analyzed. The error probability of post-accident individual actions will be calculated by means of each investigated technique based on student's computer simulator training archives

  3. Study of transient connected with WWER-1000 cluster drop with subsequent working of automatic power controller

    International Nuclear Information System (INIS)

    Kuchin, A.; Ovdiienko, I.; Khalimonchuk, V.

    2010-01-01

    Results of calculation study of transient connected with drop of WWER-1000 cluster of working group are presented. Transient was considered in the mode of automatic power control without forming of warning protection signal due to reaching of dropped cluster of core bottom. Calculations are shown that given transient can cause valuable distortion of power distribution in axial direction. At that main increase of pin power is occurred in upper part of the core, whereas power in lower part is almost not changed. The additional increase of power in the upper part of core makes conditions for initiation of DNB. This effect can be observed if in initial state axial power distribution is displaced in upper part of core nearby to rest of supported power clusters of working group. It is necessary to define conservatively with taking into account assumed working group efficiency-in which row from extracted clusters of working group the displacement of axial power in the upper part is possible. Probability of such displacement and its localization in plane of core must be properly analyzed. The work was performed in framework of orders BMU SR 2511 and BMU R0801504 (SR2611). The report describes the opinion and view of the contractor-State Scientific and Technical Centre on Nuclear and Radiation Safety-and does not necessarily represent the opinion of the ordering party - BMU-BfS/GRS and TUEV SUED. (Authors)

  4. WWER-1000 unit reliability problems from the point of view of the main supplier of technological equipment

    International Nuclear Information System (INIS)

    Bursa, V.; Holousova, M.; Turnik, S.

    1990-01-01

    At Skoda Works in Plzen, data from the period of construction of nuclear power plants are processed on an ICL DRS 300 computer. The database systems DBASE II and DATAFLEX are available for creating reliability information systems. The information system that is being developed for WWER-1000 units is tested at the WWER-440 units of the Dukovany and Mochovce nuclear power plants. Activities in the field of evaluation of structure reliability of WWER-1000 nuclear power plants are aimed at two major goals, viz., developing a methodology for testing the reliability of the whole unit and its subsystems, and performing reliability analysis and calculations of reliability indices of the secondary circuit of a WWER-1000 nuclear power plant. The reason for the latter concern is the fact that in 1984-1986, secondary circuits contributed 46% to failures of Czechoslovak WWER-440 nuclear power plants. According to existing analyses, the time fluctuations of reliability indices obey no rule that could be employed for inferring indices expected in steady-state operating conditions from indices established in the starting stage of operation. (Z.M.). 10 refs

  5. The procedure for determination of special margin factors to account for a bow of the WWER-1000 fuel assemblies

    International Nuclear Information System (INIS)

    Tsyganov, S. V.; Marin, S. V.; Shishkov, L. K.

    2008-01-01

    Starting from 1980s, the problem of bow of the WWER-1000 reactor fuel assemblies and the effect of that on the operational safety is being discussed. At the initial period of time, the extension of time for dropping control rods of the control and protection system associated with this bow posed the highest threat. Later on, new more rigid structures were developed for fuel assemblies that eliminated the problems of control rods. However, bow of the WWER-1000 reactor fuel assemblies is observed up to now. The scale of this bow reduced significantly but it still effects safety. Even a minor bow available may lead to the noticeable increase of power of individual fuel pins associated with the local variation of the coolant amount. This effect must be taken into account on designing fuel loadings to eliminate the exceeding of set limitations. The introduction of additional special margins is the standard method for taking this effect into account. The present paper describes the conservative technique for the assessment of additional margins for bow of fuel assemblies of state-of-the-art designs. This technique is employed in the WWER-1000 reactor designing. The chosen conservatism degree is discussed as well as the method for its assurance and acceptable ways for its slackening. The example of the margin evaluation for the up-to-date fuel loading is given. (authors)

  6. Dynamic behavior structural response and capacity evaluation of the standardized WWER-1000 nuclear power plants subjected to severe loading conditions

    International Nuclear Information System (INIS)

    Ambriashvili, Y.K.; Krutzik, N.J.

    1993-01-01

    In order to verify the structural capacity of standardized WWER-1000 MW nuclear power plants, comprehensive static and dynamic analyses were performed in cooperation between Siemens and Atomenergoprojekt. The main goal of these investigations was to perform of a number of seismic analyses of standardized WWER-1000 reactor buildings on the basis of 13 given seismological inputs, taking into account the local soil conditions at 17 different sites defined by in-situ investigations. The analyses were based on appropriate mathematical models (equivalent beam models as well as detailed spatial surface element models) of the coupled vibrating structures (base structure, outer structure, containment, inner structure) and of the layered soil. The analyses were mainly performed using the indirect method (substructure method). Based on the results of the seismic analysis as well as the results of static analysis (pressure and temperature due to LOCA, dead weight, prestressing) an assessment was made of the seismic safety of the containment and the reactor building. Using a complex 3-dimensional model of the structure and the soil, the influence of the flexibility of the basement structure on the structural response was also studied. The structural analyses of the WWER-1000 reactor building led to the conclusion that its design accounts well for the main factors governing the dynamic behavior of the building. The assessment of the forces acting in the structures shows that the bearing capacity of the analyzed building structure corresponds to an earthquake intensity of about 0.2 g to 0.25 g

  7. An approach to stability analysis of spatial xenon oscillations in WWER-1000 reactors

    International Nuclear Information System (INIS)

    Parhizkari, H.; Aghaie, M.; Zolfaghari, A.; Minuchehr, A.

    2015-01-01

    Highlights: • The multipoint methodology is developed for xenon oscillation in the BNPP. • The axial, radial and azimuthal offsets are calculated in the BOC and EOC. • It is shown that the all of oscillation modes are safe in the BOC. • The axial oscillation is not safe in the EOC and needs governor control system. • The multipoint kinetics show good agreement for spatial oscillations. - Abstract: Spatial power oscillations due to spatial distribution of xenon transient are well known as xenon oscillation in large reactors. Xenon-induced spatial power oscillations occur as a result of rapid perturbations to power distribution that cause the xenon and iodine distribution to be out of phase with the perturbed power distribution. This results in a shift in xenon and iodine distributions that causes the power distribution to change in an opposite direction from the initial perturbation. In this paper xenon-induced power oscillation is described by a system of differential equations with non-linearity between xenon and flux distributions; the dynamics of process is described by a discrete distributed parameter model, with the neutron flux, the delayed neutrons, the core temperature and the xenon and iodine concentrations as the “states” of the system. It is shown that it is possible to describe the discrete distributed-parameter as a set of coupled point-reactor models. It is also shown that using this scheme it is possible to analyze the control aspects of a multi-section large core reactor by treating only two adjacent sections of the core. To illustrate the capability and efficiency of the proposed scheme Bushehr Nuclear Power Plant, BNPP, which is a WWER-1000 reactor, is chosen to show the performance of the methodology. The axial, azimuthal and radial power oscillation at the beginning of cycle, BOC, and the end of cycle, EOC, for BNPP are investigated; the results are in good agreement with safety analysis report of the reference plant

  8. Seismic strengthening of overhead roads between reactor buildings of WWER-1000 MW type NPP

    International Nuclear Information System (INIS)

    Stoyanov, G.; Jordanov, M.

    2005-01-01

    This paper presents results obtained during the upgrading design of overhead roads (OHR) between WWER-1000 MW Reactor Units at Kozloduy NPP. In order to avoid the deficiencies of OHR seismic capacity different approaches were developed based on the site and structure specifics. Overhead roads are precasted RC structures. They consist of pedestrian gallery and pipeline RC box, connecting reactor buildings with auxiliary building. They are mounted at approximately 10 m above ground level. The overhead roads are evaluated at their as-is status and a seismic upgrading of the structure is designed. The analysis of the upgraded structure is performed for Review Level Earthquake (RLE). Soil-Structure Interaction (SSI) effects are taken into account through equivalent soil springs with frequency adjusted stiffnesses. The upgraded structure is checked for conformance with the specially developed technical design specification based on International, US and Bulgarian standards and codes, taking into account site specific conditions. The general approach is consistent with up-to-date practice for evaluation and upgrade of nuclear power plant facilities. The existing site conditions and Owner's requirements are taken into account during development of the upgrading. The proposed upgrading measures can be divided in two major categories global and local. Special attention is paid to improvement of the ductile behavior of the structure through new detailing and upgrading of existing connection. These measures are grouped in two final design concepts and after a comparative study one of them is chosen for implementation. For the upgraded structure response spectra are derived at locations where equipment is attached. (authors)

  9. Study on severe accidents and countermeasures for WWER-1000 reactors using the integral code ASTEC

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Altstadt, E.; Kliem, S.; Reinke, N.

    2011-01-01

    The research field focussing on the investigations and the analyses of severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises or the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. On the one side, it is the aim of the severe accident research to understand the complex phenomena during the in- and ex-vessel phase, involving reactor-physics, thermal-hydraulics, physicochemical and mechanical processes. On the other side the investigations strive for effective severe accident management measures. This paper is focused on the possibilities for accident management measures in case of severe accidents. The reactor pressure vessel is the last barrier to keep the molten materials inside the reactor, and thus to prevent higher loads to the containment. To assess the behaviour of a nuclear power plant during transient or accident conditions, computer codes are widely used, which have to be validated against experiments or benchmarked against other codes. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The results have shown that in case of unavailability of major active safety systems the reactor pressure vessel would ultimately fail. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accident, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed. The presented investigations contribute to the validation of the European integral severe accidents code ASTEC for WWER-1000 type of reactors

  10. Comparative analysis of exercise 2 results of the OECD WWER-1000 MSLB benchmark

    International Nuclear Information System (INIS)

    Kolev, N.; Petrov, N.; Royer, E.; Ivanov, B.; Ivanov, K.

    2006-01-01

    In the framework of joint effort between OECD/NEA, US DOE and CEA France a coupled three-dimensional (3D) thermal-hydraulic/neutron kinetics benchmark for WWER-1000 was defined. Phase 2 of this benchmark is labeled W1000CT-2 and consists of calculation of a vessel mixing experiment and main steam line break (MSLB) transients. The reference plant is Kozloduy-6 in Bulgaria. Plant data are available for code validation consisting of one experiment of pump start-up (W1000CT-1) and one experiment of steam generator isolation (W1000CT-2). The validated codes can be used to calculate asymmetric MSLB transients involving similar mixing patterns. This paper summarizes a comparison of the available results for W1000CT-2 Exercise 2 devoted to core-vessel calculation with imposed MSLB vessel boundary conditions. Because of the recent re-calculation of the cross-section libraries, core physics results from PARCS and CRONOS codes could be compared only. The comparison is code-to-code (including BIPR7A/TVS-M lib) and code vs. plant measured data in a steady state close to the MSLB initial state. The results provide a test of the cross-section libraries and show a good agreement of plant measured and computed data. The comparison of full vessel calculations was made from the point of view of vessel mixing, considering mainly the coarse-mesh features of the flow. The FZR and INRNE results from multi-1D calculations with different mixing models are similar, while the FZK calculations with a coarse-3D vessel model show deviations from the others. These deviations seem to be due to an error in the use of a boundary condition after flow reversal (Authors)

  11. The main conditions ensured problemless implementation of 235U high enriched fuel in Kozloduy NPP (Bulgaria) - WWER-1000 Units

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.; Minkova, K.; Michaylov, G.; Penev, P.; Gerchev, N.

    2009-01-01

    The collected water chemistry and radiochemistry data during the operation of the Kozloduy NPP Unit 5 for the period 2006-2009 (12-th, 13-th 14-th and 15-th fuel cycles) undoubtedly indicate for WWER-1000 Units (whose specific features are: Steam generators with austenitic stainless steel 08Cr18N10T tubing; Steam generators are with horizontal straight tubing and Fuel elements cladding material is Zr-1%Nb (Zr1Nb) alloy), that one realistic way for problemless implementation of 235 U high enriched fuel have been found. The main feature characteristics of this way are: Implementation of solid neutron burnable absorbers together with the dissolved in coolant neutron absorber - natural boric acid; Application of fuel cladding materials with enough corrosion resistance by the specific fuel cladding environment created by presence of SNB; Keeping of suitable coolant water chemistry which ensures low corrosion rates of core- and out-of-core- materials and limits in core (cladding) depositions and restricts out-of-core radioactivity buildup. The realization of this way in WWER-1000 Units in Kozloduy NPP was practically carried out through: 1) Implementation of Russian fuel assemblies TVSA which have as fuel cladding material E-110 alloy (Zr1Nb) with enough high corrosion resistance by presence of sub-cooled nucleate boiling (SNB) and use burnable absorber (Gd) integrated in the uranium-gadolinium (U-Gd 2 O 3 ) fuel (fuel rod with 5.0% Gd 2 O 3 ); 2) Development and implementation of water chemistry primary circuit guidelines, which require the relation between boric acid concentration and total alkalising agent concentrations to ensure coolant pH 300 = 7.0 - 7.2 values during the whole operation period. The above mentioned conditions by the passing of WWER-1000 Units in NPP Kozloduy to uranium fuel with 4.4% 235 U (TVSA fuel assemblies) practically ensured avoidance of the creation of the necessary conditions for AOA onset. The operational experience (2006-2009) of the

  12. Calculational modeling of fuel assemblies of WWER-1000 type with the use of burnable absorber Gadolinum; comparative analysis

    International Nuclear Information System (INIS)

    Yeremenko, M.L.; Kovbasenko, Yu.P.; Loetsch, T.

    2001-01-01

    In connection with the beginning of the use of fuel assemblies with burnable absorbers by integration of Gadolinum into the nuclear fuel at Ukrainian NPP the task of testing the code systems and the pertinent neutron cross section libraries for the new fuel arose. Taking into account the long term experience of German experts with calculations and evaluation of nuclear fuel containing Gadolinum it was decided to carry out a series of test calculations for fuel assembly lattices of PWR, WWER-440 and WWER-1000 types using the NESSEL/PYTHIA and CASMO/SIMULATE code systems (Authors)

  13. Evolution of ASTEC V1.2 rev.1 code for WWER-1000 reactors/SBO sequence

    International Nuclear Information System (INIS)

    Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Kalchev, B.; Passalacqua, R.

    2006-01-01

    In this paper a comparison between calculations of severe accidents occurred from WWER-1000 with ASTEC code specified for an event of full unloading with relief valves stuck opened with no hydroaccumulators intervention is presented. The purpose of the analyses provided is to present the relationship between the improvements of the actual version (ASTEC Vl.2 rev. 1) and ASTEC V1.1 p2 like: code modifications, incoming data improvements. Such discrepancies are to be examined. Case by case suggestions for ASTEC improvements are to be provided

  14. Quantum behaved Particle Swarm Optimization with Differential Mutation operator applied to WWER-1000 in-core fuel management optimization

    International Nuclear Information System (INIS)

    Jamalipour, Mostafa; Sayareh, Reza; Gharib, Morteza; Khoshahval, Farrokh; Karimi, Mahmood Reza

    2013-01-01

    Highlights: ► A new method called QPSO-DM is applied to BNPP in-core fuel management optimization. ► It is found that QPSO-DM performs better than PSO and QPSO. ► This method provides a permissible arrangement for optimum loading pattern. - Abstract: This paper presents a new method using Quantum Particle Swarm Optimization with Differential Mutation operator (QPSO-DM) for optimizing WWER-1000 core fuel management. Genetic Algorithm (GA) and Particle Swarm Optimization (PSO) have shown good performance on in-core fuel management optimization (ICFMO). The objective of this paper is to show that QPSO-DM performs very well and is comparable to PSO and Quantum Particle Swarm Optimization (QPSO). Most of the strategies for ICFMO are based on maximizing multiplication factor (k eff ) to increase cycle length and minimizing power peaking factor (P q ) in order to improve fuel integrity. PSO, QPSO and QPSO-DM have been implemented to fulfill these requirements for the first operating cycle of WWER-1000 Bushehr Nuclear Power Plant (BNPP). The results show that QPSO-DM performs better than the others. A program has been written in MATLAB to map PSO, QPSO and QPSO-DM for loading pattern optimization. WIMS and CITATION have been used to simulate reactor core for neutronic calculations

  15. Transient simulations in WWER-1000-comparison between DYN3D-ATHLET and DYN3D-RELAP5

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Mittag, S.; Rohde, U.; Weiss, F.P.

    2003-01-01

    Simulations of a real transient of an operating WWER-1000 power plant have been performed using DYN3D-ATHLET (Gru95) and DYN3D-RELAP5 (Koy01) code systems in the frame of activities aimed at a validation of the neutronic / thermal-hydraulic coupled codes. The transient initiated by a main coolant pump switching off, when three of the four main coolant pumps of the plant are in operation (scenario of the VALCO project) is chosen for the simulation. The same models of the plant (except the core nodalization) but two different libraries of macroscopic cross-sections have been used in compared calculations. Additionally, the compared code systems are based on the different / external and internal / coupling techniques. This paper contains a brief description of the coupled codes and the plant model as well as a comparison between the results from simulations (Authors)

  16. Study of accelerated unit unloading mode initiated by turbine feed pump trip with TVSA fuel assemblies operation in WWER-1000

    International Nuclear Information System (INIS)

    Borysenko, V.I.; Kadenko, I.N.; Samoilenko, D.V.

    2012-01-01

    This paper provides the study results of accelerated unit unloading mode (AUU) initiated at WWER-1000 unit operated at 100 % power and its expediency in the event of single Turbo Feed Pump (TFP) failure. Modeling was performed using an advanced calculation code RELAP/SCDAPSIM/Mod3.4 and relevant model for KhNPP Unit No. 2. As the study shows, SCRAM cannot be prevented in case of failure of 3 main circulation pumps due to steam generators (SG) level drop. Based on the results obtained, it is reasonably justified to allow SCRAM signal instead of AUU activation in case of single TFP failure at power level more than 90 % of N n om. This will provide more sparing temperature modes for fuel assemblies and equipment, as well as prevent additional thermal cycling loads and violation of safe operation limits as SG water levels

  17. Development of new chemical and electrochemical decontamination methods for selected equipment of WWER-440 and WWER-1000 reactor primary circuit

    International Nuclear Information System (INIS)

    Solcanyi, M.; Majersky, D.

    1998-01-01

    Special devices for in-situ application of decontamination technologies assigned for Steam Generator, Pressurizer and Main Circulating Casing of WWER-1000 type were designed, manufactured and tested in real conditions of their use in above Primary Circuit components. New decontamination technologies like low-concentration process NP-NHN for the decontamination of the Steam Generator, combined chemico-mechanical treatment for the Pressurizer and semi-dry electrolysis for the Main Circulating Pump Casing were developed and approved for their safe plant application from point of view of decontamination efficiency, corrosion influence and processing of secondary wastes. Main technological parameters were defined to achieve high decontamination efficiency and corrosion-safe application of all decontamination technologies. (author)

  18. Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

    International Nuclear Information System (INIS)

    Belousov, S.; Antonov, S.; Ilieva, K.

    1997-01-01

    The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

  19. Effect of high-temperature filtration on impurity composition in the primary circuit coolant of power units with WWER-1000 reactors

    International Nuclear Information System (INIS)

    Efimov, A.A.; Moskvin, L.N.; Gusev, B.A.; Leont'ev, G.G.; Nekrest'yanov, S.N.

    1992-01-01

    The effects of high-temperature filtration on changes in dispersive, chemical, radioisotope and phase compositions of impurities in primary circuit coolant of NPP with the WWER-1000 reactor are studied. Special filters are used for the studies. The data obtained confirm the applicability of high-temperature filtration for purification of WWER reactor water and steam separators at NPPs with RBMK reactors

  20. WWER identification and analysis of dominant factors affecting the fuel failure rates in WWER-1000 units in Czech Republic, Bulgaria, Ukraine and Russia

    International Nuclear Information System (INIS)

    Evdokimov, I.; Likhanskii, V.; Afanasieva, E.; Kanukova, V.; Kozhakin, A.; Maslova, L.; Chernetskiy, M.; Zborovskii, V.; Sorokin, A.

    2015-01-01

    The paper reviews the major findings of the study in the frame of the “Zero Failure Rate” project for WWER. The study included analysis and systematization of available data on leaking fuel assemblies found in 2003 through 2014 in WWER-1000 nuclear units in Russia, Ukraine, Czech Republic and Bulgaria. The study was intended to be used in preparation of recommendations and elaboration of corrective measures for enhancement of reliability and decrease of the failure rates for the WWER-1000 fuel. One of the key areas in successful implementation of the industry ‘zero failure’ goal is a challenge of significant increase of inspections of WWER-1000 fuel assemblies. It may be reasonable (with account taken for international experience) to think of development of more effective equipment for prompt fuel inspections & repair in WWER-1000 spent fuel pool. Another challenge is the elaboration of unified fuel inspection guidelines to ensure that limited industry resources are spent in the most productive way. In the frame of this work it may be helpful to implement in practice the criteria for safe removal of defective fuel rods from the leaking FA under repair

  1. Determination of neutron fluence and radiation brittleness temperature of WWER-440 and WWER-1000 pressure vessels in Kozloduy NPP

    International Nuclear Information System (INIS)

    Ilieva, K.; Apostolov, T.; Belousov, S.; Petrova, T.; Antonov, S.; Ivanov, K.; Prodanova, R.

    1993-01-01

    In Units 1-4 of Kozloduy NPP (WWER-440/230), the weld 4 of RPV undergoes the most severe irradiation embrittlement. Neither witness-samples, nor detectors are designed for these reactors. Transport calculations of fast neutron fluence on WWER-440 RPV and ex-vessel measurements by threshold activation detectors are the primary means for adequate assessment of metal state and for prognosis concerning the reactor life span. In WWER-1000 reactors (Units 5-6) the maximum neutron fluence occurs on the weld 3. The systematical observation of metal state is performed through witness-samples and threshold activation detectors ( 54 Fe (n,p), 63 Cu (n,α), 93 Nb (n,n')) placed above the reactor top edge and at the first vessel ring level. There are big differences in energy spectrum and integral neutron flux falling onto the weld 3, the RPV base metal and the staff detectors. This requires additional neutron measurements in the air gap between the RPV and the thermal insulation. (author)

  2. Experience in start-up of the South-Ukrainian-2 power unit with the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Belov, Yu.V.; Kazakov, V.A.; Kirilov, V.V.; Kryakvin, L.V.

    1987-01-01

    The volume, sequence and dates of works fulfilled during hot testing, physical and power start-ups and while bringing output of the South-Ukranian-2 power unit with the WWER-1000 reactor to the design figure are described. The works were fulfilled according to the standard schedules from October in 1984 till April in 1985. Combination of the stages and intensification of works before the physical start-up have allowed to shorten the dates by 90 days as compared to the schedule. The physical and power start-ups, including bringing reactor output to the design figure, were performed during 140 days, that permits to shorten the dates by 20 days more. The results of physical experiments carried out at the South-Ukranian-2 power unit, are in good agreement with the data obtained at the first power units of the given and Kalinin NPPs. Besides, during physical and power start-ups additional measures ensuring nuclear safety are developed

  3. Power release estimation inside of fuel pins neighbouring fuel pin with gadolinium in a WWER-1000 type core

    International Nuclear Information System (INIS)

    Mikus, J.

    2006-01-01

    The purpose of this work consists in investigation of the gadolinium fuel pin (fps) influence on space power distribution, especially from viewpoint of the values and gradient occurrence inside of neighbouring FPs that could result in static loads with some consequences, e.g., FP bowing. Since detailed power distributions cannot be obtained in the NPPs, needed information is provided by means of experiments on research reactors. As for the power release measurement inside of FPs, some special (e.g. track) detectors placed between fuel pellets are usually used. Since such works are relatively complicated and time consuming, an evaluation method based on mathematical modelling and numerical approximation was proposed by means of that, and using measured (integral) power release in selected FPs, relevant information about power release inside of needed (investigated) FP, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized in a WWER-1000 type core with 7 fuel assemblies at zero boron concentration and containing gadolinium FPs. Application of the above evaluation method is demonstrated on investigated FP neighbouring a FP with gadolinium by means of the 1) Azimuthal power distribution inside of investigated FP on their fuel pellet surface in horizontal plane and 2) Gradient of the power distribution inside of investigated FP in two opposite positions on pellets surface that are situated to- and outwards a FP with gadolinium. Similar information can be relevant from the viewpoint of the FP failures occurrence investigation (Authors)

  4. Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes

    International Nuclear Information System (INIS)

    Trgina, M.

    1993-12-01

    The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs

  5. Effect of nickel content on mechanical properties and fracture toughness of weld metal of WWER-1000 reactor vessel welded joints

    Energy Technology Data Exchange (ETDEWEB)

    Zubchenko, A.S.; Vasilchenko, G.S.; Starchenko, E.G.; Nosov, S.I

    2004-08-01

    Welding of WWER-1000 reactor vessel of steel 15X2HMPHIA is performed using the C{sub B}-12X2H2MAA wire and PHI-16 or PHI-16A flux. Nickel content in the weld metal usually lays within the limits 1.2-1.9%. The experimental data is shown on the weld metal with the nickel contents 1.28-2.45% after irradiation with fluence up to 260.10{sup 22}n/m{sup 2} at energy more than 0.5 MEV. The embrittlement was measured by shift of critical brittleness temperature. Has appeared, that the weld metal with the low nickel content is the least responsive to irradiation embrittlement. The mechanical properties and fracture toughness of the weld metal with the contents of a nickel less than 1.3% are studied. Specimens CT-1T are tested, the 'master-curve', and its confidence bounds with probability of destruction 5 and 95% is built. 'Master-curve' in the specified confidence interval is affirmed by CT-4T specimens test data. Is shown, that the mechanical properties and fracture toughness of the weld metal with the contents of nickel less than 1.3% satisfy the normative requirements.

  6. Effect of nickel content on mechanical properties and fracture toughness of weld metal of WWER-1000 reactor vessel welded joints

    International Nuclear Information System (INIS)

    Zubchenko, A.S.; Vasilchenko, G.S.; Starchenko, E.G.; Nosov, S.I.

    2004-01-01

    Welding of WWER-1000 reactor vessel of steel 15X2HMPHIA is performed using the C B -12X2H2MAA wire and PHI-16 or PHI-16A flux. Nickel content in the weld metal usually lays within the limits 1.2-1.9%. The experimental data is shown on the weld metal with the nickel contents 1.28-2.45% after irradiation with fluence up to 260.10 22 n/m 2 at energy more than 0.5 MEV. The embrittlement was measured by shift of critical brittleness temperature. Has appeared, that the weld metal with the low nickel content is the least responsive to irradiation embrittlement. The mechanical properties and fracture toughness of the weld metal with the contents of a nickel less than 1.3% are studied. Specimens CT-1T are tested, the 'master-curve', and its confidence bounds with probability of destruction 5 and 95% is built. 'Master-curve' in the specified confidence interval is affirmed by CT-4T specimens test data. Is shown, that the mechanical properties and fracture toughness of the weld metal with the contents of nickel less than 1.3% satisfy the normative requirements

  7. Modernization in-core monitoring system, of WWER-1000 reactors (W-320) by fuel assemblies with individual characteristics using

    International Nuclear Information System (INIS)

    Mitin, V. I.; Semchenkov, J. M.; Kalinushkin, A. E.

    2007-01-01

    The present report covers object, conception, engineering solution of construction of modern system of high-powered reactor in-core control, including WWER-1000 (V-320) reactors. It is known that ICMS main task is on-line monitoring distribution of power release field and its functioning independently of design programs to avoid common reason error. It is shown in what way field of power release recovery has been carrying on; rest on the signals of in-core neutron and temperature sensors. On the base of the obtained and refined information there have being automatically generated signals of preventive and emergency protection on local parameters (linear power to the maximum intensive fuel elements, reserve to heat exchange crisis, 'picking factor'). There have represented technology on sensors and processing methods of SPND and TC signals, ICIS composition and structure, program hard ware, system and applied software. Structure, composition and the taken decisions allow combining class 1E and class B and C tasks in accordance with international norms of separation and safety classes' realization. At present ICIS-M is a system, providing implementation of control, safety, information and diagnostic functions, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and NPP units. And at the same time it reduces human factor negative influence to core work thermo technical eliability in the operational process (Authors)

  8. Verification results of methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V.; Korolev, Alexander S. [JSC ' ' Atomtechenergo' ' , Novovoronezh (Russian Federation). Novovoronezh Filial ' ' Novovoronezhatomtechenergo' ' ; Ryasny, Sergei I. [JSC ' ' Atomtechenergo' ' , Moscow (Russian Federation)

    2017-09-15

    The JSC ''Atomtechenergo'' experts have developed a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants. The necessity for developing the new methodology was determined by the need to decrease the calculation error of the weighted mean coolant temperature in the hot legs because of the coolant temperature stratification. The methodology development was based on the findings of experimental and calculating research executed by the authors. The methodology verification was fulfilled through comparison of calculation results obtained with and without the methodology use in various operational states and modes of several WWER-1000 power units. The obtained verification results have confirmed that the use of the new methodology provides objective error decrease in determining the weighted mean coolant temperature in the primary circuit hot legs. The decrease value depends on the stratification character which is various for different objects and conditions.

  9. Verification results of methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

    International Nuclear Information System (INIS)

    Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V.; Korolev, Alexander S.

    2017-01-01

    The JSC ''Atomtechenergo'' experts have developed a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants. The necessity for developing the new methodology was determined by the need to decrease the calculation error of the weighted mean coolant temperature in the hot legs because of the coolant temperature stratification. The methodology development was based on the findings of experimental and calculating research executed by the authors. The methodology verification was fulfilled through comparison of calculation results obtained with and without the methodology use in various operational states and modes of several WWER-1000 power units. The obtained verification results have confirmed that the use of the new methodology provides objective error decrease in determining the weighted mean coolant temperature in the primary circuit hot legs. The decrease value depends on the stratification character which is various for different objects and conditions.

  10. Report of a consultants meeting on control rod insertion reliability for WWER-1000 nuclear power plants. Extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1995-09-01

    Starting from 1992, an increased drop time of control rods exceeding the design limit of four seconds has been observed in most of the operating WWER-1000 reactors in Russia and in the Ukraine. In some cases a dropped control rod became stuck in an intermediate position near the bottom of the core. In October 1994, a similar control rod problem was also observed at Unit 6 of the Kozloduy NPP. The issue of control rod insertion reliability was considered at a consultants' meeting on ''Core Control and Protection Strategy of WWER-1000 Reactors'' in April 1994. A consultants' meeting specifically focused on ''Control Rod Insertion Reliability'' was convened in Vienna in February 1995 attended by 15 international experts. The objectives of this meeting were: The exchange of international experience on problems and solutions related to anomalous control rod insertion; judgement of the safety concern of this issue for WWER-1000 reactors based on safety analyses; consideration of regulatory requirements and interim measures to continue operation in short term including modifications implemented or planned; and, status of root cause analyses and pending problems. The technical discussions were held in plenary sessions and in three working groups devoted to specific aspects of the issue. Refs, figs, tabs

  11. Comparative severe accident analysis of WWER 1000/B 320 LOCA DN100 computed by computer codes ASTEC V1.1 and SCDAP/RELAP5

    International Nuclear Information System (INIS)

    Kalchev, B.; Dimov, D.; Tusheva, P.; Mladenov, I.

    2005-01-01

    This paper presents the modelling approach for LOCA 100 mm sequence for WWER 1000-B 320 type of reactor with the integral ASTEC computer code and SCDAP/RELAP5 computer code. As a basic input deck the reference input file for Balakovo NPP from the released ASTEC CD has been applied. As a first part of the calculations for the SBLOCA sequence the ASTEC v1.1 modules CESAR, DIVA and CPA have been activated in a coupled mode. For SCDAP/RELAP5 calculation input deck for WWER 1000-B 320 has been applied which meant to be closer to the initial boundary conditions applied for ASTEC WWER 1000 input deck. A SBLOCA 100 mm comparison between ASTEC v1.1 and SCADAP/RELAP5 has been presented. ASTEC predicts vessel failure at 15620 s. ASTEC and SCDAP/RELAP5 give close but not similar results - this could be observed on the trends. The comparison of 100 mm-break shows that SCDAP/RELAP5 predicts clear phenomenological changes in primary pressure evolution and molten pool formation. Similar hydrogen production mass for both codes around 5000 s is detected

  12. Report of the review of WWER-1000 safety issues resolution at Temelin nuclear power plant, Temelin, Czech Republic 11 to 15 March 1996

    International Nuclear Information System (INIS)

    Almeida, C.; Hoehn, J.; Seiberling, R.; Chambon, J.L.; Fil, N.S.; Munoz, A.; Roennberg, G.; Wenk, W.

    1996-01-01

    At the request of the Government of the Czech Republic, the IAEA conducted, in the period of 11-15 March 1996, a mission to review the resolution of WWER-1000 safety issues at Temelin NPP. These safety issues have been identified for WWER-1000 model 320 reactors and ranked according to their importance to safety in the frameworks of the IAEA Extrabudgetary Programme on Safety of WWER and RBMK Nuclear Power Plants. The Temelin NPP is a WWER-1000 and was originally designed according to standards of the former Soviet Union. After a series of reviews in the 1980s, a decision was taken by the Temelin NPP management to upgrade the design of Temelin, including the supply of fuel and instrumentation and control by a western company. The objective of the mission was to review the response of Temelin to the safety issues identified by the IAEA. The mission assessed the current Temelin design, including proposed modifications and plans for operation at Temelin, in the light of the IAEA recommendations for each relevant issue. The present report contains the mission's general conclusions and recommendations and an overview of the review performed in each topical area. The attachment contains a brief summary of the discussions on each individual safety issue and associated conclusions and recommendations. 3 refs

  13. WWER-1000 steam generator integrity. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1997-07-01

    Programme was initiated by IAEA in 1990 with the aim to assist the countries of Central and Eastern Europe and former Soviet Union in evaluating the safety of their first generation WWER-440/230 nuclear power plants. The main objectives were: to identify major design and operational safety issues; to establish international consensus on priorities for safety improvements; and to provide assistance in the review of the competence and and adequacy of safety improvement programs. The scope was extended in 1992 ro include RBMK, WWER-440/312 and WWER-1000 plants in operation and under construction. Based on the operational experience of more than 90 reactor years of WWER-1000 NPPs having 80 steam generators in operation or under construction the steam generator integrity was recognized as an important issue of high safety concern. The purpose of this report is to integrate available information on the issue of WWER-1000 steam generator integrity with the focus on the steam generator cold collector damage in particular. This information covers the status of stem generators at operating plants, cause analysis of collector cracking, the damage mechanisms involved, operational aspects and corrective measures developed and implemented. Consideration is given to material, design and fabrication related aspects, operational conditions, system solutions, and in-service inspection. Detailed conclusions and recommendations are provided for each of these aspects

  14. Seismic evaluation and upgrading design of overhead roads between reactor buildings of WWER-1000 MW type NPP

    International Nuclear Information System (INIS)

    Jordanov, M.J.; Stoyanov, G.S.; Geshanov, I.H.; Kirilov, K.P.; Schuetz, W.

    2003-01-01

    This paper presents results obtained during the study of overhead roads between Reactor Building (RB) of WWER-1000 MW NPP and possible measures for their seismic upgrade. The main objective of this project is to evaluate the behavior of overhead roads under site-specific seismic loading and to determine whether this structure satisfies current international safety regulations, followed by development of upgrading concepts. Overhead roads are pre-cast RC structure, which can be divided to separate substructures. They comprise of pedestrian gallery and pipeline box, connecting reactor buildings with auxiliary building. They are mounted at approximately 10 m above ground level. The overhead roads are evaluated for Review Level Earthquake (RLE) as seismic category II structures. As seismic input motion is RLE, free field response spectra anchored to 0.2 g PGA are used with 0.5 scaling factor. Soil-Structure Interaction effects are taken into account through equivalent soil springs with frequency adjusted stiffness. In order to meet the objective of the project a technical design specification is developed for conformance with International, US and Bulgarian standards and codes, taking into account site specific conditions. The general approach is consistent with up-to-date practice for evaluation and upgrade of nuclear power plant facilities. The separate steps comprising the overall fulfillment of project's major objectives may be summarized as follows: study of all available data for initial design and as built conditions, creation of 3-D detailed finite element models for as-built structure, determination of dynamic characteristics, evaluation of adequacy of initial design under new seismic loading (calculation of D/C ratios for structural members and connections, evaluation of embedment lengths for embedded parts and rebars, deformation evaluation, stability checks), development of upgrading concepts for enhancement, verification of capability of upgraded structure

  15. Development of Generic SAM Strategies and Evaluation of their Implementability at Operating WWER-1000/V-320 Power Units

    International Nuclear Information System (INIS)

    Podshibiakin, A.K.; Fil, N.S.; Kuznetsova, T.E.; Krushelnitsky, V.N.; Peressadko, V.G.; Korshunov, A.S.

    2001-01-01

    requirements give, although implicitly, the certain directions with regard to approac h to be used when developing the BDBA management measures. The BDBA management measures shall include the following: - determination of the operational objectives in order to prevent or to terminate core melting and to mitigate radioactive release into the environment; - determination of systems and equipment on the NPP site and outside it that could be used to achieve the operational objectives; - development of the strategies for personnel's actions to achieve the safety objectives. It's worth noting that the accident management strategies should foresee the usage of all the available systems and components (including systems and components not related to safety) that could be applied not in accordance with design purpose and under off-design operational conditions. This paper presents: the Current status of AM at operating WWER-1000 (Instruction on accident liquidation, Guideline on BDBA management, Upgrading the BDBA management); the approach to generic SAM strategy development, the evaluation of generic strategies for operating WWER-1000, the severe accident mitigation in advanced WWER-1000/V-392 design (NVAES-2 design, Hydrogen control, Containment function provisions)

  16. WWER-1000 dominance ratio

    International Nuclear Information System (INIS)

    Gorodkov, S.S.

    2009-01-01

    Dominance ratio, or more precisely, its closeness to unity, is important characteristic of large reactor. It allows evaluate beforehand the number of source iterations required in deterministic calculations of power spatial distribution. Or the minimal number of histories to be modeled for achievement of statistical error level desired in large core Monte Carlo calculations. In this work relatively simple approach for dominance ratio evaluation is proposed. It essentially uses core symmetry. Dependence of dominance ratio on neutron flux spatial distribution is demonstrated. (Authors)

  17. Calculation of isotope burn-up and change in efficiency of absorbing elements of WWER-1000 control and protection system during burn-up

    International Nuclear Information System (INIS)

    Timofeeva, O.A.; Kurakin, K.U.

    2006-01-01

    The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)

  18. Lessons learned from the computational simulation of thermomechanical behaviour of the WWER-1000 reactor cores: FA development and its implantation into the Balakovo NPP unit 1

    International Nuclear Information System (INIS)

    Troyanov, V.; Likhatchev, Y.; Folomeev, V.; Demishonkov, A.

    2003-01-01

    The fuel assembly is forced under operation conditions by weight and hydraulic force; hold-down force; contact transversal forces; non-uniform thermal and neutron fields; thermomechanical forces in the fuel assembly multirod structure. The fuel assembly bowing has been investigated in this work. The objective of the study is to recommend the optimal design for fuel assemblies. Calculations are made using the code TEREMOK. A new generation fuel assemblies for WWER-1000 named FA-2 are suggested. The performance of the FA- 2 has been studied. The theoretical approach and the calculation technique have been verified

  19. Modeling of coherent heat and mass-exchange phenomena for a specter of primary circuit leaks in a NPP with WWER-1000 type reactor

    International Nuclear Information System (INIS)

    Stanev, I.; Mladenova, S.; Hadjiev, V.

    2004-01-01

    Subject of study are the inter-relations between the primary leak parameters, the containment pressure and the functioning of the WWER-1000/V320 plant systems. The model includes systems for normal operation and safety systems, which are actuated in the initial stage of the accident. The studied specter of primary leaks includes diameters from 10 to 60 mm. The model is based on the design features of unit 5 in NPP Kozloduy. This paper presents a fragment of the work, performed jointly by Atoma Consult Ltd and NPP Kozloduy Plc on the validation of the symptom oriented emergency operation procedures for units 5 and 6 in NPP Kozloduy. (authors)

  20. Calculations of fission rate distribution in the core of WWER-1000 mock-up on the LR-0 reactor using alternative methods and comparison with results of measurements

    International Nuclear Information System (INIS)

    Zaritskiy, S.; Kovalishin, A.; Tsvetkov, T.; Rypar, V.; Svadlenkova, M.

    2011-01-01

    General review of experimental and calculation researches on WWER-440 and WWER-1000 mock-ups on the reactor LR-0 was introduced on the twentieth Symposium AER. The experimental core fission rate distribution was obtained by means of gamma-scanning of the fuel pins - 140La single peak (1596 keV) measurements and wide energy range (approximately 600-900 keV) measurements. Altogether from 260 to 500 fuel pins were scanned in different experiments. The measurements were arranged in the middle of the fuel (the active part of pin). Pin-to-pin calculations of the WWER-1000 mock-up core fission rate distribution were performed with several codes: Monte Carlo codes MCU-REA/2 and MCNPX with different nuclear data libraries, diffusion code RADAR (63 energy groups library) and code SVL based on Surface Harmonics Method (69 energy groups). Calculated data are compared with experimental ones. The obtained results allow developing the benchmark for core calculations methodologies, evaluating and validating source reliability for the out-of-core (inside and outside pressure vessel) neutron transport calculations. (Authors)

  1. RELAP5/MOD3.2 investigation of loss of in-house supply power for WWER 1000/320V

    International Nuclear Information System (INIS)

    Gencheva, R.; Pavlova, M.; Groudev, P.

    2001-01-01

    This paper discusses the results of the thermal-hydraulic investigations of the 'Loss of in-house supply power' accident at the Kozloduy NPP Unit 6. The RELAP5/MOD3.2 computer code has been used to stimulate the loss of in-house supply power accident in a WWER 1000 Nuclear Power Plant model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP. The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The investigation of 'Loss of normal and reverse AC power' is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the WWER 1000 against experimental transient data obtained from Kozloduy NPP Unit 6. The comparisons between the RELAP5 results and the test data indicate good agreement

  2. Characteristics of WWER-1000 fuel rod claddings and FA components from E635 alloy at burnups up to 72 MWd/kgU

    International Nuclear Information System (INIS)

    Nikulin, A.; Novikov, A.; Peregud, M.; Shishov, V.; Shevyakov, A.; Volkova, I.; Novoselov, A.; Kobylyansky, G.

    2011-01-01

    In this paper operation experience, results of investigated E365 alloy components of Balakovo NPP Unit 1 and Kalinin NPP unit 1 fuel assemblies are presented. Appearance, shape changes and geometric size, corrosion state of guide thimbles, angles and fuel rods, corrosion of fuel claddings are studied. At the end authors concluded that: I) E635 alloy corroborated its high operation reliability as fuel claddings and WWER-1000 FA components during 6 year service to the fuel burnup of 72MWd/kgU; II) Based on the results from the post-irradiation investigations of the fuel rods and other structural elements of WWER-1000 FAA, fabricated from E635 alloy, in terms of the basic operational characteristics, their resources after the 6 year operation cycle have not been exhausted; III) The geometrical parameters, corrosion states, tensile properties of items fabricated from fuel alloy did not attain the values that would prevent their further operation: 1) the elongations of the fuel rods at the mean burnups up to 66.2 MWd/kgU do not exceed 15 mm or 4.9%; 8) the amount of the oxide coat at surface of GT and CT does not exceed 45 μm, the hydrogen content is <0.03% mass; 9) the oxide coat at the surfaces of the frame angles does not exceed 50 μm, the hydrogen content is <0.04% mass

  3. Concept for creating program-technical complex of safety monitoring with system of safety parameters presentation functions on the basis of routine WWER-1000 systems

    International Nuclear Information System (INIS)

    Dunaev, V.G.; Tarasov, M. V.; Povarov, P.V.

    2005-01-01

    Prerequisites of creating the software-hardware complex for reactor safety monitoring on the Volgodonsk NPP are analyzed and generalized. The concept of this complex is based on functions of the safety parameters presentation system. It will serve as an interface between operator and technological process and give to operator a possibility to estimate quickly the state of the safety of the nuclear power unit. The complex will be created on the basis of routine reactor monitoring and control systems intended for the WWER-1000 reactor. In addition to existing soft- and hard-wares for reactor monitoring and for analysis of technological archive, it is proposed to create and connect in parallel the new software-hardware complex which ensures calculation and presentation of generalized factors of reactor safety [ru

  4. Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

    International Nuclear Information System (INIS)

    Kolev, N.I.; Sybotinov, L.S.

    1984-01-01

    The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

  5. Nuclear safety analysis for transport cask TK-6 (for WWER-440) and cover for fresh assemblies (for WWER-1000) in implementation of new fuel types at Ukrainian NPP

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kovbasenko, Iu; Dudka, Olena

    2006-01-01

    According to the fresh fuel management procedure, fuel assemblies - after nuclear fuel delivery to the NPP fresh fuel unit - are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center 'Kurchatov Institute' for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction. Transport container TK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading (Authors)

  6. Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

    2017-03-15

    In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  7. Simulation of protected and unprotected loss of flow transients in a WWER-1000 reactor based on the Drift-Flux model

    Energy Technology Data Exchange (ETDEWEB)

    Baghban, Ghonche [Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of). Nuclear Science and Technology Research Inst.; Shayesteh, Mohsen [Imam Hussein Univ., Tehran (Iran, Islamic Republic of). Dept. of Physics; Bahonar, Majid [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2017-03-15

    In view of the importance of studying coolant transient behavior in a nuclear reactor, this work is devoted to the thermal-hydraulic analysis of protected and unprotected loss of flow transients in a WWER-1000 reactor. A series of corresponding mathematical and physical models based on the four-equation Drift-Flux model has been applied. Based on a multi-channel approach, the core has been divided into different regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. Appropriate initial and boundary conditions have been considered and two situations of tripping four and two primary pumps in a protected core in addition to situation of tripping all four pumps in an unprotected core have been analyzed. For each transient, a full range of thermal-hydraulic parameters has been obtained. For verification of the proposed model, the results have been compared with those of the RELAP5/MOD3 and Bushehr nuclear power plant Final Safety Analysis Report (FSAR). A good agreement between results has been attained for the aforementioned transients.

  8. Start up physics tests of units 5 and 6 (WWER 1000) at Kozloduy NPP by comparison with the calculated neutron physics characteristics

    International Nuclear Information System (INIS)

    Antov, A.; Stoyanova, I.

    2008-01-01

    In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests, which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies (FAs). The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6. (authors)

  9. Start up physics tests of Units 5 and 6 (WWER-1000) at Kozloduy NPP by comparison with the calculated neutron physics characteristics

    International Nuclear Information System (INIS)

    Stoyanova, I.; Antov, A.

    2007-01-01

    In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies. The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6 (Authors)

  10. Safety issues and their ranking for 'small series' WWER-1000 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

  11. Safety issues and their ranking for 'small series' WWER-1000 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    2000-09-01

    This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

  12. Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  13. Blind pre-analysis of the main building complex WWER-1000 Kozloduy. Comparison of analytical and experimental results obtained by explosive testing (task 8a of workplan 96/97)

    International Nuclear Information System (INIS)

    Krutzik, N.J.

    1999-01-01

    In accordance with the 96/97 workplan of the Research Programme on 'Benchmark Studies for Seismic Analysis and Testing of WWER-Type Nuclear Power Plants', blind pre analyses were prepared for the main building complex of the WWER-1000 based on given excitations derived from explosive tests. The investigations were performed by several institutions based on various mathematical models and procedures for consideration of soil-structure interaction effects, but on the same explosive test input data recently obtained. The methods of calculation and software tools used will also be different. The aim of this investigation is to validate different idealization concepts (mathematical models for the idealization of the structures and the soil) as well as investigation procedures (time domain and frequency domain analysis) and finally the software tools by comparing structural response results (time histories and response spectra). This report contains the results of the blind pre analysis performed by Siemens using an equivalent beam model of the main building of the WWER 1000. The calculations were performed by means of a frequency domain calculation

  14. Comparative calculations and operation-to-PIE data juxtaposition of the Zaporozhye NPP, WWER-1000 FA-E0325 fuel rods after 4 years of operation up to ∼49 MWd/kgU burnup

    International Nuclear Information System (INIS)

    Passage, G.; Stefanova, S.; Scheglov, A.; Proselkov, V.

    2006-01-01

    Operational and PIE data for the Zaporozhe NPP, FA-E0325, WWER-1000 fuel rods were provided in the OECD NEA IFPE Database and were used to perform comparative calculations among several fuel performance codes. The fuel rods had been irradiated for 4 years of operation up to ∼49 MWd/kg U burnup. The fuel rod operation histories are developed for the PINw99, TRANSURANUS (V1M1J03) and TOPRA-2 codes. The initial state fuel rod parameters are analysed and calculations are carried out. The PIE data enable the comparison of experimental measurement with code-calculated values for cladding elongation (49 rods), FGR and gas pressure (35 rods). Cladding diameter creep-down and gap closure results are juxtaposed as well. The capability of the applied codes correctly to predict the WWER fuel rod performance is shown. The WWER-1000 fuel rod data include initial geometrical and design parameters of the fuel rods, as well as description of the operation regime, NPP unit loading history and PIE results at normal conditions. The data are sufficient for modelling all 312 fuel rod and for comparison of calculations with experimental results for a limited number of fuel rods. The comparison between the calculated and measured results discussed in this paper shows that the codes PINw99, TRANSURANUS and TOPRA-2, are capable of adequate predicting the thermophysical and the mechanical performance of the WWER-1000 fuel rods. The PINw99 code predicts conservative BOL FGR values and conservative gas pressure values in the region of burnups higher than 30 MWd/kg U, which can be explained by the underprediction of the cladding gas inner volume and cladding elongation. The improved version PIN2K (not applied in the present study) predicts much better FGR and gas pressure, though, it is still under development in the high burnup FGR modelling part. In the TRANSURANUS code, there are also areas, where refinements are clearly indicated. They are subjects of the ongoing research projects and

  15. Improving the refueling cycle of a WWER-1000 using cuckoo search method and thermal-neutronic coupling of PARCS v2.7, COBRA-EN and WIMSD-5B codes

    Energy Technology Data Exchange (ETDEWEB)

    Yarizadeh-Beneh, M.; Mazaheri-Beni, H.; Poursalehi, N., E-mail: n_poursalehi@sbu.ac.ir

    2016-12-15

    Highlights: • The cuckoo search algorithm is applied to the loading pattern optimization of a nuclear reactor core. • Calculations during the cycle show a good agreement between results and reference for the original LP. • Results indicate the efficient performance of cuckoo search approach coupled with thermal-neutronic solvers. • Neutronic parameters of proposed core pattern are improved relative to original core pattern. - Abstract: The fuel loading pattern optimization is an important process in the refueling design of a nuclear reactor core. Also the analysis of reactor core performance during the operation cycle can be a significant step in the core loading pattern optimization (LPO). In this work, for the first time, a new method i.e. cuckoo search algorithm (CS) has been applied to the fuel loading pattern design of Bushehr WWER-1000 core. In this regard, two objectives have been chosen for finding the best configuration including the improvement of operation cycle length associated with flattening the radial power distribution of fuel assemblies. The core pattern optimization has been performed by coupling the CS algorithm to thermal-neutronic codes including PARCS v2.7, COBRA-EN and WIMSD-5B for earning desired parameters along the operation cycle. The calculations have been done for the beginning of cycle (BOC) to the end of cycle (EOC) states. According to numerical results, the longer operation cycle for the semi-optimized loading pattern has been achieved along with less power peaking factor (PPF) in comparison to the original core pattern of Bushehr WWER-1000. Gained results confirm the efficient and suitable performance of the developed program and also the introduced CS method in the LPO of a nuclear WWER type.

  16. Report of the review of the safety improvement programme for South Ukraine NPP units 1 and 2 and to identify the safety issues of ''small series'' WWER-1000 NPPs. South Ukraine Yuzhnoukrainsk, Nikolaev Region Ukraine, 8 to 19 July 1996. Draft

    International Nuclear Information System (INIS)

    Bastin, S.; Hoehn, J.; Lin, C.; Taylor, R.; Benitez, F.; Dale, H.; Mueller, B.; Rieg, C.Y.

    1996-10-01

    According to the Ukrainian request the purpose of the IAEA experts' mission was to review the safety improvement programme for South Ukraine NPP Units 1 and 2 in order to advise on the completeness and adequacy of safety improvements implemented and/or proposed. Another purpose of the mission was to identify major design and operational deficiencies as a basis to compile a consolidated list of generic safety issues for the units of the 'small series'' of WWER-1000 reactors (''Issue Book for ''small series'' WWER-1000 NPPs). Conclusions and recommendations from the IAEA mission are based on the combined expertise of the international group of experts who composed the team. They are intended to assist national authorities and plant operators who have the sole responsibilities for the regulation and safe operation. tabs

  17. Analysis of WWER 1000 collector cracking mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Matocha, K.; Wozniak, J. [Vitkovice J.S.C., Ostrava (Switzerland)

    1997-12-31

    The presentation reviews the large experimental program, started in 1993 in Vitkovice, where the main aim was: (1) a detailed study of strain and thermal ageing, dissolved oxygen content and temperature on subcritical crack growth in 10NiMo8.5 (10GN2MFA) steel, (2) a detailed study of the effect of high temperature water and tube expansion technology on fracture behaviour of ligaments between holes for heat exchange tubes, and (3) a detailed study of the effect of drilling, tube expansion technology and heat treatment on residual stresses on the surface of holes for heat exchange tubes. The aim of all these investigations was to find a dominant damage mechanism responsible for collector cracking to be able to judge the efficiency of implemented modifications and suggested countermeasures and to answer a very important question whether proper operation conditions (mainly water chemistry) make the operation of steam generators made in Vitcovice safe throughout the planned lifetime. 10 refs.

  18. Analysis of WWER 1000 collector cracking mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Matocha, K; Wozniak, J [Vitkovice J.S.C., Ostrava (Switzerland)

    1998-12-31

    The presentation reviews the large experimental program, started in 1993 in Vitkovice, where the main aim was: (1) a detailed study of strain and thermal ageing, dissolved oxygen content and temperature on subcritical crack growth in 10NiMo8.5 (10GN2MFA) steel, (2) a detailed study of the effect of high temperature water and tube expansion technology on fracture behaviour of ligaments between holes for heat exchange tubes, and (3) a detailed study of the effect of drilling, tube expansion technology and heat treatment on residual stresses on the surface of holes for heat exchange tubes. The aim of all these investigations was to find a dominant damage mechanism responsible for collector cracking to be able to judge the efficiency of implemented modifications and suggested countermeasures and to answer a very important question whether proper operation conditions (mainly water chemistry) make the operation of steam generators made in Vitcovice safe throughout the planned lifetime. 10 refs.

  19. Steam turbine of WWER-1000 unit

    International Nuclear Information System (INIS)

    Drahy, J.

    1986-01-01

    The manufacture was started by Skoda of a saturated steam, 1,000 MW, 3,000 rpm turbine designed for the Temelin nuclear power plant. The turbine provides steam for heating water for district heating, this either with an output of 893 MW for a three-stage water heating at 150/60 degC, or of 570 MW for a two-stage water heating at 120/60 degC. The turbine features one high-pressure and three identical low-pressure stages. The pressure gradient between the high-pressure and the low-pressure parts was optimized with respect to the thermal efficiency of the cycle and to the thermodynamic efficiency of the low-pressure part. A value of 0.79 MPa was selected corresponding to the maximum through-flow of steam entering the turbine. This makes 5,495 t/h, the admission steam parameters are 273.3 degC and 5.8 MPa. The feed water temperature is 220.9 degC. 300 cold starts, 1,000 starts after shutdowns for 55 to 88 hours and 600 starts after shutdown for 8 hours are envisaged for the entire turbine service life. (Z.M.). 5 figs., 1 tab., 6 refs

  20. Optimized core loading sequence for Ukraine WWER-1000 reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Fuel Assemblies (WFAs) experienced mechanical damage of the grids during loading at both South Ukraine 2 (SU2) and South Ukraine 3 (SU3). The grids were damaged due to high lateral loads exceeding their strength limit. The high lateral loads were caused by a combination of distortion and stiffness of the mixed core fuel assemblies and significant fuel assembly-to-fuel assembly interaction combined with the core loading sequence being used. To prevent damage of the WFA grids during core loading, Westinghouse has developed a loading sequence technique and loading aides (smooth sided dummies and top nozzle loading guides) designed to minimize fuel assembly-to-fuel assembly interaction while maximizing the potential for successful loading (i.e., no fuel assembly damage and minimized loading time). The loading sequence technique accounts for cycle-specific core loading patterns and is based on previous Westinghouse WWER core loading experience and fundamental principles. The loading aids are developed to “open-up” the target core location or to provide guidance into a target core location. The Westinghouse optimized core loading sequence and smooth sided dummies were utilized during the successful loading of SU3 Cycle 25 mixed core in March 2015, with no instances of fuel assembly damage and yet still provided considerable time savings relative to the 2012 and 2013 SU3 reload campaigns. (authors)

  1. Modernization incore monitoring system of WWER-1000 reactors (V-320)

    International Nuclear Information System (INIS)

    Mitin, Valentin; Semchenkov, Yurij; Kalinushkin, Andrey

    2008-01-01

    Modern ICIS system for VVER-1000, including a number of sensors, cable runs, corresponding measuring equipment and computer engineering, software, accumulated 30 year experience of interaction researches on VVER reactors and is capable to ensure carrying out of control, protection, informational, diagnostic functions and thus to promote real increase of quality, reliability and safety in nuclear fuel and NPP power units operation

  2. WWER-1000/320 steam generator collector rupture. Radiological consequences

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova, A; Sartmadzhiev, A; Balabanov, E [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    A model describing a hypothetical accident with direct release of primary coolant to the atmosphere is proposed. Cover lifting of the primary collector due to a rupture of the fixing bolts leads to a coolant release. The initial and boundary conditions of the accident scenario have been selected to provide for the most unfavorable conditions. The total release of primary coolant during the first 15 min of transient are estimated to 50.8 tons, of these 48.5 t with the initial activity in the primary coolant circuit. Without evacuation or sheltering, after 7 days of exposure, the expected dose at the boundary of the restricted zone is 0.0182 Sv for the whole body and 0.184 Sv for the thyroid gland. The effective equivalent dose on the site would be 0.0521 Sv. As a result of the analysis it is concluded that the steam generator collector rupture is not jeopardizing the core heat removal even with a minimum configuration of ECCS as the cooling is accomplished through the steam generators. The radiological consequences of the accident would be relatively small if an emergency procedure is applied at the 15-th minute of the transient. 1 ref.

  3. Dynamic response of Belene WWER-1000 to seismic loading conditions

    International Nuclear Information System (INIS)

    Krutzik, N.J.; Petrovski, D.; Sachanski, S.

    1993-01-01

    Within the framework of investigating of the capacity of the WWER-100 at the Belene site, an analysis was performed using revised seismic input data as well as two alternative foundation concepts (natural soil and soil exchange). The starting point for the analysis was the development of a suitable model of the coupled structures (base building, external building, containment, internal structure) and soil taking into account the real properties of the originally layered as well as the exchanged soil. The soil-structure effects were considered according to the analytical method employed, either through soil impedance (substructure method) or explicitly by a complex (direct method). On the basis of the results obtained by the two methods (substructure and direct method) the seismic safety of the complex structures for different foundation concepts was evaluated. By comparing the calculated structural response with the design spectra originally used for the design of components and systems the available safety margin was estimated

  4. New methods of building WWER 1000 nuclear power plants

    International Nuclear Information System (INIS)

    Tadzher, O.

    1983-01-01

    The shortening of construction time needed for building the 5th unit of the Kozloduj nuclear power plant in Bulgaria is discussed. In the 2nd to 4th stages of construction the builders are expected to use the large CC-4000 mobile crane with high lifting power (Demag, FRG) for construction and assembly work. This work may be done by ''open'' assembly, i.e., the assembly of equipment in an unfinished building. By using the said crane it is possible to increase the weight of the individual units and thereby to reduce their number. The advantages of ''open'' assembly is that there is no need for a opecial transport corridor for the lifting and assembly of heavy handling units. A preliminary evaluation of the new technology has shown that assembly of the reactor hall equipment can be started 6 to 8 months ahead of schedule. (E.S.)

  5. WWER-1000 simulator instructor service in NV TC

    International Nuclear Information System (INIS)

    Pogrebitsky, S.L.

    1997-01-01

    In July 1996 a full-scope simulator developed by the joint efforts of ATOMTECHENERGO, VNII AES (Russia) and Mitsubishi Heavy Industries (Japan) was put into service Novovoronezh Training Centre (NV TC). this paper describes the Instructor Station equipment and its capabilities for training process monitoring and simulation. (author)

  6. Electrode for welding steel for WWER-1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Lakatos, L.

    Of two types of electrodes, ie., with an alloyed core and with an unalloyed core, an electrode was chosen consisting of a basic coat and an unalloyed core. Fluctuations are shown of shear strength, tensile strenght and contraction with the welding mode and annealing temperature. It was found that pre-heating to 250 and 350 degC, respectively, was most suitable for welding a pressure vessel manufactured from material designated SKODA A3/II. Annealing aimed at removing stress was chosen at 650 to 700 degC. (H.S.)

  7. Risk based definition of TS requirements for NPPs with WWER-1000 type reactor

    International Nuclear Information System (INIS)

    Morozov, V.; Tokmachev, G.

    2000-01-01

    The main regulations in safety related maintenance for NPPs in Russia are defined as a part of Technical Specifications (TSs). It includes limiting conditions for operation (surveillance requirements, allowed outage time, et.). In Russian practice the two levels of TSs are presented: general TSs that have been established as a master documents for similar designed NPPs and plant specific based on operation practice of each NPP unit. This paper presents a brief review of submissions to TS changes for NPPs with WWER type reactor were issued by AEP PSA team since 1988 year. Besides it provides an approach allows to estimate the complex affect on plant risk for both Limiting Conditions of Operation (LCO) and Surveillance Test Intervals (STI) based on relevant probabilistic tool (Minimal Cut Sets method and Marcov Chains methods). (author)

  8. Frequency response analysis of NPP containment with WWER-1000 type reactor

    International Nuclear Information System (INIS)

    Zoubkov, D.; Isaikin, A.; Shablinsky, G.

    2005-01-01

    Analysis of vibration load on the building structures of nuclear power plants (NPP) has come to the fore due to extension of their operational lifetime. Such analysis could be carried out if dynamic characteristics test (natural vibration frequency and natural mode) of NPP building structures is conducted in full-scale conditions. In this paper we represent methods of frequency response analysis of the generating unit 1 (RO-1) reactor compartment at Kalininskaya NPP on the basis of technogenic vibration. Main vibration sources of RO-1 are turbine-type generators, main centrifugal pumps and pipelines of power generating units 1 and 2. Vibration of RO-1 has been measured alternately at three points of RO-1 containment which is 76 m high and 47.4 m in diameter: top of the dome (76.0 m), holdout ring (70.3 m), point on the ground next to containment (0.0 m). Three components of vibration velocity have been measured simultaneously at each point: vertical Z-component, horizontal Y-component along the axis of the apparatus room and horizontal X-component across the room axis. Magnetoelectric pendulum-type vibrometers have been used for measurement. They were modernized by ad hoc multiposition amplifier card installed into the sensor body. Vibrometers were connected to the recorder by vibration protected and jam resistant cables. The results of present researches testified that dominant frequencies of X- and Y-oscillation spectra at RO-1 of Kalininskaya NPP (1.7 and 1.9 Hz correspondingly) correspond to the first vibration mode of RO-1 as a rigid construction on elastic foundation. High peaks of spectra at 16.7 and 25 Hz result from vibrations caused by main centrifugal pumps and turbine-type generators and coincide with the number of their revolutions per minute (1000 and 1500 rpm). Other peaks of spectra are related to the vibration of pipelines of primary and secondary circuits. (authors)

  9. Application of risk-informed approaches for optimization of control of WWER 1000 pipe metal

    International Nuclear Information System (INIS)

    Kolykhanov, Viktork; Komarov, Yuriy; Skalozubov, Volodymyr; Kovryzhkin, Yuriy

    2007-01-01

    According to Ukrainian regulations, the periodic control of pipes by non-destructive methods has to be performed on a precise basis depending on the degree of influence of a system on the nuclear power plant safety. In order to improve control programs of pipes metal, risk-informed approaches have been introduced. The characteristics of a pipe (or part of a pipe) has been assessed according 4 aspects: the influence on reactor safety, the influence on the reliability of power generation, the influence on residual resource (the resource is worked-out or not), and the absence/presence of defects in the pipe. The above assessment for a pipe or part of a pipe leads to one of the 4 levels of operational metal controls. Level 0: metal control is not carried out (the pipe is fixed when a failure appears), level 1: partial metal controls are performed, level 2: operational metal controls are performed, and level 3: extended metal controls are performed. This new approach has been applied to the pipes involved in the high pressure emergency core cooling system of the VVER-1000

  10. Analysis of reactor installation accident WWER-1000 caused by collector head rupture of steam generator

    International Nuclear Information System (INIS)

    Sviridenko, I.I.

    2005-01-01

    System classification of passive heat defence of nuclear dangerous objects with usage of heat pipes. The article is continuation of series of works devoted to research and development of promising passive cooling system of nuclear dangerous objects which use evaporating-condensing heat transfer devices of closed type - heat pipes and two-phase thermosyphons. Classification of autonomous systems of passive heat defence with usage low temperature heat tubes is given. Base classificational lectures are observed. Advantages n disadvantages of various circuit (scheme) solutions are analyzed

  11. A new diagnostic concept for the primary circuit of WWER 1000 reactors

    International Nuclear Information System (INIS)

    Streicher, V.; Liska, J.

    1993-01-01

    The new concept developed by the Skoda and Siemens companies is based on their own experience, the requirements of the Czech Power Board/NPP Temelin Diagnostic Department, and on the research work of various institutes in Czechoslovakia. The development of a complex diagnostic concept is an iterative process and includes parts with different stages of experience and different goals. They can be divided into five groups: PC-based diagnostic systems with continuous or periodic functions, equipment for condition monitoring, PC-based data collectors, studies, analyses and measures which are mandatory for the correct performance and interpretation of the diagnostic systems, and the integration of all subsystems and data acquisition/evaluation units into a Local Area Network with a graphic workstation. (Z.S.) 2 figs., 4 refs

  12. The Heat Exchanger for Passive Part ECCS of WWER-1000 on Base of the Thermo siphons

    International Nuclear Information System (INIS)

    Kirov, V.; Chulkin, O.

    2008-01-01

    One of NPP's systems providing safe operation is the system of emergency core cooling system (ECCS), which primary function in accidents is to flood the nuclear reactor core and to assure the sub critical condition and core cooling. At injection of cold water in reactor thermal stresses and thermal fatigue in the vessel cladding and constructional materials are arise. Low temperature of the water injected in reactor is a reason of occurrence of these undesirable consequences. Some variants of the water heating in accumulators of ECCS are considered. Now at Ukrainian NPPs the electrical heating in accumulators is used. Electrical heaters create the essential additional loading to diesel generators at imposing of two accidents - the large break and losses of power supplies on own needs. It is offered to use a heater in accumulators that working by a principle two-phase thermal siphon which advantages is: small dimensions, small delay and design reliability. In such heat exchanger the heating medium is a direct steam and the heated up medium is water with boric acid from accumulators of ECCS. Under requirements of the service regulations of ECCS accumulators it is necessary to guarantee injected water heating up to 90 ?? in case of a small break and to 150 ?? in case of the large break. Results of calculations for different external diameters of a tube of thermal siphon which have allowed to define the constructive sizes of heat exchanger, providing necessary conditions for required functioning of passive part ECCS are submitted The calculation and analysis of operating modes of the changed circuit of passive part ECCS for various accidents is carried out. The calculated pressure drop indicates that changes do not have essential influence on system work as a whole. Thus, the submitted decision provides the increase of reliability of ECCS at small and large breaks accidents, i.e. in all modes stipulated by the project.(author)

  13. Experimental study of new generation WWER-1000 fuel assemblies at JSC NCCP

    International Nuclear Information System (INIS)

    Enin, A.; Rozhkov, V.; Sinikov, Y.; Ustimenko, A.; Shustov, M.

    2003-01-01

    An experimental program for the study of fuel assembly thermomechanical stability has been established together with RF SSC IPPE and Russian Scientific Center Kurchatov Institute. Assembly fragments and small dummy models of fuel assembly skeletons and fuel rod bundles have been used for the tests. The test results are used for the design selection, verification of the design codes and substantiation of operating capacity of fuel assemblies with a rigid skeleton. The mechanical characteristics of units make it possible to perform fuel assembly strength and rigidity calculations, including the cases of abnormal operation. The mechanical characteristics of the skeleton and fuel rod bundle dummy models make it possible to check for the adequacy of the fuel assembly design model. The mechanical characteristics obtained during fuel rods bundle push through experiments make it possible to substantiate the fuel assembly serviceability under the conditions of fuel rods bundle and skeleton interaction

  14. Algorithms for diagnostics of the measuring channels and technological equipment at NPP with WWER-1000

    International Nuclear Information System (INIS)

    Vysotskij, V.G.

    1997-01-01

    An algorithm for diagnostics of the state of measuring channels of an information computer system with usage of analysis of statistical channel characteristics is presented. An algorithm for testing the generalized state of the NPP technological equipment is proposed

  15. SPND detectors response at the control rod drop in WWER-1000. Measurement and modelling results

    International Nuclear Information System (INIS)

    Mitin, V.; Milto, N.; Shishkov, L.; Tsyganov, S.; Kuzmichev, M.

    2006-01-01

    The paper analyzes and discusses possibility of neutron flux inspection in the WWER core during fast dynamic processes applying existing in-core monitoring system. The structure and functions of the system, basic principal of detector functioning and its temporal parameters are described briefly. To assess the ability of such dynamic monitoring the event with control rod drop happened during operation of Kozloduy NPP Unit 5 is observed - at the level of power close to nominal one of the rod from control group shifted to the lowest position at-2 seconds. In-core detectors readings at the process were registered and processed with mathematical methods that allow to single out only the prompt part of the signal. Results of the processing are presented. Furthermore, the process observing have been modeled with 3D dynamic code NOSTRA. Results of modeling are presenting in a paper, and comparing with experimental ones. A good agreement achieved. The analysis of measurements and its imitation give a hope that with an aggregate signal of detectors the measurement of control rod worth could be provided, and it allows to avoid of influence of spatial effects that are significant at standard technique with ex-core ion chambers (Authors)

  16. Development of expert system for fuel monitoring and analysis in WWER-1000 units

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Kanukova, V.; Zborovskii, V.; Aliev, T.; Sokolov, N.; Shishkin, A.

    2011-01-01

    At present, an expert system (software package) for fuel monitoring in WWER units is under development in Russia. It comprises several modules which cover analysis of coolant activity, detection of failures and estimation of failure parameters, predictions of activity level and some aspects of PCI analysis. This paper outlines the current version of the fuel monitoring system, its basic features and user interface. Advances in development of computer modules for PCI analysis are reported. At present two levels of PCI analysis are used. The first is estimation of probability for pellets to get in contact with cladding in fuel rods. Estimations are made with taking into account specifications and tolerances for fuel fabrication as well as fuel operation conditions. The second level of PCI analysis implies a simplified approach for on-line calculations of stresses in cladding depending on power ramping rates. The model for PCI calculations and its application within the computer system is demonstrated. (authors)

  17. Initial data on protective containment for safety analysis of NPP with WWER-1000 Preparation

    Directory of Open Access Journals (Sweden)

    Gubeladze Oleg

    2017-01-01

    Full Text Available In the presented article the nuclear safety issues with VVER-1000 reactor are considered. The study element is the protective containment (PC, the most important function of which is localization and retention of radioactive substances within the accident localization zone. The example of possible unregulated destructive forcing (UDF on the PC for the construction and installation work period is given.

  18. Experiments for neutron fluence assessment on WWER-440 and WWER-1000 pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ilieva, K; Apostolov, T; Penev, I; Trifonov, A; Taskaev, E; Belousov, S; Antonov, S; Petrova, T; Stoeva, L [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Boyadzhiev, Z; Nelov, N; Tsocheva, V; Andreeva, I; Lilkov, B; Velichkov, V; Monev, M [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    The activity of shavings sampled out from the expected maximum embrittlement location (weld 4) on the inner pressure vessel wall of the Kozloduy-1 Unit after the 14-th cycle has been measured. The experiment was carried out along the INEI channel using Fe and Cu string and foil detectors. The axial neutron flux distribution at the Unit 3 after the cycle 11 has been measured and compared to the calculated values. The calculations of the expected activities have been carried out taking into account the local power distribution. A comparison between measured and calculated values using ACTIVAT code is made. It shows a discrepancy of about 20%. It is recommended to carry out ex-vessel neutron fluence measurements using a rack device with activation detectors in order to verify the calculation results. 8 refs., 3 figs., 2 tabs.

  19. Alternative technology of containment construction for WWER 1000 nuclear power plant

    International Nuclear Information System (INIS)

    Chalus, Z.

    1982-01-01

    A number of alternatives was assessed for the assembly of the steel elements of the cylindrical part of containment. Alternative 1 is based on the common technology of manufacture, transport and assembly of reinforced concrete blocks of ca. 3x12 m in size, used for building leak-proof walls of WWER 440 nuclear power plants. Alternative 2 is based on reinforced concrete blocks using 12x12 m blocks assembled from individual elements on the building. site. Alternative 3 is a specific variant of the previous alternative. Alternative 4 envisages the assembly of a prefabricated support structure made of steel. Alternative 5 is based on the gradual assembly of partial elements mounted onto a support structure. Alternative 6 only differs from 5 in the method of assembly and manufacture of the support structure. All alternatives are shown in diagrams. (J.B.)

  20. Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

    International Nuclear Information System (INIS)

    Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

    1996-01-01

    The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

  1. Floor response spectra of WWER-1000, NPP Kozloduy generated from local seismic excitation

    International Nuclear Information System (INIS)

    Bojadziev, Z.; Kostov, M.

    1996-01-01

    The seismic review level characteristics for the Kozloduy NPP site were set to 0.2 g and a respective free field acceleration response spectra were derived after a profound site conformation project. Accordingly a separate investigation is recommended for local seismic excitation. The goals of the analyses are: to define the seismic motion characteristics from local seismic sources; to perform structural analyses and in-structure spectra generation for local seismic excitation; and to compare the forces (spectra) from local events with those generated as seismic design review basis

  2. Main Aspects and Results of Level 2 PSA for KNPP WWER-1000/B320

    International Nuclear Information System (INIS)

    Mancheva, Kaliopa

    2014-01-01

    The PSA Level 2 for Kozloduy NPP (KNPP) is an update of an older study with wider scope of analysis. The older study represented the status of the units up to 2001. The current PSA Level 2 is based on the PSA Level 1 and represents the status of the units up to 2007 year concerning the systems and procedures included in PSA level 1 and status up to 2011 for the systems and procedures (e.g. SAMG) related to containment and severe accident aspects. The study is performed after the PSA level 1 has been finished and approved by the customer. Compare to the older analysis all modes of operation for analyzed in PSA level 1 event groups as well Spent Fuel Pool accidents are investigated. The analysis consists of both deterministic and probabilistic analysis. As part of deterministic analysis a contemporary containment strength analysis and accident progression deterministic analysis using last version of MELCOR are performed. The probabilistic analysis contains of two part: Interface PSA and CET are calculated using Riskspectrum program code. Two types of models for CET have been developed: one for conditional probabilities calculations and a set of simplified CET's for each PDS group-for integral model. The purpose of the first model is to be able to perform quick calculations and for sensitivity analyses as well. The simplified CET's are used for integral calculation of the model. Source Term analysis is mainly based on the MELCOR analyses results. All characteristics of the releases have been defined, i.e. location, mass, energy of radionuclide groups and activity of the released isotopes (most important are reported only). The main goals of the study are to analyze the status of the containment, systems designed to prevent containment failure and operator action required under the severe accident and to give quantitative assessment of the risk parameter LERF (Large Early Release Frequency). This report will present main aspects, results, finding and conclusions of the analysis. Based on the results, the effectiveness of the current means and SAMG are assessed qualitatively. (author)

  3. Reactor noise analysis applications and systems in WWER-440 and WWER-1000 type PWRs

    International Nuclear Information System (INIS)

    Por, G.

    1998-01-01

    This paper presents an introduction on different types of well selected noise diagnostic methods with their occurrence in WWER reactors with an analysis of their impact on operational safety and aging which affects the installations safety as well. The main objective is to attract the attention of NPP management staff dealing with safety, safety culture, maintenance, operation and quality assurance proving that such methods can give benefit not only to economy but impact safety of nuclear installations

  4. Application of simple approximate system analysis methods for reliability and availability improvement of reactor WWER-1000

    International Nuclear Information System (INIS)

    Manchev, B.; Marinova, B.; Nenkova, B.

    2001-01-01

    The method described on this report provides a set of simple, easily understood 'approximate' models applicable to a large class of system architectures. Constructing a Markov model of each redundant subsystem and its replacement after that by a pseudo-component develops the approximation models. Of equal importance, the models can be easily understood even of non-experts, including managers, high-level decision-makers and unsophisticated consumers. A necessary requirement for their application is the systems to be repairable and the mean time to repair to be much smaller than the mean time to failure. This ia a case most often met in the real practice. Results of the 'approximate' model application on a technological system of Kozloduy NPP are also presented. The results obtained can be compared quite favorably with the results obtained by using SAPHIRE software

  5. Results of research and development for nuclear power plants with WWER-1000 type light water reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The conference met in three sessions: 1. Project designing and construction of nuclear power plants; 2. Materials, technologies and applied mechanics; 3. Physics, thermal physics and control. The proceedings contains 82 papers of which only two have not been inputted in INIS. The final resolutions of session 1 related to the reduction of capital costs for newly built units, processing of project documentation, the introduction of step motors manufactured in Czechoslovakia, in-service diagnostics of nuclear power plants, etc. The final recommendations of session 2 dealt with the centralization of the management of research into the reliability, safety and residual life of nuclear installations, with radiation stability of weld metals, repairs of nuclear power plants by patch welding, with welding in nuclear power plants and stress calculations using mathematical methods. Session 3 centred on questions of the safety, reliability and economy of nuclear power plant operation. It was recommended to make a comparison of the results of theoretical calculations with experiments, to concentrate on the automation of measurement, to extend international division of labour and cooperation of CMEA countries, to extend publishing activities in the field of thermal physics, etc. General recommendations were related to the conception of the construction of nuclear power plants in Czechoslovakia, the implementation of original scientific, research and development work, to the question of personnel for nuclear research, the experimental base of the Czechoslovak nuclear programme and to planning and management of technical development. (E.S.)

  6. Recent advances and achievements in WWER-1000 fuel design performance and operation

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Molchanov, V.

    2009-01-01

    In this paper the main results of TVS-2 (TVS-2M) basic design operation like TVS-2 - 1216 pcs., including 687 in operation, TVS-2M - 66 pcs. in operation (Balakovo-1); reliability -1,6·10-6 in 2008 and the Balakovo NPP capacity factor - 90% are presented. Average efficient operating time of the Balakovo power units, calculated capacity factor for the Balakovo NPP at transition to fuel cycle 3x1,5, contemporary operational requirements for fuel and their realization as well as SPND axial position variation in the core with TVS - 2M and CPS AR usage are shown

  7. Building concept of cooling towers for WWER-1000 nuclear power plants

    International Nuclear Information System (INIS)

    Bucha, V.; David, M.

    1984-01-01

    A project is described of cooling towers with natural draught for the Temelin nuclear power plant. The concept proceeds from the classical design of the so-called Itterson type, i.e., the outer cladding of the draught stack is made of a monolithic reinforced concrete unit in the shape of a hyperboloid of revolution supported by a system of oblique supports mounted along the edge of the cooled water tank. The procedure is explained of the thermal calculation for the given operating conditions. The basic alternatives are considered of the choice of material and design of the cooling system. Questions are discussed relating to the design of the eliminator, the windwart wall and the shape of the shell of the draught stack and its loading by wind and seismic effects. (E.S.)

  8. Steam feeding redundancy for turbine-drives of feed pumps at WWER-1000 NPP

    International Nuclear Information System (INIS)

    Nesterov, Yu.V.; Shmukler, B.I.

    1987-01-01

    The system of steam supply for feed pump driving turbines (T) at the South Ukrainian Unit 1 according to the centralized redundancy principle is described. T is feeded through the collector of water auxiliary sytem (CWAS) to which steam from the third steam extraction line of turbine is supplied under thenormal regime. Under the reduction of turbine load, live steam from the steam generator is supplied to CWAS through the pressure regulator, possesing 10 s speed of responce. In this case the level reduction in the steam generator makes up 170 mm

  9. Improvement of operation efficiency for WWER-440 and WWER-1000 for TRIGON fuel assembly design features

    Energy Technology Data Exchange (ETDEWEB)

    Silberstein, A [European WWER Fuels GmbH, Lyon (France)

    1994-12-31

    TRIGON 440 and TRIGON 1000 fuel assemblies and their assembly matching counterparts are described. Their role in increasing the efficiency of WWER reactors is stressed. Special attention is paid to their design features as well as calibrated means of predicting behaviour under irradiation from light water reactor core operation. They reduce the fuel cycle cost as a result of the reduced need for natural uranium which have to be enriched and of the smaller number of fuel assemblies which have to be fabricated, stored or reprocessed. The improved control assemblies bring comfort to the plant operator due to intrinsic progress in safety with respect to accidental situation, trouble-free behaviour and long time utilization in the reactor. 14 figs.

  10. Experimental study of ballooning and failure of WWER-1000 fuel cans during maximum design basis accident

    International Nuclear Information System (INIS)

    Karetnikov, G.V.; Bogdanov, A.S.; Semishkin, V.P.; Bezrukov, Yu.A.; Trushin, A.M.; Frizen, E.A.

    2001-01-01

    The processes of ballooning and fracturing in tubular specimens of Eh635 and Eh110 alloy fuel cans are investigated with the use of cinematography. The investigations are carried out under steady-state conditions in the temperature range from 680 to 900 deg C and at pressure drops on the can from 2 to 12 MPa. Time dependences of circumferential strains are plotted for various temperatures of fuel cans at pressure of 2 MPa. It is shown that strain changes are of linear character at an initial portion of the curve and then an accelerated strain development takes place with transition to fracture. Using methods of nonlinear evaluation for time to fracture the approximation dependences are obtained for fuel cans. Experimental data are intended to form the equations of state for fuel can materials and to verify the program TVEL-3 [ru

  11. Modeling of the WWER-1000 fuel-rod behavior in steady-state condition with FRAPCONE-3 computer code

    International Nuclear Information System (INIS)

    Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan

    2008-01-01

    It is presented within the paper the results of the modeling and the assessment of the integral code predictions of the WWER fuel-rod behavior in steady-state condition. The assessments in this paper have used the MASSIH and ANS 5.4 subroutine in the code. The modeling and calculations have been performed with FRAPCONE-3 computer code in Argonne National Laboratory, USA

  12. Reactor physics of light water power reactors of the WWER-1000 type in the SKODA Trust. Part 2

    International Nuclear Information System (INIS)

    Zbytovsky, A.; Vyskocil, V.; Vacek, J.; Krysl, V.

    1979-01-01

    The system is described of multigroup transport programs and the respective data libraries used at the SKODA WORKS for calculating few-group diffusion constants. The importance is stressed of precisioning nuclear data. The programs for multigroup calculations are mostly written in FORTRAN-1900. The structure of microcalculations is described and the possible cost savings are indicated. (J.P.)

  13. Development of requirements for seismic upgrading of equipment of existing WWER-440 and WWER-1000 type NPPs

    International Nuclear Information System (INIS)

    Kaznovsky, S.; Ostretsov, I.

    1993-01-01

    The change in seismology data and safety demands a necessity arose for seismic upgrading of the existing operating NPPs of WWER type which have been originally designed and built without or with simplifies calculations of seismic influences. The paper describes the traditional methods and approaches and calculation-experimental method for examining and ensuring of equipment seismic resistance at the NPPs directly. Method of ground explosions is included as well

  14. Analyses with ASTEC related to release of FPs and aerosol transport in case of SBLOCA For WWER 1000

    International Nuclear Information System (INIS)

    Atanasova, B.; Stefanova, A.; Groudev, P.

    2008-01-01

    The objective of this paper is to present the results obtained from performing the calculations with ASTEC computer code for the Source Term evaluation for specific severe accident transient. The calculations have been performed with the new version of ASTEC. The ASTEC 1.3 R2 code version is released by the French IRSN (Institut de Radioprotection at de surete nucleaire) by the end of 2007. The sequences include the release of fission products into the reactor containment and environment and transport of fission products. The analyses proposed here are performed to simulate radioactive products release through the cold leg of SG under accidental conditions. This investigation has been performed in the framework of the SARNET project (under the EURATOM 6th framework program) by the FoBAUs group (Forum of Bulgarian ASTEC users). (authors)

  15. Principles of building and assembly technology of containment from steel structural blocks for WWER 1000 nuclear power plant

    International Nuclear Information System (INIS)

    Eichstedt, J.; Friedrich, F.

    1983-01-01

    This technology is being developed in cooperation between the USSR and the GDR. The cylindrical part of the containment consists of prefabricated double-sided steel blocks with inner reinforcement. The steel plates in a thickness of 20 mm provide casing and secure tightness. Blocks with one steel wall are used for the construction of the cupola. The outer slabs are assembled subsequently. The methods of assembly, concreting and quality assurance are described. (Ha)

  16. The feasibility of express in situ measurement of the isotopic composition of uranium in fresh WWER-1000 fuel

    International Nuclear Information System (INIS)

    Christoskov, I.; Tsankov, L.; Ivanov, N.

    2011-01-01

    A study of the feasibility and accuracy of measurement of the enrichment of fresh VVER-1000 fuel assemblies using a miniature CdZnTe probe is performed.The possibilities of improvement of the analytical procedure are briefly discussed. (authors)

  17. Results of TVSA fuel assemblies development and 10-years operation in WWER-1000 reactors cores. Development trends

    International Nuclear Information System (INIS)

    Molchanov, V.; Sharikov, A.; Samoilov, O.; Kaidalov, V.; Falkov, A.; Romanov, A.; Shishkin, A.

    2009-01-01

    The basic TVSA design is now operating successfully in 17 VVER-1000 power units of Russia, Ukraine and Bulgaria. TVSA design is characterized by reliability and validity of designs and resource characteristics. TVSA possesses light hydraulics and peak ratio for implementation of effective fuel cycles. Unique parameters on time of maintenance and burnup are achieved. The basic design - TVSA-5M - has been introduced into commercial operation. It is necessary to expand heading TVSA with debris filter and temperature control tubes. Two directions of evolution of a design - TVSA-ALPHA and TVSAPLUS are realized. Evolution is directed on increasing uranium content and realization of 5x1 and 3x1.5 fuel cycles with operation at raised power. The TVSA design offers a great optimization potential based on a series of proven (reference) technical solutions. TVSA is competitive and cost effective and ensures the market advantages of Russian nuclear fuel

  18. Control of power distribution and the safety of a WWER-1000 reactor during operation in a load-following regime

    International Nuclear Information System (INIS)

    Filipchuk, E.V.; Dunaev, V.G.; Luk'yanets, I.A.; Potapenko, P.T.; Timokhin, E.S.; Voznesenskij, V.A.; Mitin, V.I.

    1984-01-01

    An important task in ensuring the operational safety of a nuclear power plant with a vessel-type water-cooled, water-moderated reactor is the development of algorithms for controlling power distribution in base-load and load-following regimes. The control strategy must ensure that the form of power distribution is maintained within the permissible limits in the case of situations presenting a nuclear hazard and of xenon fluctuations when there are variations in power output. The task of controlling power distribution can be broken down into an axial problem and a radial-azimuthal problem. For maintenance of the form of axial power distribution an algorithm of optimum fast-action control is proposed; for maintenance of the radial-azimuthal form, an algorithm of optimum (i.e. best) approximation to the given distribution of control. The two control problems are solved by using a single computerized procedure based on mathematical programming. The algorithms are designed for use in computerized control. The paper presents the results of algorithm simulation. The authors propose that in the first stage, activation of the control systems should involve the operator, who can view the sequence of actions on a display screen. (author)

  19. Probability analysis of WWER-1000 fuel elements behavior under steady-state, transient and accident conditions of reactor operation

    International Nuclear Information System (INIS)

    Tutnov, A.; Alexeev, E.

    2001-01-01

    'PULSAR-2' and 'PULSAR+' codes make it possible to simulate thermo-mechanical and thermo-physical parameters of WWER fuel elements. The probabilistic approach is used instead of traditional deterministic one to carry out a sensitive study of fuel element behavior under steady-state operation mode. Fuel elements initial parameters are given as a density of the probability distributions. Calculations are provided for all possible combinations of initial data as fuel-cladding gap, fuel density and gas pressure. Dividing values of these parameters to intervals final variants for calculations are obtained . Intervals of permissible fuel-cladding gap size have been divided to 10 equal parts, fuel density and gas pressure - to 5 parts. Probability of each variant realization is determined by multiplying the probabilities of separate parameters, because the tolerances of these parameters are distributed independently. Simulation results are turn out in the probabilistic bar charts. The charts present probability distribution of the changes in fuel outer diameter, hoop stress kinetics and fuel temperature versus irradiation time. A normative safety factor is introduced for control of any criterion realization and for determination of a reserve to the criteria failure. A probabilistic analysis of fuel element behavior under Reactivity Initiating Accident (RIA) is also performed and probability fuel element depressurization under hypothetical RIA is presented

  20. Thermal-hydraulic model of the primary coolant circuits for the full-scale training facility with WWER-1000

    International Nuclear Information System (INIS)

    Kroshilin, A.E.; Zhukavin, A.P.; Pryakhin, V.N.

    1992-01-01

    The mathematical model realized in the full-scale educational facility for NPP operator training is described. The RETACT computational complex providing real time process simulation for all regimes including the maximum credible accident is used for calculation of thermohydraulic parameters of the primary coolant circuits and steam generator under stationary and transient conditions. The two-velocity two-temperature model of one-dimensional steam-water flow containing uncondensed gases is realized in the program

  1. Main trends and content of works on fabrication of fuel rods with MOX fuel for the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Tsykanov, V.A.; Golovanov, V.N.; Mayorshin, A.A.; Yurchenko, A.D.; Ilyenko, S.A.; Syuzev, V.N.

    2000-01-01

    The main trends of production of pellet MOX-fuel for the WWER reactors using the trial-experimental equipment at SSC RF RIAR are set forth. The main realized parameters of fabrication of MOX-fuel pellets are presented. The content of the reactor tests program is considered with allowance for their licensing requirements for the WWER reactors. (author)

  2. First qualitative analysis of fuel irradiation results carried out in the MR reactor on WWER-1000 fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chantoin, P [International Atomic Energy Agency, Vienna (Austria); Dubrovin, K; Platonov, P [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Onufriev, V [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    Four experiments carried out in the MR reactor are evaluated. They are aimed to assess the influence of burnup and the size of the pellet central hole on the fuel temperature and thus on the fuel swelling and fission gas release. The experiments have been performed at different linear rate and burnup of the fuel rods which are above the actual licensed values in WWER power stations. In this paper the results on WWER fuel rod behaviour are examined. The main fabrication and irradiation characteristics for each experiment are given. The main results from destructive and non-destructive examinations are summarized. They include: burnup determination by gamma spectroscopy, caesium shifting along fuel column and accumulation at the end of the fuel stack, fission gas release. fuel rod diameter and length change and macro-graphs showing the central hole size and the morphology after irradiation. From observation of fuel structure, Cs spectrometry and fission gas release, a large degradation of fuel thermal conductivity can be identified at high burnup. If the fuel burnup is the right parameter to be considered, burnup limits identified are: 0 70-75 MWd/kg for rods with large central hole; (2) 58-64 MWd/kg for rods with small central hole. As a general conclusion it is stressed the importance of the study due to irradiation beyond the usual linear rates at high burnup. Up to now the fuel life limiting factor was cladding corrosion when using Zircaloy-4. As the cladding corrosion situation improves, the next life limiting factor to be met could be the fuel itself. The decreasing fuel thermal conductivity is probably of prime importance and should be further studied and modelled. 5 tabs., 5 figs., 3 refs.

  3. Atomic Energy Research working group a on improvement extension and validation of parametrized FEW-group libraries for WWER-440 and WWER-1000

    International Nuclear Information System (INIS)

    Svarny, J.

    1998-01-01

    The AER Working Groups A and B held its sixth meeting at SKODA JS, Plzen in April 28 and 29, 1998. There were altogether 13 participants from 6 member organizations. The list of participants and the list of papers are attached. (Author)

  4. Behavior of antimony isotopes in the primary coolant of WWER-1000-type nuclear reactors in NPP Kozloduy during operation and shutdown

    International Nuclear Information System (INIS)

    Dobrevski, Ivan D.; Zaharieva, Neli N.; Minkova, Katia F.; Gerchev, Nikolay B.

    2009-01-01

    This paper focuses on the behavior of the antimony isotopes 122 Sb and 124 Sb in the coolant of the WWER reactors in the nuclear power plant Kozloduy (Bulgaria) during operation and shutdown. It is concluded that the chemical properties of their actual precursor, the isotope 121 Sb, determine the behavior of 122 Sb and 124 Sb during operation, load fluctuations, and shutdown as well as during the reactor coolant purification process. It is supposed that differences between the reactor bulk and the core fuel cladding surface chemistry as well as the presence of sub-cooled nucleate boiling at the fuel cladding may create conditions under which a local oxidizing environment may come into existence. (orig.)

  5. IMPACT OF THE CHEMICAL FORM OF IN-CONTAINMENT SOURCE ON FISSION PRODUCT RELEASE FROM WWER-1000/V-320 TYPE NPP CONTAINMENT DURING LOCA

    Directory of Open Access Journals (Sweden)

    Adam Kecek

    2016-12-01

    Full Text Available Nuclear power plant accidents may be followed by a release of fission products into the environment. This release is dependent on several phenomena, such as chemistry, pressure, type of the accident etc. The aim of this paper is to assess the impact of the chemical form of iodine on the fission product release into the environment.

  6. Assessment of realization possibility of borate-lithium water chemical regime of the PGV-100 secondary circuit for NPP with WWER-1000 type reactor

    International Nuclear Information System (INIS)

    Smykov, V.B.; Ermolaev, N.P.; Borzov, I.A.

    1994-01-01

    The theoretical and experimental assessment of chemical equilibrium reactions resulting in the weak-alkaline boric buffering water chemistry of secondary side PGV-1000 has been performed out in the article. 18 refs

  7. To the question on vibration strength of the main circulating pumps of reactor WWER-1000 at influence of the non-stationary hydrodynamic stream

    International Nuclear Information System (INIS)

    Pukhlij, V.A.

    2015-01-01

    The problem of vibration strength of working shovels of centrifugal pumps is based on definition of dynamic behaviour of shovels on the basis of the theory of flat covers, thus integration of the initial equations of movement is carried out by the modified method consecutive approach. The problem of forced and natural fluctuations of a shovel is solved, which makes it possible to identify the resonant zones of operation of centrifugal pumps of nuclear power, in particular, the main circulation pump GTSN 195M [ru

  8. The Testing of Fuel Rod Models with Zr1Nb Alloy Cladding in Water Vapor at Temperature of Hypothetical Accident Situation in WWER-1000 Type Reactors

    International Nuclear Information System (INIS)

    Krasnorutsky, V.S.; Petel'guzov, I.A.; Gritsina, V.M.; Rodak, A.G.; Belash, N.N.; Yakovlev, V.K.

    2006-01-01

    In the article happen to results of testing the fuel rod models, their welded joints, changing the mechanical characteristics of shells of models from experimental parties of pipes from Zr1Nb alloy (Zr+1 mass%Nb) at heating of models, pervaded helium before pressures, using in earned one's living fuel rods (2,2 MPa), before the temperature 770 degree C and above occurs an overblown fuels, but at temperature 820...830 degree C shells can be broken at the expense of pressure of warming gas. Swept away reduction plasticity and embrittlement shells after the heating under temperature of 900...1200 degree C and cooling before room temperature pipes-shells from Zr1Nb alloy and from the staff alloy E110

  9. Safety issues and their ranking for WWER-1000 model 320 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-03-01

    The objective of this report is to present a consolidated list of safety deficiencies, called safety issues, ranked according to their safety significance and the corrective measures to improve overall safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. To the extent that information was made available to the IAEA, the country/plant specific status with respect to each safety issue is described. Section 2 provides an overview of the impact of the relevant issues on the main safety functions in different operational conditions and other aspects important to overall plant safety. A summary of the safety issues and their respective ranking is given in Tables 1 and 2 at the end of Section 2. Section 3 deals with individual safety issues identified in the design which are presented according to the structure below. Section 4 presents the safety issues related to operational safety according to a similar structure but without the ranking. 73 refs, 3 tabs

  10. Safety issues and their ranking for WWER-1000 model 320 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1997-04-01

    The objective of this report is to present a consolidated list of safety deficiencies, called safety issues, ranked according to their safety significance and the corrective measures to improve overall safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. To the extent that information was made available to the IAEA, the country/plant specific status with respect to each safety issue is described. Section 2 provides an overview of the impact of the relevant issues on the main safety functions in different operational conditions and other aspects important to overall plant safety. A summary of the safety issues and their respective ranking is given in Tables 1 and 2 at the end of Section 2. Section 3 deals with individual safety issues identified in the design which are presented according to the structure below. Section 4 presents the safety issues related to operational safety according to a similar structure but without the ranking

  11. Mathematical modelling and quality indices optimization of automatic control systems of reactor facility

    International Nuclear Information System (INIS)

    Severin, V.P.

    2007-01-01

    The mathematical modeling of automatic control systems of reactor facility WWER-1000 with various regulator types is considered. The linear and nonlinear models of neutron power control systems of nuclear reactor WWER-1000 with various group numbers of delayed neutrons are designed. The results of optimization of direct quality indexes of neutron power control systems of nuclear reactor WWER-1000 are designed. The identification and optimization of level control systems with various regulator types of steam generator are executed

  12. Some aspects of RPV integrity of Ukrainian NPP's

    International Nuclear Information System (INIS)

    Zaritsky, N.; Kovyrshin, V.; Zhukov, P.

    1998-01-01

    The operating organisations in Ukraine implement the main IAEA recommendations concerned with NPP operational safety. Sufficient substantiation of the measures for improvement of WWER-1000 NPPs safety and reliability is provided. General information (chemical compositions and mechanical properties of reactor pressure vessel materials) is already collected on all Ukrainian WWER-1000 NPPs and transferred to International database on Pressure Vessel materials

  13. The use of energy analysis and indexes of energy efficiency in nuclear power

    International Nuclear Information System (INIS)

    D'yakonov, E.I.; Ignatenko, E.I.

    1991-01-01

    The results of calculating the indexes of energy efficiency for NPPs with the WWER-1000 and RBMK-1000 reactors, heat and power NPPs with the WWER-1000 and dictrict heating NPPs with the AST-500 reactor in three fuel cycles, namely, the open one and with uranium and plutonium recycles, are considered. Complex account for the quantity and quality of produced and consumed energy provides for objective evaluation of the indexes of energy efficiency during comparative analysis of nuclear power plants with different types of reactors. It is shown that complex use of the energy produced at a NPP provides for increase of indexes of energy efficiency. The highest indexes are obtained for heat and power NPP with the WWER-1000 reactor in the open fuel cycle, with uranium and plutonium recycle and for NPP with the WWER-1000 reactor with plutonium recycle

  14. Reserves of labour content reduction in NPP construction

    International Nuclear Information System (INIS)

    Bekerman, R.E.

    1986-01-01

    Specific labour contents when constructing NPP with RBMK-1000 and WWER-1000 type reactors are presented. Factors affecting labour content of NPP construction are shown. Measures aimed at labour content decrease are suggested

  15. Experience on reliability of insertion of control rod with PWRs in Japan

    International Nuclear Information System (INIS)

    Nakata, H.

    1995-01-01

    This paper is prepared to introduce, in response to the IAEA Letter dated 1995-02-08, national experience concerning to the reliability of insertion of control rod with PWRs which is relatively similar to WWER 1000. 3 figs

  16. Development, application and also modern condition of the calculated program Imitator of a reactor

    International Nuclear Information System (INIS)

    Aver'yanova, S.P.; Kovel', A.I.; Mamichev, V.V.; Filimonov, P.E.

    2008-01-01

    Features of the calculated program Imitator of a reactor (IR) for WWER-1000 operation simulation are discussed. It is noted that IR application at NPP provides for the project program (BIPR-7) on-line working. This offers a new means, on the one hand, for the efficient prediction and information support of operator, on the other hand, for the verification and development of calculated scheme and neutron-physical model of the WWER-1000 projection program [ru

  17. On the application of nonlinear parameters of concrete degradation for NPP containment shell calculation

    International Nuclear Information System (INIS)

    Guzeev, E.A.; Shevchenko, L.A.; Krichevskij, A.Z.

    1986-01-01

    Main work directions of the Energomontazhproekt Kharkov branch on development of a new technology, mechanisms and devices providing the reduction of labour content and time of equipment mounting at NPPs with WWER-1000 reactors, decrease of a share of manual labour, reduction of specific quantity of metal for mounting mechanisms and devices are considered. Recommendations on further improvement of effectiveness and quality of thermal-mounting works performed at NPPs with WWER-1000 reactors are given

  18. Nuclear and thermal power plant power ramping capability

    International Nuclear Information System (INIS)

    Golovach, E.A.

    1983-01-01

    The possibilities of step power increase by NPP and TPP units under emergency conditions of power grids operation are considered. The data analysis has shown that power units ramping capability with WWER-440, WWER-1000 and RBMK-1000 reactors is higher than that of 300 MW power units on fossil fuel, at the initial time interval (0-30 s). These NPP power units satisfy as to ramping capability the energy system requirements. Higher NPP power units ramping capability is explained by the fact that relative pressure before turbine valves is decreased less than in straight-through boilers while the steam volumes time constant of steam separator-superheaters is less than that of intermediate superheatings. Higher power unit ramping capability with WWER-440 and RBMK-1000 reactors as compared with the WWER-1000 reactor is pointed out as well as the increase of WWER-1000 power unit capability using high-speed turbines

  19. Prospect of increases the efficiency of nuclear energy production in advanced light water reactors

    International Nuclear Information System (INIS)

    Proskuryakov, Konstantin; Belikov, Svyatoslav; Novikov, Konstantin

    2011-01-01

    Values of quality factor and pass-band for acoustic model of reactor core of reactor unit with WWER - 1000 are defined at single-phase and biphasic conditions of the coolant. The settlement-theoretical substantiation of conditions of growth of vibrations of internals and fuel assembly within a pass-band limits. Results of calculation and measurements have suitable coincidence. The technique of forecasting is developed and results of forecasting of vibration acoustical resonances of fuel assemblies with coolant in stationary and transitive operating modes of reactor unit with WWER - 1000 are worked out. (author)

  20. Status report: Nuclear fuel operating experience in implementing the program for power generation increase at VVER NPPs of JSC concern Rosenergoatom

    International Nuclear Information System (INIS)

    Ryabinin, Y.

    2015-01-01

    The power uprate program of operating WWER-1000 plants was performed by Rosenergoatom using FA-2M and FAA-PLUS for 18-month fuel cycles. Their operation was justified at 104% of the rated power, and extension to 18-month fuel cycles was carried out at WWER-1000 units (except for Kalinin NPP-1). The analysis of actual performance data confirmed the efficiency of the actions implemented, and issues addressed related to the introduction of new fuel type, extended fuel cycles and spent nuclear fuel storage and removal

  1. The FARC fuel archive of WWER

    International Nuclear Information System (INIS)

    Zizin, M.N.; Parfenova, N.A.; Proselkov, V.N.; Shishkov, L.K.

    1998-01-01

    The principles of organisation are explained and the structure of the FARC fuel archive for WWER reactors is circumscribed. The objective of archive is accumulation of fuel data, data storage and obtaining of fuel using characteristics. The working version of fuel archive on 01 July 1998 is realised, in which the data tables for fuel assemblies for 169 WWER-440 cycles and 35 WWER-1000 cycles are stored. There are two different versions of fuel archive - for WWER-440 (FARC) and for WWER-1000 (FARC1000). A structure of some tables and the texts of programs for them differ. (Authors)

  2. Safety problems of nuclear power plants with reactors of new generation; Voprosy bezopasnosti v proehktakh AEhS novogo pokoleniya s WWER

    Energy Technology Data Exchange (ETDEWEB)

    Fedorov, V; Rogov, M; Biryukov, G [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1996-12-31

    Modernization schemes for safety enhancement of WWER-1000 reactors are proposed. In the case of WWER-1000/V-392 it is based on introduction of additional safety systems and overall design improvement. For WWER-1100 (1000-1100 MW) the safety is enhanced by passive systems built from two-stage heat exchangers. For WWER-500/600 the use of passive safety system is extended to emergency cooling of the active zone and removal of the residual heat emissions from the reactor. The technical characteristics of the three reactors are compared. 3 figs., 1 tab.

  3. National report on nuclear power plant control and instrumentation in Czechoslovakia

    International Nuclear Information System (INIS)

    Stirsky, P.; Karpeta, C.

    1986-01-01

    Research, development and design efforts in the field of nuclear power plant I and C systems in Czechoslovakia have been recently aimed at solving the following problems: setting the parameters of the WWER 440 units control and protection systems and testing them in the power phase of commissioning; design and simulation of the WWER 440 units control system performance under the conditions of steam bleeding for a centralized heat supply system; development of a simulation model of the unit WWER 1000 dynamics for the purpose of I and C systems investigation and design; design of innovated I and C systems for WWER 440 and WWER 1000 units

  4. The Role of Computer-Based Educational Laboratories in Nuclear Engineering University Programmes

    International Nuclear Information System (INIS)

    Korolev, S.A.; Kosilov, A.N.; Chernov, E.V.; Vygovskiy, S.B.

    2014-01-01

    The specialized Educational and research laboratory 'Reactor physics, control and safe operation of WWER type NPP’ is based on the computer simulator of WWER -1000 and offers the real-time monitoring of data available to the WWER -1000 NPP control room operators, and provides a possibility to investigate reactor behavior in normal and abnormal situations. The laboratory supports interactive technologies and team-based activities that enable students to build their knowledge through required gateway courses and explore problems relevant to real life situations

  5. New welding and finishing procedures

    International Nuclear Information System (INIS)

    1992-01-01

    The proceedings contain texts of the 31 contributions presented at the conference. One of them, dealing with the reheat crackability in the heat affected zone of steels for WWER-440 and WWER-1000 reactor vessels, has been inputted in INIS. The behavior of the steels was determined by the simulation technique. (M.D.). 4 figs., 5 tabs., 5 refs

  6. Thermal-hydraulic causes of increase of dynamic stresses and cracks in the covers of vessel reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.

    2006-01-01

    Paper presents data of measurements and of analysis of the WWER-1000 reactor NPP equipment process noises manifesting occurrence of the latent dynamic processes. The NPP is shown to have unused abilities to prolong the equipment service life and to reduce possibility of the equipment thermally and hydraulically caused unexpected malfunctions [ru

  7. Informational system to assist decision making at spent nuclear fuel transportation from VVER-440, VVER-1000 and RBMK-1000 nuclear power plants

    International Nuclear Information System (INIS)

    Kuryndin, A.V.; Kirkin, A.M.; Stroganov, A.A.

    2012-01-01

    The developed informational system provides an automated estimations of nuclear and radiation safety parameters during spent nuclear fuel transportation from WWER-440 and WWER-1000 and RBMK-1000 nuclear power plants to the nuclear fuel cycle facilities, and allows us to determine the optimum cask loading from the dose rates distribution outside of protection point of view [ru

  8. The technologies of zirconium production for nuclear fuel components in Ukraine

    International Nuclear Information System (INIS)

    Semenov, G.R.

    2000-01-01

    Perspectives of development zirconium alloys and WWER-1000 assemble components production in Ukraine are considered. Basic technological production processes of zirconium alloys in conditions of Ukrainian enterprises and modern requirements are analyzed. The critical processes on technical and economic criteria are defined. The main directions of activity and steps on technological processes improvement for production quality providing are offered. (author)

  9. Innovation of genetic algorithm code GenA for WWER fuel loading optimization

    International Nuclear Information System (INIS)

    Sustek, J.

    2005-01-01

    One of the stochastic search techniques - genetic algorithms - was recently used for optimization of arrangement of fuel assemblies (FA) in core of reactors WWER-440 and WWER-1000. Basic algorithm was modified by incorporation of SPEA scheme. Both were enhanced and some results are presented (Authors)

  10. TORT application in reactor pressure vessel neutron flux calculations

    International Nuclear Information System (INIS)

    Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

    1994-01-01

    The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

  11. Perspective decisions of WWER nuclear fuel: Implementation at Russian NPPs

    International Nuclear Information System (INIS)

    Molchanov, V.

    2003-01-01

    The scientific and technical policy pursued by JSC TVEL has managed to create a new generation of fuel assembly design on the basis of solutions tested at various units of Russian NPPs - Kola NPP, Kalinin NPP, Unit 1, Balakovo NPP Unit 1. The requirements set for the new generation nuclear fuel for WWER are: 1) High fuel burnup - up to 70 MWxdays/kgU; 2) Extended operation cycle - up to 6 years; 3) Increase of uranium charge to the core; 4) Increased lateral stability - bow not more than 7 mm; 5) High level of operating reliability - fuel rod leakage not worse than 10-5 1/year; 6) Demountable fuel assembly design. Post-irradiation examination results of fuel assemblies discharged from WWER-1000 reactors demonstrate that fuel rods have substantial reserve in general characteristics including that of dealing with planned burnup. In order to meet the requirements, trials are started for: implementation of rigid skeleton (WWER-1000); fuel column length extension (WWER-1000 and WWER-440); increase of UO 2 charge (WWER-1000 and WWER-440); enhancing of operational reliability and demountable design. It is concluded that the Russian nuclear fuel for WWER-type reactors is competitive and enables the implementation of state-of-the-art cost effective fuel cycles

  12. Automatized welding equipment for manufacturing steel cells for special buildings

    International Nuclear Information System (INIS)

    Weikert, F.; Winter, K.P.

    1986-01-01

    In GDR's nuclear power plant construction, reinforced concrete wall cells are used to construct pressure and full pressure containments for WWER-440 and WWER-1000 reactors, respectively. Welding processes for the prefabrication of steel cells as reinforcement have been automatized in order to increase both labor productivity and quality assurance. 11 figs

  13. Nuclear power project planning and implementation

    International Nuclear Information System (INIS)

    Roumiantsev, A.N.

    1976-01-01

    General description of NVRs. Development of NVR technology. Basic peculiarities of NVR: technological scheme, cooling circuits, core design, initial load of core, core discharging, control and safety, operating regime, economy. Basic technical characteristics of WWER-440 and WWER-1000. (orig.) [de

  14. Peculiarities of cyclic deformation and fracture of heat-resistant steel 10GN2MFA under conditions typical for the steam generator PGV-1000 collector material

    International Nuclear Information System (INIS)

    Giginyak, F.F.

    1998-01-01

    In the present paper, the results are discussed concerning investigations into the regularities of deformation and fracture of steel, 10GN2MFA under conditions close to those of actual operation when used in collectors of PGV-1000-type steam generators, which are in service with WWER-1000-type reactors of nuclear power plants (NPP). (author)

  15. The IAEA extrabudgetary programme on the safety of WWER and RBMK plants

    International Nuclear Information System (INIS)

    Havel, R.

    1995-01-01

    Data on WWER-440/213, WWER-440/230, WWER-1000 and RBMK reactors in operation are presented. Organizational chart for the IAEA extrabudgetary programme on the safety of WWER and RBMK plants, general programme objectives and main components are outlined

  16. Current state of WWER SNF storage in Russia and the perspectives

    International Nuclear Information System (INIS)

    Anisimov, O.; Kozlov, Y.; Razmashkin, N.; Safutin, V.; Tikhonov, N.

    2006-01-01

    In the Russian Federation WWER-440 Spent Nuclear Fuel (SNF) is reprocessed at RT-1 plant near Cheliabinsk. WWER-1000 SNF is supposed to be reprocessed at RT-2 plant, which will be built about 2020. The information on the capacity and fill up level of the at-reactor pools at NPP with WWER reactors considering its modification up to May 2005 is given. The regulatory requirements to all SNF 'wet' storage facilities; the principle design and engineering solutions as well as the complex of measures for radiation safety and the environmental protection of spent fuel storage are presented. WWER-440 SNF management, WWER-1000 SNF management and dry storage of WWER-1000 SNF are discussed. In the conclusion it is noted than neither Russia, nor any other country have the experience of construction of vault-type 'dry' storage facilities of such a capacity to store WWER-1000 SNF (9000 tU). The experience and design solutions approved earlier in creation of other dangerous facilities were used. The calculations were based on conservative assumptions allowing with a large assurance to guarantee the nuclear and radiation safety and the environmental protection. At present, a program is developed for scientific-technical support of the dry storage facility design and operation, aimed at the studies whose results will allow to optimize the taken technical decisions, simplify SNF management technology and, possibly, to reduce the cost of the storage facility itself

  17. Use of NESTLE computer code for NPP transition process analysis

    International Nuclear Information System (INIS)

    Gal'chenko, V.V.

    2001-01-01

    A newly created WWER-440 reactor model with use NESTLE code is discussed. Results of 'fast' and 'slow' transition processes based on it are presented. This model was developed for Rovno NPP reactor and it can be used also for WWER-1000 reactor in Zaporozhe NPP

  18. Effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu K; Malachenko, L L; Medvedev, A V; Solyany, V I; Sukhanov, G I; Tonkov, V Yu

    1987-05-01

    Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. 4 references, 7 figures, 3 tables.

  19. The effect of reference parameters and properties of materials for WWER-type fuel elements on their reliability

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Malachenko, L.L.; Medvedev, A.V.; Solyany, V.I.; Sukhanov, G.I.; Tonkov, V.Yu.

    1987-01-01

    Present approach to requirements for reference parameters and properties of materials for WWER-1000 fuel elements is presented as well as evaluation of their effects on fuel reliability. Some results of investigations with the aim of improving fuel element reliability in operational NPP conditions are discussed. (author)

  20. Water chemistry-related activities at the IAEA

    International Nuclear Information System (INIS)

    Cheng, H.; Onufriev, V.

    2005-01-01

    Water chemistry activities and publications in the past are listed. IAEA Coordinated Research Programmes, WWER-1000 SG water chemistry database, materials issues TM in Vienna, TC workshops and attendance of international meetings, publications. There is a list of IAEA publications related to water chemistry and corrosion. Finally water chemistry activities planned for 2006-2008 are detailed. (N.T.)

  1. Organizational aspects of recruiting, training, maintaining the level of professional skills and retraining of NPP operations personnel in Russia

    International Nuclear Information System (INIS)

    Veltchinsky, V.

    1993-01-01

    The organization of personnel work during WWER-1000 operation is presented as recommended by the Russian operating organization. The system is described of recruiting, training control of professional activities, maintaining of professional skills and retraining of the Russian nuclear power plant operation personnel (PKPO system). The basic documentation of the PKPO system is listed. (Z.S.) 1 fig

  2. Maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    Lashgari, Farbod.

    1995-01-01

    This paper is about maintenance of nuclear power plants. In part one, the outage management of nuclear power plants has described. Meaning of the outage and objectives of outage management is given in introduction. The necessity of a long-term outage strategy is shown in chapter one. The main parts of an outage are as follows: Planning; Preparation; Execution, Each of them and also post-outage review have been explained in the followed chapters. Part two deals with technical details of main primary components of nuclear power plant type WWER. After an introduction about WWER reactors, in each chapter first the general and detailed description of main primary components has given and then their maintenance schedules and procedures. Chapter about reactor and steam generator is related to both types of WWER-440 and WWER-1000, but chapter about reactor coolant pump has specified to WWER-1000 to be more in details.(author)

  3. Calculation methods and algorithms development of dynamic loadings on NPP secondary circuit equipment at shock and pulse actions

    International Nuclear Information System (INIS)

    Kuznetsov, D.V.; Kormilitsyn, V.M.; Proskuryakov, K.N.

    2010-01-01

    Calculation results of acoustic parameters fluctuations in low-pressure regenerative heating system of NPP with WWER-1000 type reactor were presented. The spectral structure of acoustic fluctuations was shown to depend on configuration of secondary circuit equipment, its geometrical sizes and operation mode. Estimations of natural oscillations frequencies of working medium pressure in the secondary circuit equipment were resulted. The developed calculation methods and algorithms are intended for revealing and prevention of initiation conditions of vibrations resonances in elements of the secondary circuit equipment with acoustic oscillations in working medium, both under operating conditions and in the design stage of the second circuit of NPP with WWER-1000 type reactor. Analysis of pass-band dependence on operation mode was carried out to solve the given problem [ru

  4. Experience of TVSA fuel implementation at Kozloduy NPP

    International Nuclear Information System (INIS)

    Kamenov, K.; Kamenov, AI.; Hristov, D.

    2011-01-01

    The base design of the Russian fuel assemblies TVSA have been under operation at Kozloduy NPP WWER-1000 reactors since 2004. The old type fuel assemblies TVS-M were gradually substituted till 2008. The TVSA assembly distinguishes itself with much stronger construction. As a burnable absorber it has a mixture of uranium and uniformly distributed Gd in 6 or more fuel rods. This enables to increase the safety and effectiveness of fuel cycles. The experience gained during TVSA fuel implementation on units 5 and 6 and KASKAD code package validation was presented at the eightieth International conference on WWER 'Fuel performance, modelling and experimental support in 2009'. Additional information about TVSA fuel implementation at Kozloduy NPP WWER-1000 units in a 4-year fuel cycle with 42 and 48 fresh fuel assemblies reloading scheme is presented in the paper. (Authors)

  5. Approaches for accounting and prediction of fast neutron fluence on WWER pressure vessels and results of validation of calculational procedure

    International Nuclear Information System (INIS)

    Borodkin, P.G.; Khrennikov, N.N.; Ryabinin, Yu.A.; Adeev, V.A.

    2015-01-01

    A description is given of the universal procedure for calculation of fast neutron fluence (FNF) on WWER vessels. Approbation of the calculation procedure was carried out by comparing the calculation results for this procedure and measurements on the outer surface of the WWER-440 and WWER-1000 vessels. In addition, an estimation of the uncertainty of the settlement procedure was made in accordance with the requirements of regulatory documents. The developed procedure is applied at Kola NPP for independent fast neutron fluence estimates on the WWER-440 reactor vessels when planning core loads taking into account the introduction of new fuels. The results of the pilot operation of the procedure for calculating FNF at the Kola NPP were taken into account when improving the procedure and its application to the calculations of FNF on the WWER-1000 vessels [ru

  6. Some aspects of nuclear fuel use at Ukrainian NPPs during last two years

    International Nuclear Information System (INIS)

    Bilodid, Y.; Shevchenko, I.; Ieremenko, M.; Ovdiienko, I.

    2015-01-01

    For many years SSTC NRS actively participates in licensing of fuel reloading and in the implementation of new nuclear fuel types at the nuclear power plants in Ukraine. Results of the nuclear fuel use for last years are presented in the paper. The results are based on NPP documentation submitted for licensing to the regulating body of Ukraine and based on our estimations and independent calculations. The first part of the paper contains a brief characteristic of the fuel cycles at Ukrainian NPPs. Types of loaded fuel are described also. Experience of new fuel type implementation is presented (Westinghouse FA and TVSA-12 for WWER-1000 reactors). The next part of the paper presents a new regulatory document under development and further new fuel implementation (WWER-1000 reactors). The last part of the paper describes some issues with fuel use. (authors) Keywords: WWER, TVSA, TVSA-12, TVS-W, TVS-WR, Westinghouse, NPP

  7. Calculational results for radiation embrittlement of WWER pressure vessel at the Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Apostolov, T; Ilieva, K; Petrova, T [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    Determination of radiation impact on metal state in the case of WWER-440/230 is made only by calculation methods since a special sample-witness (SW) incorporation had not been implemented. In WWER-1000 reactors such SW are foreseen but their spots are high above the active core. This is why in both reactors the appliance of a calculational procedure for radiation embrittlement determination is compulsory. The authors propose such a procedure accounting for the change in critical temperature of neutron brittleness by the neutron fluence. The neutron fluence and the shift of critical embrittlement temperature have been calculated for the maximum overloaded location and for the weld metal of the Kozloduy-5 and Kozloduy-6 reactors (WWER-1000). The shift of critical temperature in weld 4 of the Units 1-4 (WWER-440) is plotted versus work cycles and compared to experimental values. 4 figs., 5 tabs.

  8. Experience in scientific guidance during start-up and normal operation of NPPs with WWER reactors; Opyt nauchnogo rukovodstva puskom i soprovozhdeniya pri ehkspluatatsii ehnergoblokov AEhS s WWER

    Energy Technology Data Exchange (ETDEWEB)

    Abagyan, A; Kazakov, V; Kryakvin, L [All-Russian Inst. for NPP Operation, Moscow (Russian Federation)

    1996-12-31

    The experience gained in the All-Russian Research Institute of NPP Operation (RU) in improvement of reactor start-up and normal operation of NPPs has been reported. The work was carried out on NPPs in Russia, Ukraine, Bulgaria, Hungary and Czechoslovakia. All scientific programmes applied use the system approach, modern methods of safety assessment and the principle of loss effect optimization. For the Kozloduy NPP the following projects have been implemented: design of accelerated unloading system for WWER-1000; refinement of water feeding unit for the WWER-1000 steam generator; response analysis of the safety anti-breakdown system of the second loop. The importance of technical substantiation in the process of decision making is stressed. 6 refs.

  9. Probability safety assessment of the Kozloduy-5 and Kozloduy-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, A; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    A probability safety assessment (PSA) of Level 1 (assessment of plant failures leading to the determination of core damage frequency) has been carried out for the NPP Kozloduy Units 5 and 6 (reactors WWER-1000). The scope of the study includes all significant accident initiators including seismic (earthquake) and fire initiators. Event trees for all initiators and fault trees for front line systems, support systems and major safety systems have been built. A distribution of the different initiators has been established as follows: internal initiators - 85%, seismic initiators - 5%, fire initiators- 10%. The loss of offsite power was identified as main contributor from the internal initiators with frequency 1,1.10{sup -4}/y. It is concluded that the safety functions of WWER-1000 are adequately covered by the safety systems. 4 refs., 2 tabs.

  10. Adaptation the Abaqus thermomechanics code to simulate 3D multipellet steady and transient WWER fuel rod behavior

    International Nuclear Information System (INIS)

    Kuznetsov, A.V.; Kuznetsov, V.I.; Krupkin, A.V.; Novikov, V.V.

    2015-01-01

    The study of Abaqus technology capabilities for modeling the behavior of the WWER-1000 fuel element for the campaign, taking into account the following features: multi-contact thermomechanical interaction of fuel pellet and fuel can, accounting for creep and swelling of fuel, consideration of creep of the can, setting the mechanisms of thermophysical and mechanical behavior of the fuel - cladding gap. The code was tested on the following developed finite element models: 3D fuel element model with five fuel pellets, 3D fuel element model with one fuel pellet and cleavage in the gap, 3D model of the fuel rod section with one randomly fragmented tablet. The position of the WWER-1000 fuel rod section in the middle of the core and the loads and material properties corresponding to this location were considered. The principal possibility of using Abaqus technology for solving fuel design problems is shown [ru

  11. Specific features of the determination of the pellet-cladding gap of the fuel rods by non-destructive method

    International Nuclear Information System (INIS)

    Amosov, S.V.; Pavlov, S.V.

    2002-01-01

    This report describes the specific features of determining the pellet-cladding gap of the irradiated WWER-1000 fuel rods by nondestructive method. The method is based on the elastic radial deformation of the cladding up to its contact with the fuel. The value of deformation of cladding till its contacting fuel when radial force changes from F max to 0 is proposed as a measuring parameter for determination of the diametrical gap. Because of the features of compression method, the obtained gap value is not analog of the gap measured on micrograph of the fuel rod cross-section. Results of metallography can provide only qualitative evaluation of its method efficiency. Comparison of the values determined by non-destructive method and metallography for WWER-1000 fuel rods with burnup from 25 to 55 MWd/kg U testified that the results of compression method can be used as a low estimate of the pellet-cladding gap value. (author)

  12. NPP Temelin. Status of safety improvements

    International Nuclear Information System (INIS)

    1999-01-01

    The WWER-1000 Temelin NPP under construction has been subjected as other NPPs of the same type to numerous project reviews resulting in quite a number of recommendations for design changes. Results of the IAEA mission to review the resolution of WWER-1000 safety issues at Temelin NPP are cited in this paper. The main conclusions emphasize that a combination of eastern and western technology and practices led to safety improvements in comparison with the international practices. Plant managers are clearly committed to implementation of operational programs which are consistent with effective western operational safety practices. Considerable effort remains to bring planned programs to successful implementation, in particular in meeting the need to foster strong safety culture among all personnel

  13. Information resources for nuclear power construction

    International Nuclear Information System (INIS)

    Rozin, V.M.

    1987-01-01

    Problems of designing and introduction of data resources (in the form of magnetic carrier data) for the construction of NPP with the WWER-1000 type reactors are considered. It is suggested that the data resources should be supplied to the construction as independent production alongside with preliminary estimation and material resources. It is pointed out that the introduction of the data resource will permit to increase the efficiency of NPP construction management

  14. Rossendorf Research Center, Institute of Safety Research. Annual report 1991

    International Nuclear Information System (INIS)

    Boehmert, J.; Weiss, F.P.

    1992-08-01

    The working program covers above all topics concerning the assessment of design basis safety and the increase of operational safety of the WWER type reactors. The topics are directed to the WWER-440/213 type and to the WWER-1000 type, and are dealt with by the three departments, i.e. incident analysis, neutron embrittlement, and mechanical integrity. One paper is concerned with the determination of the neutron field of HERA. (orig.) [de

  15. Nuclear power development

    International Nuclear Information System (INIS)

    Povolny, M.

    1980-01-01

    The development and uses of nuclear power in Czechoslovakia and other countries are briefly outlined. In the first stage, the Czechoslovak nuclear programme was oriented to the WWER 440 type reactor while the second stage of the nuclear power plant construction is oriented to the WWER 10O0 type reactor. It is envisaged that 12 WWER 440 type reactors and four to five WWER 1000 type reactors will be commissioned till 1990. (J.P.)

  16. WWER reactor physics code applications

    International Nuclear Information System (INIS)

    Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.

    1994-01-01

    The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs

  17. On the advisability of nuclear power plant construction and some solutions on their safety increase

    International Nuclear Information System (INIS)

    Sokolov, I.N.

    1989-01-01

    Neccesity in a further development of nuclear power in the USSR is proved. Specific features of layouts of nuclear heat supplying plants with increased safety including WWER-440, WWER-1000 and VK-50 reactors are considered. It is stressed that even under the most incredible accidents referring to the class of hypothetic ones, radioactive effect on the population, according to evaluations by experts, whould be much lower than maximum admissible one according to the existing standards

  18. Some aspects of nuclear fuel utilisation at Ukrainian's NPPs during last two years

    International Nuclear Information System (INIS)

    Ieremenko, M.; Bilodid, Y.; Ovdiienko, Y.

    2011-01-01

    In the first part of the report the brief characteristic of the realized fuel cycles on Ukraine's NPPs, types of loaded fuel is described. Experience of new fuel type implementation are present (FA Second Generation for WWER-440. Westinghouse FA for WWER-1000). Next issue of report is the some of problem with fuel utilisation (leakage FA, problem with C b calculation and other). And the last issue of report is presentation of future new fuel implementation (WWER-440/1000). (authors)

  19. Comparative analysis of methods for workload assessment of the main control room operators of NPP

    International Nuclear Information System (INIS)

    Georgiev, V.; Petkov, G.

    2008-01-01

    The paper presents benchmarking workload results obtained by a method for operator workload assessment – NASA Task Load Index, and a method for human error probability assessment - Performance Evaluation of Teamwork. Based on the archives of FSS-1000 training in the accident “Main Steam Line Tube Rupture at the WWER-1000 Containment” the capacities of the two methods for direct and indirect workload assessment are evaluated

  20. Status of the PSA use in the Czech regulatory process

    International Nuclear Information System (INIS)

    Dusek, J.

    1994-01-01

    A review of previous probabilistic safety assessment (PSA) activities initiated by regulatory body and preparation of the preliminary PSA study and final PSA study (released in January 1994) for the nuclear power plant Dukovany with WWER-440 type 213 reactor is described. A brief information about the NPP Temelin (with WWER-1000) PSA Study, shutdown and PSA risk monitor current activities for the NPP Dukovany, next PSA activities in 1994 and about planned PSA activities in future is attached. (author). 21 refs

  1. Testing of components on the shaking table facilities of AEP and contribution to full scale dynamic testing of Kozloduy NPP. Final report

    International Nuclear Information System (INIS)

    Ambriashvili, Y.

    1999-01-01

    Units 5 and 6 of the Kozloduy NPP were designed as a standardized WWER-1000 on the basis standardized scheme designs, including building structures and reactor building. The accelerograms and response spectrum were adopted based on seismological, geological, geophysical and other conditions of the region where the NPP is expected to be located. This paper presents the results design floor response spectra analysis as well as method and some results of the tests of electrical equipment as identified by the Kozloduy NPP staff

  2. Manufacture of rings of 08Kh18N10T sheet for internal structures of WWER type reactors

    International Nuclear Information System (INIS)

    Fojta, A.; Nitka, B.

    1984-01-01

    Technology is presented of the manufacture of rings for the jacket, shaft, core catcher and shaft bottom of WWER-440 reactors produced by Vitkovice Steel Works. The rings are manufactured from sheets of austenitic steel 08Kh18N10T. The materials and technology problems are discussed of sheet production, ring welding technology and annealing following welding. The plastic properties are assessed of the welded joints and problems are outlined of ring production for WWER-1000 reactors. (B.S.)

  3. Proceedings of the twenty-first symposium of atomic energy research on WWER physics and reactor safety

    International Nuclear Information System (INIS)

    Vidovszky, I.

    2011-10-01

    The present volume contains 61 papers, presented on the twenty-first symposium of atomic energy research, held in Dresden, Germany, 19-23 September 2011. The papers are presented in their original form, i. e. no corrections or modifications were carried out. The content of this volume is divided into thematic groups: Improvement, extension and validation of parameterized few-group libraries for WWER-440 and WWER-1000.

  4. Design safety improvements of Kozloduy NPP

    International Nuclear Information System (INIS)

    Hinovski, I.

    1999-01-01

    Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

  5. Mathematical models in nuclear safety and radiation protection of nuclear energy

    International Nuclear Information System (INIS)

    1993-01-01

    This collection of papers contains the full texts of 33 reports presented at the Seminar, all of which are indexed and abstracted separately for the INIS database. The topics of the reports cover the mathematical models and computer codes for risk analysis, reactor reliability simulating, safety-related benchmarks, thermohydraulic studies, reactor kinetics, hypothetical accidents and their radiological consequences, etc. The investigations refer mainly to WWER-440 and WWER-1000 type reactors of the Kozloduy NPP

  6. On possibility of application of the parallel-mixed type coolant flow scheme to NPP steam generators linked with superheaters

    International Nuclear Information System (INIS)

    Malkis, V.A.; Lokshin, V.A.

    1983-01-01

    Optimum distribution of the coolant straight-through flow between the superheater, evaporator and economizer is determined and the parallel-mixed type flow scheme is compared with other schemes. The calculations are performed for the 250 MW(e) steam generator for the WWER-1000 reactor unit the inlet and outlet primary coolant temperature of which is 324 and 290 deg C, respectively, while the feed water and saturation temperatures are 220 and 278.5 deg C, respectively. The rated superheating temperature is 300 deg C. The comparison of different schemes has been performed according to the average temperature head value at the steam-generator under the condition of equality as well as essential difference in the heat transfer coefficients in certain steam-generator sections. The calculations have shown that the use of parallel-mixed type flow permits to essentially increase the temperature head of the steam generator. At a constant heat transfer coefficient in all steam generator sections the highest temperature head is reached. At relative flow rates in the steam generator, economizer and evaporator equal to 6, 8 and 86%, respectively. The superheated steam generator temperature head in this case by 12% exceeds the temperature head of the WWER-1000 reactor unit wet steam generator. In case of heat transfer coefficient reduction in the superheater by a factor of three, the choice of the primary coolant, optimum distribution permits to maintain the steam generator temperature head at the level of the WWER-1000 reactor unit wet-steam steam generator. The use of the parallel-mixed type flow scheme permits to design a steam generator of slightly superheated steam for the parameters of the WWER-1000 unit

  7. High-power condensation turbine application to district heating

    International Nuclear Information System (INIS)

    Virchenko, M.A.; Arkad'ev, B.A.; Ioffe, V.Yu.

    1982-01-01

    In general outline the role of condensation turbines in NPP district heating is considered. The expediency of expansion of central heating loading of turbines of operating as well as newly designed condensation power plants on the basis of the WWER-1000-type reactors is shown. The principle heat flowsheet of the 1000 MW power turbine is given. An advantage in using turbines with uncontrolled steam bleeding is pointed out [ru

  8. Recent activities in the field of nuclear power plant control and instrumentation in Czechoslovakia

    International Nuclear Information System (INIS)

    Rubek, J.

    1992-01-01

    The report presents a review of Czechoslovak nuclear power plants that are in operation and in the course of construction. The present state of the instrumentation of the newly built NPP's is described, and a special attention is given to work on the control of spatial power distribution in the reactor core of WWER 1000, the first unit to be installed in NPP Temelin. A project of a secondary circuit diagnostic system of this unit is described. (author). 6 refs, 1 tab

  9. Country report: Bulgaria

    International Nuclear Information System (INIS)

    1993-01-01

    Bulgaria has only one site with operating nuclear energy reactors - Kozloduy nuclear power plant with six units in operation, four of them are WWER-440 type and two are WWER-1000. Spent fuel is to be stored for a three year period in water pools at reactors and after that to be transported to the away from reactor storage (AFRS) or back to the supplier. The AFRS operator has not been granted operating license until more stringent safety requirements will be met

  10. Specific features of accounting probable exercussions of power in periphery WWER rods caused by in-process gap changes between fuel assemblies

    International Nuclear Information System (INIS)

    Mikailov, E. F.; Shishkov, L. K.

    2011-01-01

    The paper discusses the way for accounting the uncertainties of calculated WWER-1000 rod powers caused by the changes of fuel assembly shape in the course of operation. The trouble is that the gap affected power distribution does not obey the normal distribution law, while the dispersion does not influence the total uncertainty in power distribution. The paper proposes the methods for accounting the uncertainty and gives the exercise. (Authors)

  11. FERONIA-CZ/PHARE Progress Report, January 1997 - December 1997

    International Nuclear Information System (INIS)

    Lassmann, K.; Laar, J. van de; Miasnikov, A.; Svoboda, R.; Valach, M.

    1998-02-01

    The report describes work performed within the programme 'Assistance to the State Office for Nuclear Safety of the Czech Republic by European Institute for Transuranium Elements Karlsruhe in Performing Fuel Rod Thermo-Mechanical Behaviour Analyses for the licensing process support, FERONIA (acronym for Fuel Rod Modelling and Performance)'. The main objective is a reliable prediction of fuel performance at the Temelin WWER-1000 units and Dukovany WWER-440 units. The TRANSURANUS code was employed

  12. The effect of the PWR secondary circuit water chemistry on erosion corrosion

    International Nuclear Information System (INIS)

    Kaplan, J.

    1993-07-01

    The secondary circuit of WWER-440 and WWER-1000 reactors is described. The causes of erosion corrosion are outlined, and the effects of the physical properties and chemical composition of water are discussed with emphasis on specific conductivity and concentrations of oxygen, ammonia, iron, sodium, silicon and organics. Described are corrective actions to eliminate the deviations from the normal state during reactor power reduction or reactor shutdown. (J.B.)

  13. Perspectives for utilization of the experience from the electrical equipment modernization of Kozloduy NPP units 5 and 6 in case of construction of new NPP

    International Nuclear Information System (INIS)

    Bogdanov, D.

    2005-01-01

    In 2005 a governmental decision for continue the project for construction of Belene NPP was adopted. In case of start of this significant project in short terms, a good option will be to utilize the already achieved experience by the modernization program of units 5 and 6 of Kozloduy NPP. The process of construction of new NPP is associated with significant amount of engineering, civil construction works, equipment order and delivery, installation works, testing and commissioning activities. It is vital necessity to have optimal technology, strict organization for project implementation and excellent documentation. From technical point of view the experience from the modernization of units 5 and 6 of Kozloduy NPP will be very important in case of utilization of similar types of WWER 1000 units. According to the 'Feasibility study for construction of Belene NPP', the most favorable option is to accomplish the already partially constructed Unit 1 on the basis of upgraded WWER B-320 design - WWER 1000 B-392, in order to fulfill IEC 331-3A, IEC 332-3A, and of US NRC Regulatory Guide 1.70 and to construct the second unit on the basis of ASE WWER/1000 B-466. However, for any type of unit selected, the achieved experience will be an important basis. This article is focused mostly on the modernization aspects of some electrical systems of Kozloduy NPP

  14. Information about AER WG A on improvement, extension and validation of parametrized few-group libraries for VVER 440 and VVER 1000

    International Nuclear Information System (INIS)

    Mikolas, P.

    2010-01-01

    Joint AER Working Group A on 'Improvement, extension and validation of parameterized few-group libraries for WWER-440 and WWER-1000' and AER Working Group B on 'Core design' nineteenth meeting was hosted by VUJE a. s. in Modra - Harmonia (Slovakia) during the period of 20. to 22. April 2010. There were present altogether 12 participants from 8 member organizations and 9 papers were presented (8 of them in written form). Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-groups libraries preparation, their accuracy and validation. Presentations were devoted to some aspects of transport and diffusion calculations and to the benchmark dealing with WWER-1000 core periphery power tilt. Tamas Parko (co-authors Istvan Pos and Sandor Patai Szabo) described in his presentation 'Application of Discontinuity factors in C-PORCA 7 code', Radoslav Zajac (co-authors Petr Darilek and Vladimir Necas) spoke about 'Fast Reactor Nodalisation in HELIOS Code', Gabriel Farkas presented 'Calculation of Spatial Weighting Functions of Ex-Core Neutron Detectors for WWER-440 Using Monte Carlo Approach' and Daniel Sprinzl (co-authors Vaclav Krysl, Pavel Mikolas and Jiri Svarny) provided a definition of a benchmark in ' 'MIDICORE' WWER-1000 core periphery power tilt benchmark proposal'. (Author)

  15. Express diagnostics of WWER fuel rods at nuclear power plants

    International Nuclear Information System (INIS)

    Pavlov, S.; Amosov, S.; Sagalov, S.; Kostyuchenko, A.

    2009-01-01

    Higher safety and economical efficiency of nuclear power plants (NPP) call for a continuous design modification and technological development of fuel assemblies and fuel rods as well as optimization of their operating conditions. In doing so the efficiency of new fuel introduction depends on the completeness of irradiated fuel data in many respects as well as on the rapidity and cost of such data obtaining. Standard examination techniques of fuel assemblies (FA) and fuel rods (FR) intended for their use in hot cell conditions do not satisfy these requirements in full extent because fuel assemblies require preliminary cooling at NPP to provide their shipment to the research center. Expenditures for FA transportation, capacity of hot cells and expenditures for the examined fuel handling do not make it possible to obtain important information about the condition of fuel assemblies and fuel rods after their operation. In order to increase the comprehensiveness of primary data on fuel assemblies and fuel rods immediately after their removal from the reactor, inspection test facilities are widely used for these purposes. The inspection test facilities make it possible to perform nondestructive inspection of fuel in the NPP cooling pools. Moreover these test facilities can be used to repair failed fuel assemblies. The ultrasonic testing of failed fuel rods inside the fuel assembly was developed for stands of inspection and repair of TVSA WWER-1000 for the Kalinin NPP and Temelin NPP. This method was tested for eight leaking fuel assemblies WWER-440 and WWER-1000 with a burnup of ∼14 up to 38 MW·day/kgU. The ultrasonic testing proved its high degree of reliability and efficiency. The defectoscopy by means of the pulsed eddy-current method was adapted for the stand of inspection and repair of TVSA WWER-1000 for the Kalinin NPP. This method has been used at RIAR as an express testing method of FR claddings during the post-irradiation examinations of fuel assemblies WWER

  16. Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

    International Nuclear Information System (INIS)

    Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

    2001-01-01

    The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

  17. Results of trial operation of the WWER advanced fuel assemblies

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Dragunov, Y.; Mikhalchuk, A.

    2001-01-01

    The paper describes results from experimental operation of advanced WWER-1000 fuel assemblies (AFA) at five units in Balakovo NPP. Advanced fuel is developed according to the concept of standard WWER-1000 fuel assembly (jacket-free). The new features includes: 1) zirconium guiding channels (alloy E-635 and E-110) and spacer grids (alloy E-110); 2) integrated burnable absorber gadolinium; 3) extended service life of fuel assemblies (FA) and absorber rods (possibility of repair of FA); 4) improved adoption to reactor conditions. Some results of AFA pilot operation of a three year operation are presented and analyses of effectiveness of improvements are made concerning application of zirconium channels and grids; application of integrated burnable absorbers; extension of FA and absorbing rods service life and FA repairability. These new features of WWER-1000 fuel design allow: 1) to reduce the average fuel enrichment to the 3.77% instead of 4.31% in U-235; 2) to reduce the FA axial load in reactor hot state by 40%,; 3) increasing of fuel operation in reactor to the 30000 effective days with possibility to have a 5-year residence time in the reactor. The design of new generation FA for WWER-440 reactors involves few key changes. Fuel inventory in new fuel design is increased due to elongation of fuel stack and reducing the diameter of the central hole. Vibration stability is enhanced as a result of: no-play junction of the fuel rod with the lower grid; change of SG arrangements; strengthening of the lower grid unit; secure of the central tube in the gap. Water-uranium ration is increased. Introduction of all these kinds of modernization in a 5-year fuel cycle reduces fuel component in the energy cost to the 7%

  18. Experimental research on safety assurance of advanced WWER fuel cycles

    International Nuclear Information System (INIS)

    Krainov, Ju.; Kukushkin, Ju.

    2002-01-01

    The paper presents the results of experimental investigations on substantiation of implementation of a modernized butt joint for the WWER-440 reactor, carried out in the critical test facility 'P' in the RRC 'Kurchatov Institute'. The comparison results of the calculation and experimental data obtained in the physical startup of Volgodonsk NPP-1 with the WWER-1000 are also given. In the implementation of four-year fuel cycle in the WWER-440 with the average enrichment of fuel makeup 3.82% it was solved to conduct experimental research of power distribution in the vicinity of control rod butt junction. Moreover, it was assumed that adequate actions should be applied to eliminate inadmissible power jumps, if necessary. It is not available to measure their values in NPP conditions. Therefore, the power distribution near the butt joint was studied in a 19-rod bank installed in the critical test facility 'P' first for the normal design of the joint when surrounding fuel assemblies enrichment goes up. Then a set of calculation and tests was fulfilled to optimize a butt junction design. On the base of this research the composition of a butt junction was advanced by placing Hf plates into the junction. The effectiveness of modernized butt joint design was experimentally confirmed. In Volgodonsk NPP-1 with WWER-1000 the four-year fuel cycle is being implemented. During the physical startup of the reactor the measurements of the reactivity effects and coefficients were measured at the minimum controlled flux level, and the parameters of a number of critical states were recorded. The data obtained were compared with the calculation. The validity of the certified code package for forecasting the neutronic characteristics of WWER-1000 cores in the implementation of a four year fuel cycle has been supported (Authors)

  19. Verification calculations for the WWER version of the TRANSURANUS code

    International Nuclear Information System (INIS)

    Elenkov, D.; Boneva, S.; Georgieva, M.; Georgiev, S.; Schubert, A.; Van Uffelen, P.

    2006-01-01

    The paper presents part of the work performed in the study project 'Research and Development for Licensing of Nuclear Fuel in Bulgaria'. The main objective of the project is to provide assistance for solving technical questions of the fuel licensing process in Bulgaria. One important issue is the extension of the predictive capabilities of fuel performance codes for Russian-type WWER reactors. In the last decade, a series of international projects has been based on the TRANSURANUS fuel performance code: Specific models for WWER fuel have been developed and implemented in the code in the late 90's. In 2000-2003, basic verification work was done by using experimental data of nuclear fuel irradiated in WWER-440 reactors. While the present paper focuses on the analysis of WWER-1000 standard fuel under normal operating conditions, the above study project covers additional tasks: 1) Post-irradiation calculations of ramp tests performed in the DR3 test reactor of the Risoe National Laboratory (five instrumented fuel rods of the Risoe 3 dataset contained in the IFPE database) using the TRANSURANUS code; 2) Compilation of cross-section libraries for isotope evolution calculations in WWER-440 and WWER-1000 fuel assemblies using the ORIGEN-S code; 3) Analysis of current situation and needs for an extension of the curriculum in Nuclear Engineering at the Technical University of Sofia. In this paper the post-irradiation calculations of steady-state irradiation experiments with nuclear fuel for Russian-type WWER-1000 reactors, using the latest release of the TRANSURANUS code (v1m1j03)are presented. Regarding a comprehensive verification of modern fuel performance codes, the burn-up region above 40 MWd/kgU is of increasing importance. A number of new phenomena emerge at high fuel burn-up, implying the need for enlarged databases of postirradiation examinations (PIE). For one fuel assembly irradiated in a WWER-1000 reactor with a rod discharge burn-up between 50 and 55 MWd

  20. Instrumentation and control doctrine for VVER upgrading

    International Nuclear Information System (INIS)

    Pernette, R.; Cucciniello, J.; Trebot, R.

    1997-01-01

    The main results are reported of analyses carried out on the WWER-400 and WWER-1000 series, and the reasons are summed up underlying the decision to establish a policy of WWER instrumentation and control upgrading. It is shown that such a policy forms a key stage in any upgrading project and contributes to the improvement of reactor safety at several levels: it helps to optimize the integration of safety requirements; it makes it easier to achieve a high level of automation and a better man-machine interaction; it should raise the culture of reactor safety. Four annexes demonstrate the factual fundamentals and structure of such a policy. (A.K.)

  1. On improving the NPP construction long-term planning

    International Nuclear Information System (INIS)

    Solov'ev, M.M.

    1980-01-01

    General approaches to solving the problems of long-term planning of the construction-mounting work volume and introduction of power capacities at different paces of power unit commission and norms of construction time are considered. Problems associated with location and development of subsidiary industries providing NPP construction are discussed. It is shown that decrease of WWER-1000 reactor type NPP construction time from 24 to 12 months results in reduction of required high-skilled engineers, managerial staff, unique construction mechanisms and engines as well as in improvement of conditions for conducting measures on work arrangement during preparation of construction industry and mounting of the first assemblies

  2. Complex approach to power supply system in the Tomsk region

    International Nuclear Information System (INIS)

    Meshcheryakov, V.N.; Bojko, V.I.; Koshelev, F.P.; Kolpakov, G.N.; Shamanin, I.V.

    2001-01-01

    The problem of power supply of the Tomsk region is under consideration. Power sources are compared with regard to economical and ecological effectiveness. It is shown that nuclear power industry is promising for the Tomsk region due to a unique nuclear park in a town of Seversk with developed scientific, industrial, skilled personnel and construction bases. It is noted that the construction of the NPP based on two WWER-640, WWER-1000 reactors of new generation permits solving the problem of power supply of the Tomsk region [ru

  3. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L. [Inst. of Material Engineering, Ostrava (Switzerland)

    1995-12-31

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

  4. The behaviour of water-cooled reactor fuel rods in steady state and transient conditions; Zachowanie sie pretow paliwowych reaktorow chlodzonych woda w stanach ustalonych i nieustalonych

    Energy Technology Data Exchange (ETDEWEB)

    Strupczewski, A.; Marks, P. [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1997-12-31

    In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author) 38 refs, 40 figs, 15 tabs

  5. NPP capital equipment life control facilities and tools

    International Nuclear Information System (INIS)

    Nekrasov, A.V.; Proskuryakov, K.N.

    2009-01-01

    Algorithms and simulation package for calculation of eigen modes of coolant pressure oscillations in equipment of primary coolant circuit of WWER-1000 NPP, tuned-circuit Q-factor of an oscillating contour of the coolant, fluctuation frequencies pass band and attenuation factors of were developed. Calculation results for cases of presence and absence of gaseous state radiolysis products in the coolant were obtained. Results analysis makes it possible to disclose operating conditions which result in raised vibrations, and to develop control response in NPP industrial control system for their prevention [ru

  6. Design works organization during the IandC system upgrading on an operating VVER power plant from the general designer's standpoint

    International Nuclear Information System (INIS)

    Krepel, V.

    1997-01-01

    Two replacements of nuclear power plant instrumentation and control systems have already been implemented according to the project design of Energoprojekt: at the Mochovce WWER-440 and the Temelin WWER-1000 plants. In the present contribution the design problems encountered during the replacement are described, the causes of the problems are discussed, and a policy of organizing project design work for an instrumentation and control system replacement at an operating WWER power plant is established such as would avoid these problems. (A.K.)

  7. Experimental investigations and seismic analyses for benchmark study of 1000 MW WWER type (water-cooled and moderated reactor) nuclear power plant Kozloduy. Final report 15 June 1993 - 14 June 1994

    International Nuclear Information System (INIS)

    Sachansky, S.

    1995-01-01

    This report includes preparation and compilation of all existing studies related to seismic safety assessment of Kozloduy WWER-1000, i.e. Units 5 and 6; description of previous full scale testing of Unit 5; and the results obtained from seismic analyses performed under benchmark experimental studies. The results are concerned with analysis of the geological conditions; analysis of the seismic wave velocities in the soil layers; analysis of the predominant natural periods; dynamic characteristics of the Unit 5; soil-structure interaction and laboratory testing and analysis of the reactor containment tenders

  8. Studies on the behaviour of heat transfer flowsheets with fluids

    International Nuclear Information System (INIS)

    Munser, H.

    1989-01-01

    A model is presented which provides a dynamic description of plant components in the secondary coolant circuit of WWER-1000-type nuclear power stations. These components employ combined heat exchangers as their main constituents which themselves are composed of simple water-water and steam-water heat exchangers. The simple heat exchangers are described by means of linearized sets of partial differential equations which can be solved by Laplace transformation. The code system DAW has been established for dynamic modulation of plant components. (author)

  9. Status and objectives of the Bulgarian nuclear energy sector

    International Nuclear Information System (INIS)

    Kouzmanov, K.

    1996-01-01

    A brief overview of the current situation and problems to be solved in the Bulgarian nuclear energy is presented. The main task is to increase the availability of the Kozloduy NPP in terms of time and power. The following technical solutions for reconstruction of units 1-4 (WWER-440) are discussed: defence against external events; emergency cooling of the primary and secondary circuits; reliability analysis used in probabilistic safety assessment; management of beyond design basis accident conditions. The necessity of retrofitting of units 5-6 (WWER-1000) and completion of the Belyane NPP project is also stressed

  10. Refurbishing the reactor protection systems of VVER-440/230 and VVER-1000/320 nuclear power plants with exclusively digital IandC systems

    International Nuclear Information System (INIS)

    Martin, M.

    1997-01-01

    The refurbishment of reactor protection systems of nuclear power plants is based on two sets of requirements: engineering aspects such as performance, qualification and licensing, as well as interfaces to other systems; and cost-benefit relationships, ease of service and maintenance as well as installation during scheduled outages. A number of WWER-440 and WWER-1000 nuclear plants have announced their intention to refurbish their protection systems. Since 1994, these plants have been placing orders with Siemens for new protection systems, including the neutron flux monitoring system utilizing the advanced system TELEPERM XS. This exclusively digital IandC system provides an excellent foundation for the remaining plant service life

  11. Determination of reactor parameters during start up test at the Taiwan NPP, Unit 1

    International Nuclear Information System (INIS)

    Astakhov, S.; Kravchenko, A.; Kraynov, Ju.; Nasedkin, A.; Tsyganov, S.

    2006-01-01

    Unit 1 of Taiwan NPP with WWER-1000 reactor reached the first criticality at December 20 of 2005. Series of start up experiments were carried out under scientific advisory of RRC 'Kurchatov Institute' specialists. At the Hot Zero Power state the reactivity coefficients, control rod group and scram worth were measured and symmetry of the core loading and power reactivity effect due to reaching 1% of nominal were assessed. Paper describes in brief special features of experiments and presented some results obtained at a measurements (Authors)

  12. LWR physics in SKODA Works

    International Nuclear Information System (INIS)

    Zbytovsky, A.; Lehmann, M.; Vyskocil, V.; Vacek, J.; Krysl, V.

    1980-01-01

    Computation of nuclear power reactors of the WWER-1000 type is described as are computer programs used by Skoda Works for the solution of neutron problems. The programs are analyzed for applicability in the unified program system of the CMEA countries which will be used in the preparation of safety reports, the evaluation of safety hazards, the design of fuel charges, economical studies etc. A detailed description is also presented of multigroup transport calculations and of the preparation of input data for macrocalculations of the heterogeneous lattices of LWR's. (author)

  13. The behaviour of water-cooled reactor fuel rods in steady state and transient conditions

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.

    1997-01-01

    In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author)

  14. Standards, criteria, comparative study. Final report 1993-1994

    International Nuclear Information System (INIS)

    David, M.

    1995-01-01

    The goal of this report is to compare standards and criteria used for the design of seismic resistant Nuclear Power plants in former USSR and Eastern Europe to the actual practice in Western Europe and USA. A review of USSR and Czech-Slovak standards for design of WWER-1000 NPPs is included. Comparative study deals involves a general comparison of standards and the comparison of seismic loads according to Eastern standards with the to-day practices. According to these comparisons it can be theoretically concluded that despite some differences the structural seismic forces calculated according to Soviet-Russian standards are in good agreement with the ones resulting from Western Standards

  15. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Pappx, L.

    1994-01-01

    After modification of Dukovany NPP steam generator feedwater system, the increased concentration of minerals was measured in the cold leg of modified steam generator. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators, has focused this attention on the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of flow distribution in the secondary side of SG was developed. (Author)

  16. Designing a nuclear power plant with 1000 MW WWER-type units

    Energy Technology Data Exchange (ETDEWEB)

    Berkovich, V; Kaloshin, J; Tatarnikov, V; Shenderovich, A

    1977-06-01

    A brief description is presented of a WWER-1000 nuclear power plant also considering its environmental impact and the problem of core poisoning. The following indicators are graphically shown in relation to the reactor output: turbogenerator unit outputs, efficiency, specific capital costs and own costs of electric power generated by the Voronezh nuclear power plant. Also listed are the specific consumption of metal and concrete, specific equipment weight and the specific volume of the buildings of the main generating unit as well as the cross section thereof.

  17. Designing a nuclear power plant with 1000 MW WWER-type units

    International Nuclear Information System (INIS)

    Berkovich, V.; Kaloshin, J.; Tatarnikov, V.; Shenderovich, A.

    1977-01-01

    A brief description is presented of a WWER-1000 nuclear power plant also considering its environmental impact and the problem of core poisoning. The following indicators are graphically shown in relation to the reactor output: turbogenerator unit outputs, efficiency, specific capital costs and own costs of electric power generated by the Voronezh nuclear power plant. Also listed are the specific consumption of metal and concrete, specific equipment weight and the specific volume of the buildings of the main generating unit as well as the cross section thereof. (J.B.)

  18. Analysis and planning of the utilization of nuclear power plants

    International Nuclear Information System (INIS)

    Skvarka, P.

    1985-01-01

    The utilization coefficient as one of the characteristics of availability of nuclear power plants and the operation results (like maximum power, block number, and electric energy generation) are investigated by different statistic methods for several nuclear power plants with PWR type reactors and compared with those of WWER 440-type reactors. By means of linear many-parameter regression analysis the utilization coefficient is studied in dependence on block power and time after reactor commissioning. Forecastings of mean utilization coefficients are presented for the power of WWER 1000-type reactors

  19. Seismic soil-structure-equipment interaction analysis of unit 5/6, Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M [Bulgarian Academy of Sciences, Central Laboratory for Seismic Mechanics and Earthquake Engineering, Sofia (Bulgaria)

    1995-07-01

    This research project is aimed to analyse problems of soil-structure-equipment interaction under seismic excitation in case of Kozloduy NPP. Reevaluation and upgrading of Kozloduy NPP has started after 1977 Vrancea earthquake. New Safe Shutdown Earthquake (SSE) level was defined, upgrading most of structural equipment was performed, seismic instrumentation was installed. New investigations were initiated after 1990 IAEA mission visited the site. A comprehensive site confirmation project was started with a subsequent structural and equipment reevaluation and upgrading. This work deals with Units 5 and 6 of WWER-1000 type only.

  20. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L [Inst. of Material Engineering, Ostrava (Switzerland)

    1996-12-31

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

  1. Safety parameter display system (SPDS) for Russian-designed NPPs

    International Nuclear Information System (INIS)

    Anikanov, S.S.; Catullo, W.J.; Pelusi, J.L.

    1997-01-01

    As part of the programs aimed at improving the safety of Russian-designed reactors, the US DoE has sponsored a project of providing a safety parameter display system (SPDS) for nuclear power plants with such reactors. The present paper is focused mostly on the system architecture design features of SPDS systems for WWER-1000 and RBMK-1000 reactors. The function and the operating modes of the SPDS are outlined, and a description of the display system is given. The system architecture and system design of both an integrated and a stand-alone IandC system is explained. (A.K.)

  2. Experimental investigations and seismic analyses for benchmark study of 1000 MW WWER type (water-cooled and moderated reactor) nuclear power plant Kozloduy. Final report 15 June 1993 - 14 June 1994

    Energy Technology Data Exchange (ETDEWEB)

    Sachansky, S [Building Research Institute (NISI), Sofia (Bulgaria)

    1995-07-01

    This report includes preparation and compilation of all existing studies related to seismic safety assessment of Kozloduy WWER-1000, i.e. Units 5 and 6; description of previous full scale testing of Unit 5; and the results obtained from seismic analyses performed under benchmark experimental studies. The results are concerned with analysis of the geological conditions; analysis of the seismic wave velocities in the soil layers; analysis of the predominant natural periods; dynamic characteristics of the Unit 5; soil-structure interaction and laboratory testing and analysis of the reactor containment tenders.

  3. Derivation of design response spectra for analysis and testing of components and systems

    International Nuclear Information System (INIS)

    Krutzik, N.

    1996-01-01

    Some institutions participating in the Benchmark Project performed parallel calculations for the WWER-1000 Kozloduy NPP. The investigations were based on various mathematical models and procedures for consideration of soil-structure interaction effects, simultaneously applying uniform soil dynamic and seismological input data. The methods, mathematical models and dynamic response results were evaluated and discussed in detail and finally compared by means of different structural models and soil representations with the aim of deriving final enveloped and smoothed dynamic response data (benchmark response spectra). This should be used for requalification by analysis testing of the mechanical and electrical components and systems located in this type of reactor building

  4. Status and prospects of WWER in-core fuel management activities

    International Nuclear Information System (INIS)

    Novikov, A.N.; Pavlov, V.I.; Pavlovichev, A.M.; Proselkov, V.N.; Saprykin, V.V.

    1994-01-01

    A short review is given of recent extensive calculational and experimental studies carried out in Russia and Bulgaria for WWER fuel cycle modernization. The main activities performed at Kola NPP, Novovoronezh NPP, Kozloduy NPP and Balakovo NPP are outlined. Based on experience gained, the following improvements in the fuel cycle have been introduced: 1) increased fuel burnup; 2) reduced natural uranium consumption and decreased amount of separation work per energy output unit; 3) increased efficiency of the reactor emergency protection; 4) reduced fast neutron flux onto the reactor vessel. The main characteristics of modernized fuel cycles of WWER-440 and WWER-1000 are presented. 4 tabs., 3 figs., 14 refs

  5. Safety investigation of team performance in accidents

    International Nuclear Information System (INIS)

    Petkov, G.; Todorov, V.; Takov, T.; Petrov, V.; Stoychev, K.; Vladimirov, V.; Chukov, I.

    2004-01-01

    The paper presents the capacities of the performance evaluation of teamwork (PET) method. Its practicability and efficiency are illustrated by retrospective human reliability analyse of the famous nuclear and maritime accidents. A quantitative assessment of operators' performance on the base of thermo-hydraulic (T/H) calculations and full-scope simulator data for set of NPP design basic accidents with WWER is demonstrated. The last data are obtained on the 'WWER-1000' full-scope simulator of Kozloduy NPP during the regular practical training of the operators' teams. An outlook on the 'evaluation system of main control room (MCR) operators' reliability' project, based on simulator data of operators' training is given

  6. Determining the Residual Lifetime of Qualified Electrical and I and C Equipment in Units 5 and 6 of Kozloduy NPP

    International Nuclear Information System (INIS)

    Radoslavov, Biser

    2012-01-01

    The proposed paper clarifies the environmental qualification and residual lifetime of installed equipment and electrical cables in operating WWER 1000 type units 5 and 6 of Kozloduy NPP. The qualification approach is a combination of analyses and tests in consideration of some specific features of WWER equipment. As a result of the study, selected equipment is divided in 26 groups. Qualification groups include similar types of equipment with equal resistance to foreign impacts, analogous critical parts, and same methods of control and estimation of residual lifetime. Qualification protocols were made for each group. (author)

  7. International Economic Association on organization of co-operative production and development of equipment and providing technical assistance in construction of nuclear power plants - ''INTERATOMENERGO''

    International Nuclear Information System (INIS)

    Mal'tsev, N.D.

    1979-01-01

    History is stated of foundation of the International Economic Association ''Interatomenergo''. Structure is given of the Association and the list of main problems to be solved by it. Project is given of the programm of co-operation in the field of scientific and technical works as well as of design and projecting works in creation of new types of equipment for nuclear power plants, in particular, creation of serial power units with improved WWER-1000 reactor. Directions are stated of activity of the Association in the field of providing assistance in construction and exploitation of nuclear power plants as well as in training of operational personnel [ru

  8. Advances in new WWER designs to improve operation and maintenance

    International Nuclear Information System (INIS)

    Dragunov, Y.G.; Ryzhov, S.B.; Podshibiakin, A.K.; Vasilchenko, I.N.; Repin, A.I.; Nikitenko, M.P.; Konoplev, N.P.; Fil, N.S.

    2000-01-01

    Economic operational indices of WWER-type reactors show their competitiveness in all the countries where these reactors operate. Advanced WWERs being designed and constructed now have the improved characteristics of economical efficiency and are more convenient for operation and maintenance. Many technical solutions aimed at improvement of the operational performance are implemented in the design of WWER-1000/V-392 and WWER-640/V-407, and these reactors are the important basis for the nuclear power expansion in Russia. Some of these solutions are considered in the present paper. (author)

  9. AER working group D on WWER safety analysis - report of the 2007 meeting

    International Nuclear Information System (INIS)

    Kliem, S.

    2007-01-01

    The AER working group D on WWER reactor safety analysis held its sixteenth meeting in Paris, France during the period 08-09 May 2007. The meeting was hosted by the CEA France. It followed the final workshop on the OECD/DOE/CEA WWER-1000 Coolant Transient Benchmark held at 07 May. Altogether 11 participants attend the meeting of the working group D, 7 from AER member organizations and 4 guests from non-member organizations. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting. The meeting started with a general information exchange about the recent activities in the participating organizations. The given presentations and the discussions can be attributed to the following topics: -Code development and benchmarking for reactor dynamics applications; -Safety analysis methodology and results; -Future activities. New solutions for three different benchmarks were presented and discussed. These are the Second AER Dynamic Benchmark on control rod ejection at hot zero power (S. Kliem, FZD), the WWER-1000 Coolant Transient Benchmark (E. Syrjaelahti, VTT) and the stationary AER-FCM101 Benchmark considering a WWER-1000 reactor (C. Parisi, UniPisa). A. Kereszturi (AEKI) presented a statistical evaluation of the possibility to observe a fuel assembly mis loading event. The second presentation of E. Syrjaelahti was dedicated to the description how best-estimate coupled code calculations at VTT are supported by uncertainty and sensitivity analyses. K. Velkov (GRS) presented preliminary results of BIPR8KN/ATHLET calculations with a very detailed resolution of the calculation grid on the assessment of coolant mixing inside WWER-1000 assembly heads. Coolant mixing experiments at three different mixing test facilities, modeling different reactor types, were presented and compared by S. Kliem. A calculation study using the coupled code system KORSAR/GP on the consequences of the injection of a slug of un borated water into the reactor core was

  10. Passive system with steam-water injector for emergency supply of NPP steam generators

    International Nuclear Information System (INIS)

    Il'chenko, A.G.; Strakhov, A.N.; Magnitskij, D.N.

    2009-01-01

    The calculation results of reliability indicators of emergency power supply system and emergency feed-water supply system of serial WWER-1000 unit are presented. To ensure safe water supply to steam generators during station blackout it was suggested using additional passive emergency feed-water system with a steam-water injector working on steam generators dump steam. Calculated analysis of steam-water injector operating capacity was conducted at variable parameters of steam at the entrance to injector, corresponding to various moments of time from the beginning of steam-and-water damping [ru

  11. Accident beyond the design basis management with the coolant loss at the NPP with WWER

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

    2010-01-01

    The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

  12. Surveillance specimen programmes for WWER reactor vessels in the Czech Republic

    International Nuclear Information System (INIS)

    Brynda, J.; Hogel, J.; Brumovsky, M.

    2003-01-01

    The present state of materials degradation in WWER reactor pressure vessels manufactured in the Czech Republic is highlighted. The standard surveillance program for WWER-440/V-213 type reactors is described and its deficiencies together with the main results obtained are discussed. A new supplementary surveillance program meeting all requirements for PWR type reactors has been developed and launched. An entirely new design was chosen for the surveillance programme for WWER-1000/V-320 type reactor pressure vessels. Materials selection, container design and location as well as the withdrawal plan connected with ex-vessel fluence monitoring are described

  13. The BN-1800 advanced sodium cooled fast reactor meeting requirements to nuclear power engineering of the XXI century

    International Nuclear Information System (INIS)

    Poplavskij, V.M.; Tsibulya, A.M.; Kamaev, A.A.

    2004-01-01

    Basic principles and direction of the elaboration of sodium fast reactor BN-1800 are discussed. The elaboration of the BN-1800 reactor is based on the scientific justified technical feasibilities of BN-350, BN-600 and BN-800 reactors. Descriptions of power blocks and reactor core of the elaborated reactor are presented. Characteristics of the BN-1800 steam generator are given. Safety of reactor unit is estimated, fundamental technical and economic indexes of BN-1800 are discussed. Economic indexes of the BN-1800 reactor are noted to be on the level of WWER-1000 and WWER-1500 reactors [ru

  14. 'Kazmer' a complex noise diagnostic system for 1000 MWe PWR WWER type nuclear power units

    International Nuclear Information System (INIS)

    Por, G.

    1992-06-01

    Noise diagnostic systems have previously been developed and installed for the WWER-440 type reactors at the Paks Nuclear Power Plant, Hungary. Based on the experiences, the system has been extended and modified for use in 1000 MWe, WWER-1000 type units. KAZMER consists of three subsystem, the KARD reactor noise diagnostic system, ARGUS vibration monitoring system for rotation machinery, and ALMOS acoustic monitoring system. The installation of the KAZMER system at the Kalinin Nuclear Power Station, Russia, and the first operational experiences are outlined. (R.P.) 15 refs.; 9 figs

  15. Activities in the Czech Republic for reactor pressure components lifetime management

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1994-01-01

    The following activities in the Czech republic for reactor pressure components lifetime management are described: upgrading and safety assurance of nuclear power plants (NPP) with reactors of WWER-440/V-230 type, safety assurance of NPPs with reactors of WWER-440/V-213, lifetime management programme of NPPs with WWER-440/V-213 reactors, preparation of start-up of NPPs with WWER-1000/V-320 reactors, preparation of guides for lifetime as well as defect allowability evaluation in main components of primary and secondary circuits. 3 figs

  16. New requirements for the WWER fuel and their consideration in designing the fuel assemblies

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Ananyev, Y.

    2003-01-01

    In 2001-2002 the base designs of the new generation fuel assemblies for the WWER-440 and WWER-1000 reactors were developed. The ways of their further modernisation were defined. The present report deals with the urgent requirements and how they have been implemented in these designs. The assessment of the efficiency of new designs is carried out on the basis of the existing data of the world market on the cost of: Uranium concentrate; dividing operations; fabrication. It is additionally possible also to take into account the cost of transportation, storage and processing of the irradiated fuel including burial of wastes

  17. Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

  18. The supply model of the IandC system modernization at VVER 1000 units of the NPP Temelin

    International Nuclear Information System (INIS)

    Tomicek, L.

    1997-01-01

    The history of the instrumentation and control system of the Temelin nuclear power plant is briefly mentioned. An upgrading of this system was initiated, aimed at enhancement of nuclear safety and at increasing the economical effectiveness of electricity production. The final supplier of the new system is SKODA PRAHA, the Westinghouse Electric Co. being the sub-supplier. By modernization of the IandC system, conditions have been created for an up-to-date WWER-1000 type unit which copes with the current requirements on nuclear safety and operating economy. (A.K.)

  19. Rod drop in the LR-0 reactor core comprising 55 fuel assemblies

    International Nuclear Information System (INIS)

    Hadek, J.; Grundmann, U.

    1989-09-01

    Data from the third stage of kinetic measurements on the LR-0 reactor, performed in 1988, were employed for additional calculations using the 3-dimensional neutron kinetics code HEXDYN3D. The reactor consists of subassemblies similar to those in the WWER-1000 (PWR) reactor. The theoretical and experimental results are compared for the time behavior of the neutron flux caused by drop of the control rod cluster in various subassemblies of the reactor. The results demonstrate that the HEXDYN3D code is well suited to the treatment of the space-time behavior of the neutron flux. (author). 21 figs., 2 tabs., 16 refs

  20. Comparative analysis of different methods of modelling of most loaded fuel pin in transients

    International Nuclear Information System (INIS)

    Ovdiyenko, Y.; Khalimonchuk, V.; Ieremenko, M.

    2007-01-01

    Different methods of modeling of most loaded fuel pin are presented at the work. Calculation studies are performed on example of accident related to WWER-1000 cluster rod ejection with using of spatial kinetic code DYN3D that uses nodal method to calculate distribution of neutron flux in the core. Three methods of modeling of most loaded fuel pin are considered - flux reconstruction in fuel macrocell, pin-by-pin calculation by using of DYN3D/DERAB package and by introducing of additional 'hot channel'. Obtained results of performed studies could be used for development of calculation kinetic models during preparing of safety analysis report (Authors)

  1. Power generation in the 12-th five-year plan

    International Nuclear Information System (INIS)

    Troitskij, A.A.

    1986-01-01

    The state of electric power generation in the 11-th five-year plan is summed up. Perspectives of development of heat and electric power generation in the 12-th five-year plan are considered. Thermal power generation of NPPs in 1990 will increase by a factor of 8.4 as compared with 1975. The NPP development will be mainly realized on the basis of the WWER-1000 type reactors. It is planned to commission fast reactors of up to 800 MW

  2. Status and prospects of WWER in-core fuel management activities

    Energy Technology Data Exchange (ETDEWEB)

    Novikov, A N; Pavlov, V I; Pavlovichev, A M; Proselkov, V N; Saprykin, V V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    A short review is given of recent extensive calculational and experimental studies carried out in Russia and Bulgaria for WWER fuel cycle modernization. The main activities performed at Kola NPP, Novovoronezh NPP, Kozloduy NPP and Balakovo NPP are outlined. Based on experience gained, the following improvements in the fuel cycle have been introduced: (1) increased fuel burnup; (2) reduced natural uranium consumption and decreased amount of separation work per energy output unit; (3) increased efficiency of the reactor emergency protection; (4) reduced fast neutron flux onto the reactor vessel. The main characteristics of modernized fuel cycles of WWER-440 and WWER-1000 are presented. 4 tabs., 3 figs., 14 refs.

  3. The concept of electro-nuclear facility for useful power generation and minor actinides transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Bergelson, B.R.; Balyuk, S.A. [ITEP, Moscow (Russian Federation)

    1995-10-01

    The possibility is shown to design in principle the double-purpose liquid fuel electro nuclear facility for useful power generation and minor actinides transmutation in U-Pu fuel cycle conditions. D{sub 2}O and a melt of fluorine salts are considered as a working media for liquid fuel. Such facility replenished with depicted or natural uranium only makes it possible to generate power of 900 MW (c) for external consumers and serve 20 WWER-1000 reactors for transmutation of MA. The facility could be thought as an alternative to fast reactors since appr. 30% of the total power confined in uranium is utilized in it.

  4. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Papp, L.

    1995-01-01

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed

  5. Estimation of dynamic loading on a design of the NPP caused by seismic influences

    International Nuclear Information System (INIS)

    Proskuryakov, Konstantin

    2011-01-01

    Methods and algorithms of calculations of quality factor of a stream of the coolant are developed. Quantitative estimations of a range of frequency of vibration - acoustical resonance between the coolant flowing through the reactor core and fuel assembly vibration in the NPP with WWER-1000 are provided. The design procedure of quality factor of a stream of the coolant and a band - width in advanced light water reactor is developed. The experimental substantiation of sharp increase of intensity of vibrations at occurrence of vibration - acoustical resonance is received. The reasons of abnormal growth of level of vibrations are identified at stationary modes of cold - ops, hot - ops of the equipment of reactor installations with WWER-1000. It is showed that for prevention of vibration - acoustical resonance of the coolant and fuel assembly it is necessary and sufficient to deduce own frequency of fuel assembly vibrations from band - width limits. The technique of designing of cartograms of a reactor core with indication of quantity and location of fuel assemblies with high level of vibration is worked out. (author)

  6. Organization and mechanization of maintenance operations at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    Titov, A.A.

    1983-01-01

    The structure of capital investments defining organization and mechanization of maintepance operations at NPPs with the WWER type reactors is analyzed. The trends in development of optimum decisions for organization and mechanization of repair obs at NPPs being designed taking into account the prospects of nuclear powep enginerning development, the system of NPP maintenance servicing, as well as the structure of repair-productive capacities are discussed. On the basis of the analysis of the data obtained in designing the Zaporozhskaya NPP it is shown that the capital investments for organizing and mechanization of maintenance operations at the unified NPP site with four WWER-1000 reactors reach nearly 18 roubles/kW. A conclusion is drawn that at present the design of an NPP with the WWER-1000 reactor totally meets the requirements of realization of periodic maintenance operations. It is advisable to cooperate the NPP management with that of a thermal power station from the viewpoint of using manpower, which would improve the operating conditions and labour productivity of workers engaged in repair and, consequently, reduce the capital investments and repair expenditures

  7. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 5A. Experience data. Working material. Experience database of Romanian facilities subjected to the last three Vrancea earthquakes. Final report from November 1994 - October 1995

    International Nuclear Information System (INIS)

    Coman, O.; Stevenson, J.D.

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains the report experience database for Romanian facilities contingent to the three Vrancea earthquakes

  8. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4A. Paks NPP: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to seismic analyses of structures of Paks and Kozloduy reactor buildings and WWER-440/213 primary coolant loops with different antiseismic devices

  9. Westinghouse experience in Kozloduy NPP units 5 and 6 I and C modernization

    International Nuclear Information System (INIS)

    Sechensky, B.

    2005-01-01

    The paper presents the background, current implementation approach and experience on the largest ever modernization program on operating units WWER 1000 (PWR) at Kozloduy Nuclear Power Plant in Bulgaria. The Modernization Program itself includes more than 212 measures. Westinghouse is modernizing the major I and C systems at WWER 1000. The major topics of the modernization program and specific approach described in this paper are as follows: 1) Design Approach and Feature; 2) Installation Approach; 3) Test Strategy; 4) Licensing Strategy, applicable codes and standards. At the end author summarized that: 1) Specific design solutions were required and developed in order to address the specific plant features. At each stage, representatives of the Client are being involved in the process of designing and testing of the equipment and systems; 2) Phase-by-phase installation efforts were developed and extensive installation design documentation was prepared to fit in the limited outage window and to successfully complete the installation activities; 3) Well-prepared, multi-phase testing strategy was developed and is being implemented to assure the proper and adequate operation of the equipment at the factory and at the real plant

  10. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4B. Paks NPP: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on dynamic study of the main building of the Paks NPP; shake table investigation at Paks NPP and the Final report of the Co-ordinated Research Programme

  11. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3E. Kozloduy NPP units 5/6: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to floor response spectra of Kozloduy NPP; calculational-experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating WWER-type NPPs; analysis of design floor response spectra and testing of the electrical systems; experimental investigations and seismic analysis Kozloduy NPP; testing of components on the shaking table facilities and contribution to full scale dynamic testing of Kozloduy NPP; seismic evaluation of the main steam line, piping systems, containment pre-stressing and steel ventilation chimney of Kozloduy NPP

  12. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 5B. Experience data. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on the effects of Armenia earthquakes on selected power, industry and commercial facilities and seismic functional qualification of active mechanical and electrical components tested on shaking table

  13. The international WWER fuel market

    International Nuclear Information System (INIS)

    Gingold, G.E.; Goldstein, L.; Strasser, A.A.

    1994-01-01

    The state of the world nuclear fuel market and its economic complexities are described. Currently the nuclear fuel market is oversupplied and nuclear fuel fabrication in the West far exceeds the anticipated demands. Actually the current demand is not much more than half of the capacity available to supply it. The Eastern Europe (excluding the plants in the Russian Federation) with its 20 WWER-440 and 12 WWER-1000 reactors in operation and additional 4 WWER-440 and 8 WWER-1000 units under construction is considered as a potential long-term market for the Western fuel fabricators. The following significant benefits of competition in the WWER fuel market for the operators of these reactors are : 1) lower cost; 2) more favorable contract terms and improved vendor cooperation with the customer; 3) accelerated technological development. A brief description of the main WWER fuel suppliers TVEL, ABB Atom, BNFL, EVF and Westinghouse, as well as the status of some new companies as CEZ and SEP is given. The principal differences between Western and WWER fuels are outlined. The advanced features offered by the Western vendors and Russian fuel supply organisations are discussed. 2 tabs., 1 fig

  14. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 2. Generic material: Codes, standards, criteria. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports related to generic material, namely codes, standards and criteria for benchmark analysis.

  15. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4A. Paks NPP: Analysis/testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to seismic analyses of structures of Paks and Kozloduy reactor buildings and WWER-440/213 primary coolant loops with different antiseismic devices.

  16. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP.

  17. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4C. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material involves comparative analysis of the seismic analysis results of the reactor building for soft soil conditions, derivation of design response spectra for components and systems; and upper range design response spectra for soft soil site conditions at Paks NPP

  18. Cooperation of Atomehnergoehksport and Skodaexport in the sphere of atomic power engineering

    International Nuclear Information System (INIS)

    Gulko, V.D.

    1986-01-01

    The cooperation is briefly outlined of the Czechoslovak foreign trade organization Skodaexport and its Soviet counterpart Atomehnergoehksport in nuclear power engineering. Thanks to this cooperation within socialist economic integration, 12 WWER-440 units of a total capacity of 5,280 MWe will be built in Czechoslovakia by 1990 followed by the construction of WWER-1000 reactors. Thus, by the year 2000, the total installed capacity in Czechoslovakia will be 15 to 16 thousand MWe. The reactors are being built according to Soviet project designs, with active participation of Soviet experts and organizations. Within the project of cooperation, the Soviet Union also supplies Czechoslovakia with nuclear fuel and regularly provides spent fuel disposal. On the other hand, Czechoslovakia supplies the Soviet Union and other CMEA member countries with reactor components, especially WWER-440 primary circuit equipment, and will become the main supplier of WWER-1000 primary circuit equipment, both according to Soviet project designs. This and close cooperation of Czechoslovak and Soviet organizations ranks Czechoslovakia among the major suppliers of technological equipment for nuclear power plants in CMEA countries. (L.O.)

  19. A novel optimization method, Gravitational Search Algorithm (GSA), for PWR core optimization

    International Nuclear Information System (INIS)

    Mahmoudi, S.M.; Aghaie, M.; Bahonar, M.; Poursalehi, N.

    2016-01-01

    Highlights: • The Gravitational Search Algorithm (GSA) is introduced. • The advantage of GSA is verified in Shekel’s Foxholes. • Reload optimizing in WWER-1000 and WWER-440 cases are performed. • Maximizing K eff , minimizing PPFs and flattening power density is considered. - Abstract: In-core fuel management optimization (ICFMO) is one of the most challenging concepts of nuclear engineering. In recent decades several meta-heuristic algorithms or computational intelligence methods have been expanded to optimize reactor core loading pattern. This paper presents a new method of using Gravitational Search Algorithm (GSA) for in-core fuel management optimization. The GSA is constructed based on the law of gravity and the notion of mass interactions. It uses the theory of Newtonian physics and searcher agents are the collection of masses. In this work, at the first step, GSA method is compared with other meta-heuristic algorithms on Shekel’s Foxholes problem. In the second step for finding the best core, the GSA algorithm has been performed for three PWR test cases including WWER-1000 and WWER-440 reactors. In these cases, Multi objective optimizations with the following goals are considered, increment of multiplication factor (K eff ), decrement of power peaking factor (PPF) and power density flattening. It is notable that for neutronic calculation, PARCS (Purdue Advanced Reactor Core Simulator) code is used. The results demonstrate that GSA algorithm have promising performance and could be proposed for other optimization problems of nuclear engineering field.

  20. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 1. Data related to sites and plants: Paks NPP, Kozloduy NPP. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to sites and NPPs Paks and Kozloduy

  1. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 1. Data related to sites and plants: Paks NPP, Kozloduy NPP. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on data related to sites and NPPs Paks and Kozloduy.

  2. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

    1996-01-01

    The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

  3. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 5A. Experience data. Working material. Experience database of Romanian facilities subjected to the last three Vrancea earthquakes. Final report from November 1994 - October 1995

    Energy Technology Data Exchange (ETDEWEB)

    Coman, O [Stevenson and Associates Bucharest Office, Bucharest (Romania); Stevenson, J D [Stevenson and Associates (United States)

    1996-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains the report experience database for Romanian facilities contingent to the three Vrancea earthquakes.

  4. Working group on VVER safety analysis - report of the 2010 meeting

    International Nuclear Information System (INIS)

    Kliem, S.

    2010-01-01

    The AER Working Group D on WWER reactor safety analysis held its nineteenth meeting in Pisa, Italy, during the period 15-16 April, 2010. The meeting was hosted by the San Piero a Grado Nuclear Research Group of the University of Pisa and was held in conjunction with the second workshop on the OECD/NEA Benchmark for the Kalinin-3 WWER-1000 NPP and the fourth workshop on the OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of LWRs. Altogether 12 participants attended the meeting of the working group D, 8 from AER member organizations and 4 guests from non-member organization. The co-ordinator of the working group, Mr. S. Kliem, served as chairman of the meeting. The meeting started with a general information exchange about the recent activities in the participating organizations. The given presentations and the discussions can be attributed to the following topics:-Code validation and benchmarking including the calculation of the OECD/NEA Benchmark for the Kalinin-3 WWER-1000 NPP;-Safety analyses and code developments;-Future activities A list of the participants and a list of the handouts distributed at the meeting are attached to the report. The corresponding PDF-files can be obtained from the chairman. (Author)

  5. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 2. Generic material: Codes, standards, criteria. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports related to generic material, namely codes, standards and criteria for benchmark analysis

  6. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 4D. Paks NPP: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. The two prototypes are Kozloduy Units 5/6 for WWER-1000 and Paks for WWER-440/213 nuclear power plants. This volume of Working material contains reports on seismic margin assessment and earthquake experience based methods for WWER-440/213 type NPPs; structural analysis and site inspection for site requalification; structural response of Paks NPP reactor building; analysis and testing of model worm type tanks on shaking table; vibration test of a worm tank model; evaluation of potential hazard for operating WWER control rods under seismic excitation

  7. METMET fuel with Zirconium matrix alloys

    International Nuclear Information System (INIS)

    Savchenko, A.; Konovalov, I.; Totev, T.

    2008-01-01

    The novel type of WWER-1000 fuel has been designed at A.A. Bochvar Institute. Instead of WWER-1000 UO 2 pelletized fuel rod we apply dispersion type fuel element with uniformly distributed high uranium content granules of U9Mo, U5Nb5Zr, U3Si alloys metallurgically bonded between themselves and to cladding by a specially developed Zr-base matrix alloy. The fuel meat retains a controllable porosity to accommodate fuel swelling. The optimal volume ratios between the components are: 64% fuel, 18% matrix, 18% pores. Properties of novel materials as well as fuel compositions on their base have been investigated. Method of fuel elements fabrication by capillary impregnation has been developed. The primary advantages of novel fuel are high uranium content (more than 15% in comparison with the standard UO 2 pelletized fuel rod), low temperature of fuel ( * d/tU) and serviceability under transient conditions. The use of the novel fuel might lead to natural uranium saving and reduced amounts of spent fuel as well as to optimization of Nuclear Plant operation conditions and improvements of their operation reliability and safety. As a result the economic efficiency shall increase and the cost of electric power shall decrease. (authors)

  8. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  9. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  10. Criticality calculations of various spent fuel casks - possibilities for burn up credit implementation

    International Nuclear Information System (INIS)

    Apostolov, T; Manolova, M.; Prodanova, R.

    2001-01-01

    A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion K eff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)

  11. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  12. The international WWER fuel market

    Energy Technology Data Exchange (ETDEWEB)

    Gingold, G E; Goldstein, L; Strasser, A A [Stoller (S.M.) Corp., Pleasantville, NY (United States)

    1994-12-31

    The state of the world nuclear fuel market and its economic complexities are described. Currently the nuclear fuel market is oversupplied and nuclear fuel fabrication in the West far exceeds the anticipated demands. Actually the current demand is not much more than half of the capacity available to supply it. The Eastern Europe (excluding the plants in the Russian Federation) with its 20 WWER-440 and 12 WWER-1000 reactors in operation and additional 4 WWER-440 and 8 WWER-1000 units under construction is considered as a potential long-term market for the Western fuel fabricators. The following significant benefits of competition in the WWER fuel market for the operators of these reactors are : (1) lower cost; (2) more favorable contract terms and improved vendor cooperation with the customer; (3) accelerated technological development. A brief description of the main WWER fuel suppliers TVEL, ABB Atom, BNFL, EVF and Westinghouse, as well as the status of some new companies as CEZ and SEP is given. The principal differences between Western and WWER fuels are outlined. The advanced features offered by the Western vendors and Russian fuel supply organisations are discussed. 2 tabs., 1 fig.

  13. The relevance of the IFPE Database to the modelling of WWER-type fuel behaviour

    International Nuclear Information System (INIS)

    Killeen, J.; Sartori, E.

    2006-01-01

    The aim of the International Fuel Performance Experimental Database (IFPE Database) is to provide, in the public domain, a comprehensive and well-qualified database on zircaloy-clad UO 2 fuel for model development and code validation. The data encompass both normal and off-normal operation and include prototypic commercial irradiations as well as experiments performed in Material Testing Reactors. To date, the Database contains over 800 individual cases, providing data on fuel centreline temperatures, dimensional changes and FGR either from in-pile pressure measurements or PIE techniques, including puncturing, Electron Probe Micro Analysis (EPMA) and X-ray Fluorescence (XRF) measurements. This work in assembling and disseminating the Database is carried out in close co-operation and co-ordination between OECD/NEA and the IAEA. The majority of data sets are dedicated to fuel behaviour under LWR irradiation, and every effort has been made to obtain data representative of BWR, PWR and WWER conditions. In each case, the data set contains information on the pre-characterisation of the fuel, cladding and fuel rod geometry, the irradiation history presented in as much detail as the source documents allow, and finally any in-pile or PIE measurements that were made. The purpose of this paper is to highlight data that are relevant specifically to WWER application. To this end, the NEA and IAEA have been successful in obtaining appropriate data for both WWER-440 and WWER-1000-type reactors. These are: 1) Twelve (12) rods from the Finnish-Russian co-operative SOFIT programme; 2) Kola-3 WWER-440 irradiation; 3) MIR ramp tests on Kola-3 rods; 4) Zaporozskaya WWER-1000 irradiation; 5) Novovoronezh WWER-1000 irradiation. Before reviewing these data sets and their usefulness, the paper touches briefly on recent, more novel additions to the Database and on progress made in the use of the Database for the current IAEA FUMEX II Project. Finally, the paper describes the Computer

  14. Seismic Probabilistic Risk Assessment (SPRA), approach and results

    International Nuclear Information System (INIS)

    Campbell, R.D.

    1995-01-01

    During the past 15 years there have been over 30 Seismic Probabilistic Risk Assessments (SPRAs) and Seismic Probabilistic Safety Assessments (SPSAs) conducted of Western Nuclear Power Plants, principally of US design. In this paper PRA and PSA are used interchangeably as the overall process is essentially the same. Some similar assessments have been done for reactors in Taiwan, Korea, Japan, Switzerland and Slovenia. These plants were also principally US supplied or built under US license. Since the restructuring of the governments in former Soviet Bloc countries, there has been grave concern regarding the safety of the reactors in these countries. To date there has been considerable activity in conducting partial seismic upgrades but the overall quantification of risk has not been pursued to the depth that it has in Western countries. This paper summarizes the methodology for Seismic PRA/PSA and compares results of two partially completed and two completed PRAs of soviet designed reactors to results from earlier PRAs on US Reactors. A WWER 440 and a WWER 1000 located in low seismic activity regions have completed PRAs and results show the seismic risk to be very low for both designs. For more active regions, partially completed PRAs of a WWER 440 and WWER 1000 located at the same site show the WWER 440 to have much greater seismic risk than the WWER 1000 plant. The seismic risk from the 1000 MW plant compares with the high end of seismic risk for earlier seismic PRAs in the US. Just as for most US plants, the seismic risk appears to be less than the risk from internal events if risk is measured is terms of mean core damage frequency. However, due to the lack of containment for the earlier WWER 440s, the risk to the public may be significantly greater due to the more probable scenario of an early release. The studies reported have not taken the accident sequences beyond the stage of core damage hence the public heath risk ratios are speculative. (author)

  15. Twenty years nuclear power sector in Bulgaria - an attempt at striking a balance

    Energy Technology Data Exchange (ETDEWEB)

    Shervashidze, N

    1994-12-31

    The advantages of nuclear power and the mistakes that cause the opposition against it on the example of the history of the Bulgarian nuclear power sector are considered in a polemical manner. Four WWER 440/230 and two WWER 1000 units have been put into operation in Kozloduj NPP in the period 1974 - 1993. Before 1991, however, safety operation criteria were compromised numerous times and the atmosphere of secrecy, self-complacency and the series of absurdities about the Chernobyl accident were clearly used by the opponents of nuclear energy. Now the attitude towards nuclear power is changing for the better. Much progress has been made in recovering the nuclear power sector under complicated economic and political conditions. The role of NPP as an anti-inflation factor, an element providing the national security, a stimulator of industry and infrastructure, and an ecological salvation factor, is stressed. (I.M.).

  16. Post-test examination and evaluation of CORA experiments; Nachuntersuchung und Auswertung der CORA-Experimente

    Energy Technology Data Exchange (ETDEWEB)

    Leistikow, S.; Schanz, G.; Metzger, H.

    1995-08-01

    The experimental program has been completed with the tests CORA-W1 and CORA-W2, bundles without and with absorber material, prepared in cooperation with the Kurchatov Institute in order to study WWER-1000 typical core materials and fuel element configuration. Reported are results of post-test microstructural investigations of CORA-W2 towards the complex interdependence of the Zr-1 % Nb cladding oxidation, the temperature escalation, the destruction of the absorber rod, initiating the melting and the chemical interaction of components, the melt relocation and blockage formation. The results of this contribution are qualitatively comparable to the behavior of western LWR type fuels under SFD-conditions. (orig./HP)

  17. Researches of WWER fuel rods behaviour under RIA accident conditions

    International Nuclear Information System (INIS)

    Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

    2003-01-01

    Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

  18. 'AER Working Group D On WWER Safety Analysis' Report Of The Meeting In Pisa, Italy, 26-27 April 2006

    International Nuclear Information System (INIS)

    Siltanen, P.

    2006-01-01

    AER Working Group D on WWER reactor safety analysis held its 15 t h meeting in Grand Hotel Duomo in Pisa, Italy during the period 26-27 April 2006. The meeting was hosted by the University of Pisa following the fourth workshop on the OECD/DOE/CEA WWER-1000 Coolant Transient Benchmark (W1000-CT) held at the same location on 24-25 April. Altogether 15 participants attended the Working Group D meeting, 11 from AER member organizations and 4 guests from non-member organizations. The coordinator for the working group, Mr. P. Siltanen (FNS) served as chairman. In addition to general information exchange on recent activities in the participating organizations, the topics of the meeting included: 1) Code development and benchmarking for reactor dynamics applications. 2) Safety analysis methodology and results. 3) Future activities. A list of participants and a list of handouts distributed at the meeting are attached to the report (Author)

  19. Quality assurance monitoring during nuclear fuel production in JSC 'TVEL'

    International Nuclear Information System (INIS)

    Filimonov, G.; Tchirkov, V.

    2000-01-01

    The paper describes Quality Assurance (QA) monitoring during fabrication of nuclear fuel in Russian Federation. Joint Stock Company 'TVEL', natural state monopoly of the type of holding that fabricates and supplies nuclear fuel for the NPPs of Russia, CIS and Europe, incorporates the major enterprises of the nuclear fuel cycle including JSC 'Mashinostroitelny zavod', Electrostal (fabrication of fuel pellets, rods and assemblies for different types of reactors), JSC 'Novosibirsky zavod khimconcentratov', Novosibirsk (fabrication of fuel rods and assemblies for WWER-440 and WWER-1000), JSC 'Tchepetsky mechanitchesky zavod', Tchepetsk (fabrication of Zr tubing). Monitoring of QA is an important element of Quality Management System (QMS) developed and implemented at the above-mentioned enterprises of the JSC 'TVEL' and it is performed on three levels including external and internal audits and author's supervision. Paper also describes short- and long-term policies of the JSC 'TVEL' in nuclear fuel quality field. (author)

  20. Experience with the WIMS computer code at Skoda Plzen

    International Nuclear Information System (INIS)

    Vacek, J.; Mikolas, P.

    1991-01-01

    Validation of the program for neutronics analysis is described. Computational results are compared with results of experiments on critical assemblies and with results of other codes for different types of lattices. Included are the results for lattices containing Gd as burnable absorber. With minor exceptions, the results of benchmarking were quite satisfactory and justified the inclusion of WIMS in the production system of codes for WWER analysis. The first practical application was the adjustment of the WWER-440 few-group diffusion constants library of the three-dimensional diffusion code MOBY-DICK, which led to a remarkable improvement of results for operational states. Then a new library for the analysis of WWER-440 start-up was generated and tested and at present a new library for the analysis of WWER-440 operational states is being tested. Preparation of the library for WWER-1000 is in progress. (author). 19 refs

  1. Structural steels for power generating equipment and heat and chemical heat treatments

    International Nuclear Information System (INIS)

    Astaf'ev, A.A.

    1979-01-01

    Development of structural steels for power generating equipment and for reactor engineering, in particular, is elucidated. Noted is utilization of the 15Kh2NMFA steels for the WWER-1000 reactor vessels, the 10GN2MFA steels for steam generators, pressurizers, vessels of the automatic emergency shut down and safety system; the 00Kh12N3DL steel for cast pump vessels and main locking bars. The recommendations on heat treatment of big forgings, for instance, ensuring the necessary complex of mechanical properties are given. Diffusion chromizing with subsequent nitriding of austenitic steels which increase durability of the components in BN reactors more than 4 times, is practised on a large scale

  2. WWER radial reflector modeling by diffusion codes

    International Nuclear Information System (INIS)

    Petkov, P. T.; Mittag, S.

    2005-01-01

    The two commonly used approaches to describe the WWER radial reflectors in diffusion codes, by albedo on the core-reflector boundary and by a ring of diffusive assembly size nodes, are discussed. The advantages and disadvantages of the first approach are presented first, then the Koebke's equivalence theory is outlined and its implementation for the WWER radial reflectors is discussed. Results for the WWER-1000 reactor are presented. Then the boundary conditions on the outer reflector boundary are discussed. The possibility to divide the library into fuel assembly and reflector parts and to generate each library by a separate code package is discussed. Finally, the homogenization errors for rodded assemblies are presented and discussed (Author)

  3. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  4. Forecasting of resonances vibration equipment with elastic waves coolant and with the external periodic loads on NPP with WWER

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Zaporozhets, M.V.; Fedorov, A.I.

    2015-01-01

    Forecasting are carried out for external loads in relation to the main circulation circuit - dynamic loads caused by the rotation of the MCP, dynamic loads caused by the earthquake, dynamic loads caused by damage to the MCP in the earthquake. A comparison of the response spectrum of one of the variants of the base of the NPP, with the frequency vibration of the primary circuit equipment for NPP with WWER-1000 and self-frequency of elastic waves in the fluid. Analysis of the comparison results shows that the frequency of vibration of the main equipment of the reactor plant and elastic waves are in the frequency band in the spectrum response corresponding to the maximum amplitude of the seismic action [ru

  5. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.

    2007-01-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes

  6. Operating and mathematical representation of resonances between flow parameters oscillations and structure vibrations of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Yang Shan Afshar, E.; Polyakov, N.I. [Nuclear Power Plant Department of Moscow Power Engineering Institute Technical Univ., Moscow (Russian Federation)

    2007-07-01

    The experimental data that have been obtained from the measurements of noise signals in primary circuit of NPP with reactor of WWER-1000 are presented. The causes of resonant interaction between Eigen-Frequencies of Oscillations of the Coolant Pressure (EFOCP) and structure vibrations are discussed. An application-oriented approach to the problem of identification of abnormal phenomena of thermal-hydraulic parameters is proposed. Logarithmic Decrement {delta} is determined. The bigger damping ratio {zeta} provides bigger {delta} and correspondingly smaller values of Q-factor and amplitude X(t)max. All experimental units intended for NPP severe accident investigation must satisfy to the NPP Q-factor criterion of similarity. (authors)

  7. Operating and mathematical representation of resonances between flow parameters oscillations and structure vibrations of NPP

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Yang Shan Afshar, E.; Polyakov, N.I.

    2007-01-01

    The experimental data that have been obtained from the measurements of noise signals in primary circuit of NPP with reactor of WWER-1000 are presented. The causes of resonant interaction between Eigen-Frequencies of Oscillations of the Coolant Pressure (EFOCP) and structure vibrations are discussed. An application-oriented approach to the problem of identification of abnormal phenomena of thermal-hydraulic parameters is proposed. Logarithmic Decrement δ is determined. The bigger damping ratio ζ provides bigger δ and correspondingly smaller values of Q-factor and amplitude X(t)max. All experimental units intended for NPP severe accident investigation must satisfy to the NPP Q-factor criterion of similarity. (authors)

  8. Protection of WWER type primary loops against extreme effects

    International Nuclear Information System (INIS)

    Podrouzek, J.; Rejent, B.

    1985-01-01

    Dynamic analyses of the WWER-440 primary loops for the Mochovce nuclear power plant showed that the unprotected primary loop is very soft with a first eigenfrequency of 0.38 Hz. Protection with amortisseurs and viscous shock absorbers was compared and the viscous shock absorber in all cases proved to be more suitable. GERB viscous absorbers will be installed at the Mochovce nuclear power plant. First calculations of the dynamic resistance of the WWER-1000 primary loops for the Temelin nuclear power plant to extreme events were also made. It was shown that the unprotected primary loop is rather soft with a first eigenfrequency of 0.9 Hz, or 0.6 Hz at the pressurizer branch. It will therefore be necessary to protect the primary loops with viscous shock absorbers. (Z.M.)

  9. CASTOR {sup registered} 1000/19. Transport and storage cask for the disposal of spent fuel from the nuclear power plant Temelin (Czech Republic); CASTOR {sup registered} 1000/19. Transport- und Lagerbehaelter zur Entsorgung abgebrannter Brennelemente aus dem Kernkraftwerk Temelin in Tschechien

    Energy Technology Data Exchange (ETDEWEB)

    Fopp, Stefan; Kuehne, Bernhard; Schroeder, Jens [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2011-07-01

    The transport and storage cask CASTOR {sup registered} 1000/19 was designed for a dry interim storage of 19 spent fuel elements of WWER-1000 reactors. The project performed by GNS mbH included design, manufacture, licensing and delivery of 36 casks. The specific requirements for NPP Temelin concern loading and dispatch of the casks to be performed during the outage period of the reactor, thus high reliability and functionality of the casks and the manipulation equipment. The paper describes the mechanical design of the cask, stress analyses for a hypothetical fall accident from 9 m height, performed using the FEM program ANSYS and LS-DYNA. The 3D simulation models are based on conservative material characteristics and upper-bound boundary conditions. The safety analysis was performed using qualified software programs validated by the Czech authorities.

  10. Results of the scientific and technical activities of the Nuclear Reactors and Thermal Physics Institute for 2014. Scientific and technical collection

    International Nuclear Information System (INIS)

    Trufanov, A.A.; Sorokin, A.P.; Vereshchagina, T.N.

    2015-01-01

    In the collection there are the main results of research and development obtained by the researchers of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in 2014, the problems and questions of further investigations are formulated and discussed. Considerable body of data on neutronic, thermohydraulic and technological studies carried out in the frameworks of Proryv project are presented, calculational and experimental justification of design choices and safety of projects on RU BN-1200, multipurpose research reactor MBIR with sodium coolant, RU BREST-OD-300 with lead coolant are among them. The results of experimental and calculational thermophysical investigations in justification of operation conditions and safety of nuclear power plants with water-cooled reactors (WWER-1000, WWER-TOI), pilot studies on innovation project WWER-SKD with supercritical water, in justification of thermonuclear reactor blanket are given [ru

  11. Main research results in the field of nuclear power engineering of the Nuclear Reactors and Thermal Physics Institute in 2014

    International Nuclear Information System (INIS)

    Trufanov, A.A.; Orlov, Yu.I.; Sorokin, A.P.; Chernonog, V.L.

    2015-01-01

    The main results of scientific and technological activities for last years of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in solving problems of nuclear power engineering are presented. The work have been carried out on the following problems: justification of research and development solutions and safety of NPPs with fast reactors of new generation with sodium (BN-1200, MBIR) and lead (BREST-OD-300) coolants, justification of safety of operating and advanced NPPs with WWER reactor facilities (WWER-1000, AEhS 2006, WWER-TOI), development and benchmarking of computational codes, research and development support of Beloyarsk-3 (BN-600) and Bilibino (BN-800) NPPs operation, decommissioning of AM and BR-10 research reactors, pilot scientific studies (WWER-SKD, ITER), international scientific and technical cooperation. Problems for further investigations are charted [ru

  12. Safety provisions of nuclear power plants

    International Nuclear Information System (INIS)

    Niehaus, F.

    1994-01-01

    Safety of nuclear power plants is determined by a deterministic approach complemented by probabilistic considerations. Much use has been made of the wealth of information from more than 6000 years of reactor operation. Design, construction and operation is governed by national and international safety standards and practices. The IAEA has prepared a set of Nuclear Safety Standards as recommendations to its Member States, covering the areas of siting, design, operations, quality assurance, and governmental organisations. In 1988 the IAEA published a report by the International Nuclear Safety Advisory Group on Basic Safety Principles for Nuclear Power Plants, summarizing the underlying objectives and principles of excellence in nuclear safety and the way in which its aspects are interrelated. The paper will summarize some of the key safety principles and provisions, and results and uses of Probabilistic Safety Assessments. Some comments will be made on the safety of WWER 440/230 and WWER-1000 reactors which are operated on Bulgaria. 8 figs

  13. Report of a consultants meeting on accidents during shutdown conditions for WWER nuclear power plants. Extrabudgetary programme on the safety of WWER NPPs

    International Nuclear Information System (INIS)

    1996-07-01

    The main objectives of the meeting were to exchange information on the operational occurrences, studies performed and countermeasures taken for the accidents during shutdown for WWERs, and to define the necessity and directions of the further activities which may promote the improvement of WWER safety under shutdown conditions. The consultants have discussed some aspects concerning vulnerability of safety functions during shutdown conditions, several steps required to performed accident analysis and selected operational aspects for shutdown conditions. The discussion was supported by an evaluation of selected operational occurrences. The consultants have agreed that the discussion during the meeting in major parts is relevant to all the WWER designs (i.e. WWER-1000, WWER-440/213 and WWER-440/230). As for the plant conditions, the consultants have agreed to bound the discussion mainly by the cold shutdown and refuelling modes. Refs, figs, tabs

  14. Construction of foundation slab of Temelin reactor building

    International Nuclear Information System (INIS)

    Lebr, P.; Vyleta, M.

    1988-01-01

    The concreting is described of the foundation slab under the WWER-1000 reactor in the Temelin nuclear power plant. The slab area is 68x68 m and thickness 2.4 m. For ease of concreting, the slab was divided in 12 blocks with vertical partition walls of steel mesh. The total thickness was concreted in three stages in which the partial thicknesses slightly differed for operating reasons. The first two partial thicknesses were concreted in layers of 0.45 m each, the third thickness consisted of two layers of 0.30 m each. The reinforcement was completely cleaned of the concrete residues from the previous stages in the break between the second and the third stages. Totally, 11,050 m 3 concrete were used. Briefly described is quality control during concreting and experiences and recommendations are summed up for other concreting jobs. (Z.M.). 19 figs

  15. Atom for peace in the socialist countries

    International Nuclear Information System (INIS)

    Panasenkov, A.F.

    1979-01-01

    In general terms presented are the achievements of the Council of Mutual Economic Assistance (CMEA) countries in the sphere of nuclear science and technology. The activities of the Joint Institute of Nuclear Research in Dubna and the CMEA Commission on peaceful Use of Nuclear power are described. Works on the development and construction of plants with the WWER-1000 reactor are under completion, the BN-1600 fast reactor technical project is under way. The CMEA countries energy program and their cooperation in this field are presented in short. The directions of studies on the radioactive waste reprocessing and desposal are described. The work of International Economic Union (the ''Interinstrument''), created with the aim of specialization and cooperation on the CMEA countries in the field of nuclear instrument-industry is presented

  16. Recent developments in the external hazard risk assessment in Ukraine

    International Nuclear Information System (INIS)

    2000-01-01

    Ukrainian legislation prescribes safety analysis reports for all operating and future NPPs. Apart from main report they must include: safety analysis supplement; design basis accident analysis; beyond design basis accident analysis; probabilistic safety assessment (PSA); technical; substantiation of safety. Regulatory requirements to PSA contents cover the criteria for core damage frequency and large radioactive release frequency. Initiating events taken into account are internal events; internal hazards and external hazards. External hazards to be considered are seismic events, external fires, external floods, extreme ambient temperatures, aircraft crashes, etc. Current status of PSA development is related to operating WWER-440 and WWER-1000 NPPs and NPPs under construction. This presentation describes in detail the external hazard risk assessment for South Ukraine including methodology applied and expected future activities

  17. Hygiene problems in building a nuclear power and heat plant near Bratislava

    International Nuclear Information System (INIS)

    Chorvat, D.; Mizov, J.; Hladky, E.; Kubik, I.; Carach, J.

    1976-01-01

    The results are presented of the calculation of the population exposure due to the release of radioactive products from a nuclear power and heating plant accident into the ambient atmosphere (primary coolant circuit rupture) providing the nuclear power and heating plant is sited approximately 500 m from the Slovnaft chemical works in Bratislava. Ground water contamination was analyzed assuming the infiltration of radioactive products from a surface deposit due to fallout and the direct infiltration of the products into the soil in the area of the plant. The results of the assessment of the design basis accident of a WWER-1000 nuclear power and heating plant show unequivocally that the emergency core cooling system, full-pressure containment and the correct function of the containment spray system are able to keep the accident consequences within acceptable limits thus meeting radiation hygiene requirements related to the siting of similar installations in the vicinity of large housing estates. (Oy)

  18. Improving the safety of Ukrainian NPP to reach an internationally accepted level

    International Nuclear Information System (INIS)

    Bozhko, S.; Helske, J.; Janke, R.; Mayoral, C.

    2013-01-01

    This paper summarizes the safety status and the modernization progress of Ukrainian NPPs towards an internationally accepted level of safety. After a brief discussion of the concept of what is called an 'international accepted level' for new and operating NPPs, the status of Russian type WWER and in particular the Ukrainian NPPs is presented. Then, the performed investigations of the gaps between international accepted level and the original status of Ukrainian NPPs are presented. The safety objectives of the modernization programs, some examples of defence in depth improvements, and an overall view of the modernization programs of Ukrainian NPPs are produced. Then, few important safety improvements implemented at the oldest Ukrainian WWER-1000 South Ukraine-1 are given in more detail. Finally, a conclusion presents the current status on the way to fulfill the national safety targets and to reach an internationally accepted level for all the Ukrainian NPPs. The paper is followed by the slides of the presentation. (authors)

  19. Perturbation measurements in reactor LR-0 and their evaluation

    International Nuclear Information System (INIS)

    Rypar, W.; Faehrmann, K.H.

    1988-07-01

    To investigate space-dependent kinetic effects in reactors of the WWER-1000 type, two central and one eccentric perturbation measurements were performed in the zero power reactor LR-0 of the UJV Rez (CSSR) by trapeze-form movements of an absorber cluster. The measurements were based ona computer aided CAMAC system for the simultaneous data acquisition of 20 spatially distributed neutron detectors and for cluster movement control. The measurements were followed by a detailed evaluation in the ZfK Rossendorf (GDR) with respect to the calculation results of flux response obtained by nodal code HEXDYN3D, the aim of which was to demostrate the limits of the point reactor model and to account for space-dependent effects by approximative methods. A sensitive check of the calculation methods was made possible especially by the eccentric perturbation where the space dependent effects,due to a larger distance of cluster movement, were most significant. (author). 17 figs., 9 refs

  20. Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

    International Nuclear Information System (INIS)

    Boyadzhiev, Z.; Vapirev, E.

    1995-01-01

    A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs

  1. Estimation of the nuclear fuel assembly eigenfrequencies in the probability sense

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2014-12-01

    Full Text Available The paper deals with upper and lower limits estimation of the nuclear fuel assembly eigenfrequencies, whose design and operation parameters are random variables. Each parameter is defined by its mean value and standard deviation or by a range of values. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of fuel assembly eigenfrequencies in the probability sense. Presented analytical approach used for the calculation of eigenfrequencies sensitivity is based on the modal synthesis method and the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and load-bearing skeleton linked by spacer grids. The method is applied for the Russian TVSA-T fuel assembly in the WWER1000/320 type reactor core in the Czech nuclear power plant Temelín.

  2. Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model

    Directory of Open Access Journals (Sweden)

    Reza Akbari

    2017-08-01

    Full Text Available Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the distribution of heat sources, which is needed to perform the thermal-hydraulic analysis. In this study single heated channel model as a very fast model for predicting thermal hydraulics behavior of pressurized water reactor core has been developed. For verifying the results of this model, we used RELAP5 code as US nuclear regulatory approved thermal hydraulics code. The results of developed single heated channel model have been checked with RELAP5 results for WWER-1000. This comparison shows the capability of single heated channel model for predicting thermal hydraulics behavior of reactor core.

  3. Results of post-irradiation examination of WWER fuel assembly structural components made of E110 and E635 alloys

    International Nuclear Information System (INIS)

    Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Ivashchenko, A.; Strozhuk, A.

    2006-01-01

    The paper presents the main examination results on the condition of fuel rods claddings, guide tubes and spacer grids of the WWER FA made of E110 and E635 alloys operated under standard operating conditions. The paper is based on the data obtained during the examination of 28 WWER-1000 FA and 12 WWER-400 FA. E110 alloy is shown to be suitable material for the WWER fuel rod claddings under the normal operating conditions. E635 alloy is attractive to manufacturing of the skeleton components. The currently used combination (E110 as a material of fuel rods claddings and E635 - as a material of the skeleton components) is the optimal solution for the WWER fuel assembly because the advantages of the both alloys are used. (authors)

  4. Fuel pin failure root causes and power distribution gradients in WWER cores

    International Nuclear Information System (INIS)

    Mikus, J.

    2008-01-01

    The purpose of this work is to investigate the influence of some core heterogeneities and reactor construction materials on space power distribution in WWER type cores, especially from viewpoint of the values and gradient occurrence that could result in static loads with some consequences, e.g., fuel pin (FP) or fuel assembly (FA) bowing and possible contribution to the FP failure root causes. Presented information were obtained by means of experiments on research reactor LR-0 concerning the: 1) Power distribution estimation on pellet surface of the FPs neighbouring a FP containing gadolinium (Gd 2 O 3 ) burnable absorber integrated into fuel in WWER-440 and -1000 type cores; 2) Power distribution measurement in periphery FAs neighbouring the baffle in WWER-1000 type cores and 3) Power distribution in FAs neighbouring the control rod absorbing part in a WWER-440 type core. (author)

  5. A new mechanism of hydrogen absorption in water-water reactor core materials

    International Nuclear Information System (INIS)

    Gann, V.V.; Gann, A.V.

    2012-01-01

    The spectrum of fast protons, generated in water by fast neutrons of WWER-1000 reactor core, has been calculated using the code MCNPX. The main mechanism of fast proton generation in the moderator is found to be elastic scattering of fast neutrons on hydrogen nuclei. Fast protons with mean energy 1 MeV flow towards the surface of cladding material at flux density ∼ 0.1 μA/cm 2 . Proton range distribution profile in cladding material is calculated. The range of fast protons in zirconium averages 20 μm, the maximal proton range is larger than 200 μm. The rate of hydrogen deposition in 40 μm layer amounts to 5 x 10 -5 H/n/μ. A role of the suggested mechanism in process of zirconium clad hydrogenation during reactor irradiation is discussed.

  6. Comparative analysis of the results obtained by computer code ASTEC V2 and RELAP 5.3.2 for small leak ID 80 for VVER 1000

    International Nuclear Information System (INIS)

    Atanasova, B.; Grudev, P.

    2011-01-01

    The purpose of this report is to present the results obtained by simulation and subsequent analysis of emergency mode for small leak with ID 80 for WWER 1000/B320 - Kozloduy NPP Units 5 and 6. Calculations were performed with the ASTEC v2 computer code used for calculation of severe accident, which was designed by French and German groups - IRSN and GRS. Integral RELAP5 computer code is used as a reference for comparison of results. The analyzes are focused on the processes occurring in reactor internals phase of emergency mode with significant core damage. The main thermohydraulic parameters, start of reactor core degradation and subsequent fuel relocalization till reactor vessel failure are evaluated in the analysis. RELAP5 computer code is used as a reference code to compare the results obtained till early core degradation that occurs after core stripping and excising of fuel temperature above 1200 0 C

  7. Storage of spent nuclear fuel: the problem of spent nuclear fuel in Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, Z; Vapirev, E [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    A review of existing technologies for wet and dry storage of spent nuclear fuel (SNF) and the reprocessing policies is presented. The problem of SNF in Bulgaria is arising from nonobservance of the obligation to return SNF back to the former Soviet Union as agreed in the construction contract. In November 1994 approximately 1800 fuel assemblies have been stored in away-from-reactor (AFR) facility and another 1060 in at-reactor (AR) pools. The national policy is to export SNF out of the country. The AFR facility has a limited capacity and it is designed only for WWER-440 fuel although work is going on to extend it in order to store WWER-1000 SNF. 14 refs.

  8. Twenty years nuclear power sector in Bulgaria - an attempt at striking a balance

    International Nuclear Information System (INIS)

    Shervashidze, N.

    1994-01-01

    The advantages of nuclear power and the mistakes that cause the opposition against it on the example of the history of the Bulgarian nuclear power sector are considered in a polemical manner. Four WWER 440/230 and two WWER 1000 units have been put into operation in Kozloduj NPP in the period 1974 - 1993. Before 1991, however, safety operation criteria were compromised numerous times and the atmosphere of secrecy, self-complacency and the series of absurdities about the Chernobyl accident were clearly used by the opponents of nuclear energy. Now the attitude towards nuclear power is changing for the better. Much progress has been made in recovering the nuclear power sector under complicated economic and political conditions. The role of NPP as an anti-inflation factor, an element providing the national security, a stimulator of industry and infrastructure, and an ecological salvation factor, is stressed. (I.M.)

  9. Kozloduy NPP - its 20 years operational experience; AEhS `Kozloduy`: 20-letni opyt ehksploatatsii

    Energy Technology Data Exchange (ETDEWEB)

    Yankov, M; Grizanov, I; Sabinov, S [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    The main technical and economic results of 20 year operation of the Kozloduy NPP are summarised. The total electricity produced amounts to more than 200 billions KWh. The energy produced by the plant accounts for more than 40% of the national electricity output since 1993. Details of electricity production in different years by all 6 units are given. The efficiency of units 1-4 (WWER-440) is 29-30% and of units 5-6 (WWER-1000) is 29-33%. The Unit 6 has best characteristics. Since 1989 a safety test is being carried out in cooperation with the IAEA. Units 1 and 2 have been reconstructed in order to enhance safety and reliability. 1 tab.

  10. Team of experts concludes review of safety issues at Temelin

    International Nuclear Information System (INIS)

    2001-01-01

    Full text: At the request of the Czech Government, the International Atomic Energy Agency (IAEA) assembled a team of national experts from Bulgaria, France, Germany, Spain, and the United Kingdom, with an observer from Austria, to review safety issues at the Temelin power plant that were identified in 1996 as relevant to reactors of the generic Temelin design (WWER-1000/320 type). Following a detailed on-site review from 18 to 23 November 2001, the experts concluded that most identified issues had been addressed and resolved. Work is continuing on the few remaining issues. These issues, however, are not judged by them to be significant and would not from the experts' standpoint preclude the safe operation of the Temelin nuclear power plant. The final report of the team of experts will be available to the Czech Government in one month's time. (author)

  11. Linear step drive

    International Nuclear Information System (INIS)

    Haniger, L.; Elger, R.; Kocandrle, L.; Zdebor, J.

    1986-01-01

    A linear step drive is described developed in Czechoslovak-Soviet cooperation and intended for driving WWER-1000 control rods. The functional principle is explained of the motor and the mechanical and electrical parts of the drive, power control, and the indicator of position are described. The motor has latches situated in the reactor at a distance of 3 m from magnetic armatures, it has a low structural height above the reactor cover, which suggests its suitability for seismic localities. Its magnetic circuits use counterpoles; the mechanical shocks at the completion of each step are damped using special design features. The position indicator is of a special design and evaluates motor position within ±1% of total travel. A drive diagram and the flow chart of both the control electronics and the position indicator are presented. (author) 4 figs

  12. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3A. Kozloduy NPP units 5/6: Analysis/testing. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. This volume of Working material contains reports related analyses and testing of Kozloduy nuclear power plant, units 5 and 6

  13. Report of the ASSET (Assessment of Safety Significant Events Team) mission to the Khmelnitsky nuclear power plant in Ukraine 8-19 March 1993. Root cause analysis of operational events with a view to enhancing the prevention of accidents

    International Nuclear Information System (INIS)

    1993-01-01

    This IAEA Assessment of Safety Significant Events Team (ASSET) Report presents the result of an ASSET team's assessment of their investigation of the effectiveness of the plant policy for prevention of incidents since 1988 at Khmelnitsky nuclear power plant. The plant's one WWER 1000 MW(e) type unit has been in commercial operation since 1987. The results, conclusions and suggestions presented herein reflect the views of the ASSET experts. They are provided for consideration by the responsible authorities in Ukraine. The ASSET team's views presented in this report are based on visits to the plant, on review of documentation made available by the operating organization and on discussion with utility personnel. The report is intended to enhance operational safety at Khmelnitsky by proposing improvements to the policy for the prevention of incidents at the plant. The report includes, as a usual practices, the official responses of the regulatory body operating organization to the ASSET recommendations. Figs, tabs

  14. Algorithmization of problems on the personnel information support in the automatic chemical control systems at NPP

    International Nuclear Information System (INIS)

    Vilkov, N.Ya.; Kryukov, Yu.V.; Cheshun, A.V.

    2001-01-01

    When elaborating software for the standard algorithms of the information support of the efficient control (keeping) of water chemistry operation (WCO) at the NPP power units one introduces an approach when the systems of chemical control are realized as the systems of quality control of in-loop physical and chemical processes gathering force in the course of time. Elaboration of algorithms to proceed data of the operational chemical control seeks for elaboration of the statistic procedures to detect anomalies of the processes at the early stages of their development more efficient in contrast to the standard procedures of control. The introduced procedure is used in the demonstration model of the system for diagnostics of some typical reasons of violation of the first circuit WCO of WWER-1000 power units [ru

  15. Behaviour of Spent WWER fuel under long term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kadarmetov, I M [A.A.Bochvar All-Russia Research Institute of Inorganic Materials, Moscow (Russian Federation)

    1999-07-02

    Results of experimental investigation into thermomechanical properties of pre-irradiated Zr-1%Nb alloy over a range temperatures 500-570 grad C are presented. Safety examination of the Ventilation Storage Casks dry storage system has been carried out. Preliminary safety criteria under dry storage conditions in an environment of inert gas are follows: maximum cladding temperature under normal conditions of dry storage should not exceed 330 grad C after 5-year cooling in water-filled pools; maximum allowable temperature of spent fuel rod cladding under operational mode with infringement of heat removal should not exceed 440 grad C over 8 hours. As each SFA dry storage project comprises its individual technology of spent fuel management, it is necessary to evaluate allowable parameters (terms of storage, maximum temperatures of fuel) for each project respectively. The programme of experimental investigations for the justification of safety criteria for WWER-1000 dry spent fuel storage systems is underway. (author)

  16. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3B. Kozloduy NPP units 5/6: Analysis/testing. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The Co-ordinated research programme on the benchmark study for the seismic analysis and testing of WWER-type nuclear power plants was initiated subsequent to the request from representatives of Member States. The conclusions adopted at the Technical Committee Meeting on Seismic Issues related to existing nuclear power plants held in Tokyo in 1991 called for the harmonization of methods and criteria used in Member States in issues related to seismic safety. The Consulltants' Meeting which followed resulted in producing a working document for CRP. It was decided that a benchmark study is the most effective way to achieve the principal objective. Two types of WWER reactors (WWER-440/213 and WWER-1000) were selected as prototypes for the benchmark exercise to be tested on a full scale using explosions and/or vibration generators. This volume of Working material contains reports related analyses and testing of Kozloduy nuclear power plant, units 5 and 6.

  17. Chemical technologies and life management of Ukrainian NPPs

    International Nuclear Information System (INIS)

    Arkhipenko, A.V.; Barbashev, S.V.; Litvinskij, L.L.; Masko, A.N.

    2000-01-01

    Now 11 units with WWER-1000 reactors, 2 units with WWER-440 and 1 unit with RBMK-1000 are operated in Ukraine. State of chemical technologies of NPPs essentially influences on unit operating resource by the next ways: decreasing of corrosion intensity of equipment metal; decreasing of contamination on thermal exchanged surfaces of equipment; decreasing of amounts of radioactive waste. Improvement of these parameters can be achieved by the next measures: improvement of purification schemes for feed water of main systems; introduction of more effective water-chemical regimes (WCR); implementation of new methods and instruments of chemical monitoring for WCR; providing of without-scale regime of thermal-exchanged equipment operating by reactor division users through optimisation of the WCR of the NPP spray pool. (author)

  18. Experimental data and calculation studies of critical heat fluxes at local disturbances of geometry of WWER fuel assemblies

    International Nuclear Information System (INIS)

    Kobzar, L.L.; Oleksyuk, D.A.

    2001-01-01

    The results of experiments executed in RRC 'Kurchatov Institute on the thermal-physical critical facility SVD are presented herein. The experiments modeled the drawing of two fuel rods to each other till touching WWER-1000 reactor in FA. The experimental model is a 7-rod bundle with the heated length of 1 m. The primary goal of experiments was to acquire the quantitative factors of the reduction in the critical heat fluxes as contrasted to the basic model (without disturbances of FA geometry) at the expense of local disturbance of a rod bundle geometry. As it follows from the experiment, the effect of decrease of the critical heat rate depends on combination of regime parameters and it makes 15% in the most unfavorable case (Authors)

  19. Saturated steam turbines for power reactors of WWER-type

    International Nuclear Information System (INIS)

    Czwiertnia, K.

    1978-01-01

    The publication deals with design problems of large turbines for saturated steam and with problem of output limitations of single shaft normal speed units. The possibility of unification of conventional and nuclear turbines, which creates the economic basis for production of both types of turbines by one manufacturer based on standarized elements and assemblies is underlined. As separate problems the distribution of nuclear district heating power systems are considered. The choice of heat diagram for district heating saturated steam turbines, the advantages of different diagrams and evaluaton for further development are presented. On this basis a program of unified turbines both condensing and district heating type suitable for Soviet reactors of WWER-440 and WWER-1000 type for planned development of nuclear power in Poland is proposed. (author)

  20. Report of an experts' mission to review the modernization programme of the Kozloduy NPP units 5 and 6 Kozloduy, Bulgaria 26 June - 1 July 1995

    International Nuclear Information System (INIS)

    1995-01-01

    The purpose of the IAEA Safety Review Mission was to review the safety aspects of the Kozloduy 5/6 modernization programme, and to advise on the completeness and adequacy of the safety improvements proposed. The IAEA draft report ''Ranking of Safety Issues for WWER-1000 Model 320 NPPs'', WWER-SC-104, 1995-05-29 (referred to as the Issue Book in the subsequent text) was the basis for the experts' review. The IAEA NUSS standards and guides, and internationally accepted safety concepts and practices and national standards, complemented by the Russian safety rules in force, were used in the preparation of the issue book. The scope of the review covered the plant design and operational safety aspects as proposed in the modernization programme. The upgrading measures, which are only related to the improvement of plant availability, were not included in the review

  1. Peculiarities of highly burned-up NPP SNF reprocessing and new approach to simulation of solvent extraction processes

    Energy Technology Data Exchange (ETDEWEB)

    Fedorov, Y.S.; Zilberman, B.Y.; Goletskiy, N.D.; Puzikov, E.A.; Ryabkov, D.V.; Rodionov, S.A.; Beznosyuk, V.I.; Petrov, Y.Y.; Saprykin, V.F.; Murzin, A.A.; Bibichev, B.A.; Aloy, A.S.; Kudinov, A.S.; Blazheva, I.V. [RPA ' V.G.Khlopin Radium Institute' , 28, 2 Murinsky av., St-Petersburg, 194 021 (Russian Federation); Kurenkov, N.V. [Institute of Industrial Nuclear Technology NRNU MEPHI, 31, Kashirskoye shosse, Moscow, 115409 (Russian Federation)

    2013-07-01

    Substantiation, general description and performance characteristics of a reprocessing flowsheet for WWER-1000 spent fuel with burn-up >60 GW*day/t U is given. Pu and U losses were <0.1%, separation factor > 10{sup 4}; their decontamination factor from γ-emitting fission products was 4*10{sup 4} and 3*10{sup 7}, respectively. Zr, Tc, Np removal was >98% at U and Pu losses <0.05%. A new approach to simulation of extraction equilibrium has been developed. It is based on a set of simultaneous chemical reactions characterized by apparent concentration constants. A software package was created for simulation of spent fuel component distribution in multistage countercurrent extraction processes in the presence of salting out agents. (authors)

  2. Analysis of emergency operating procedures effectiveness for core damage prevention using computer code RELAP for nuclear power plants with VVER-1000/B-320 in reference to primary to secondary circuit leak with external power loss and BRU-A stuck open failure

    International Nuclear Information System (INIS)

    Arkhangelski, L.; Sheveliov, D. V.

    1999-01-01

    This report presents analysis of development emergency operating procedures effectiveness for possible accident on nuclear power plant with WWER-1000 reactor type. Accident initiating event is the primary to secondary circuit leak caused by steam generator primary cover lift-up. In according to conservative assumptions the following additional failures were considered: dump valve BRU-A stuck open failure; loss of external power. The results of this work are represented as a comparative analysis of two possible ways of accident evolution: according to functioning automatic safety systems responses; according to accident management based on development emergency operating procedures with operator intervention. Developed emergency operating procedures assure the following significant goals to mitigate accident sequences: optimal use of ECCS water inventory; severe core damage prevention; mitigation of environment radioactive contamination. (authors)

  3. Experience of development of the methods and equipment and the prospects for creation of WWER fuel examination stands

    International Nuclear Information System (INIS)

    Pavlov, S.; Smirnov, V.

    1998-01-01

    The report presents the basic methods and equipment developed for inspection of the fuel elements and fuel assemblies in the spent fuel pools. It considers their characteristics and results of the tests under laboratory and experimental fuel examination stand conditions. In particular, the following techniques are presented: visual inspection, measurement of the geometrical dimensions, definition of the form change in fuel assemblies and fuel elements, detection of the failed fuel elements, etc. The experience of the experimental fuel examination stand operation is generalized. The concept of the creation of the WWER-440 and WWER-1000 FA and FE inspection stands is presented. The concept is based on the modular principle which runs as follows. A set of the basic functional blocks is being developed based on which it is possible to make such a stand configuration which is necessary to fulfil the specific program of the examination at the particular nuclear power plant. (author)

  4. Increasing of leak-tightness of hermetic zone and other important components in NPPs

    International Nuclear Information System (INIS)

    Murani, J.

    2001-01-01

    The performed by VUEZ resealing works in the hermetic compartments/containment of WWER 440 and WWER 1000 NPPs in Slovakia (Jaslovske Bohunice V-1 NPP, Jaslovske Bohunice V-2 NPP, Mochovce NPP), the Czech Republic (Dukovany NPP, Temelin NPP) and Hungary (Pask NPP) are presented. The leakage rate from hermetic compartments at individual NPP reactor units are shown.The VUEZ activities in field of resealing of weld joints in pools and tanks in the Jaslovske Bohunice V-1 NPP and innovation of the seal of reactor pit protective hood (kolpak) are described. The principle of operation and practical experience of the SMU-V system implemented in the Jaslovske Bohunice V-1 NPP (WWER 440 - V 230) for NPP primary circuit leakage monitoring are also given

  5. Comparison of the APA-H (Westinghouse) calculations with the operational data for ZpNPP unit 3 cycles 16-19

    International Nuclear Information System (INIS)

    Abdullayev, A. M.; Gorbachenko, O. V.; Ignatchenko, A.I.; Maryokhin, S.V.; Zhukov, A. I.

    2007-01-01

    The computer simulation of ZpNPP Unit 3 (WWER-1000) Cycles 16-19 core depletion has been performed on the basis of the operational data. The changes in reactor heat rate, lead bank position and inlet temperature during the core operation have been taken into account. These calculations were performed by using Westinghouse APA-H (ALPHA/PHOENIX/ ANC-H) code system. The main objectives of the calculations were the comparison with operational data for core loading with TVS-M (Cycle 16) and transition core loading with TVS-A (Cycle 17-19). The calculation results were compared with the results of Critical boric acid concentration vs. Cycle Burnup measurements and Start up Physics Test measurements (at HZP, BOC, NoXe core conditions). Additionally, the comparison between the results of assemblies power calculation performed by ANC-H and BIPR-7A codes is presented (Authors)

  6. Databases on safety issues for WWER and RBMK reactors. Users' manual. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-04-01

    At the beginning of the IAEA Extrabudgetary Programme on the safety of WWER reactors a great number of findings and recommendations (safety items) were collected as a result of design review and safety review missions of the WWER-440/230 type reactors. On the basis of these findings a technical database containing more than 1300 records was established to support the consolidation of the information obtained and to help in identification of safety issues. After the scope of the WWER extrabudgetary programme was extended similar data sets were prepared for the WWER-440/213, WWER-1000 and RBMK nuclear power plants. This publication describes the structure of the databases on safety issues of WWER and RBMK NPPs, the information sources used in the databases and interrogation capabilities for users to obtain the necessary information. 14 refs, 9 figs, 5 tabs

  7. Comparison of CASMO and NESSEL few group cross section libraries and their usage in DYN3D

    International Nuclear Information System (INIS)

    Kuchin, A.; Ovdiyenko, Y.; Loetsch, T.

    2007-01-01

    This work presents comparative analysis of two group diffusion cross sections libraries which were generated by NESSEL-4 and CASMO-4 lattice codes. Diffusion parameters were calculated for WWER-1000 fuel assemblies with stainless steel spacing grids and guiding tubes. These cross section sets were introduced into reactor core code DYN3D and tested on the base of real reactor core states. In this case operation data of the first three fuel cycles of sixth unit of Zaporizhzhya NPP were used. The work was performed in the framework of the order BMU SR 2511 - 862 500/09, UA-2575. The report describes the opinion and view of the contractor - TUV ENERGIE CONSULT - and does not necessarily represent the opinion of the ordering party - Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (Authors)

  8. SKODA Nuclear Machinery - tradition and expertise in nuclear power industry

    International Nuclear Information System (INIS)

    Svitak, F.

    1997-01-01

    The SKODA Nuclear Machinery company is a major manufacturer of nuclear reactor assemblies and supplier of WWER type primary coolant circuits. In the past, the company was nearly a monopolistic manufacturer of WWER reactor assemblies supplied to the Central and East European countries (except the USSR) grouped in the former Council of Mutual Economic Assistance. Over the 1980-1993 period, 21 units of the WWER-440 type and 3 units of the WWER-1000 type were manufactured. The company keeps abreast of technological progress and has been switching to new manufacturing areas, such as compact storage racks for spent fuel pools, hermetic cable bushings, spent fuel storage and transport casks, and cooperation in the manufacture of neutron flux measuring channels. Technological services provided to nuclear power plants constitute another important field of the company's business. The company's combined expertise in Soviet and Western designed PWRs is a considerable asset. (P.A.)

  9. Effluents and releases of tritium from Novo-Voronezh-5 reactor

    International Nuclear Information System (INIS)

    Babenko, A.G.; Mekhedov, B.N.; Podporinova, L.E.; Popov, S.V.; Shalin, A.N.

    1990-01-01

    Results of systematic measurements of tritium concentration within technological systems of reactor of Novo-Voronezh NPP conducted to evaluate tritium effluents and releases and radiation doses to population from these effluents and releases are given. It is shown that 68% concerning tritium total amount were disposed into sewerage while 17% - through vent tube and 15% - with water and steam from secondary circuit systems. Standartized tritium effluents from WWER-1000 reactor for 5 year run constitute 15±1.9 GBq/MWxyear and it corresponds to mean value of effluents for foreign NPPs. Tritium concentration in the atmosphere constituted according to calculations (4.1-20)x10 -5 Bq/l. Conclusion is made about insignificant dose to population from tritium gaseous effluents. Detail study is necessary for dose connected with tritium contained in water effluents

  10. Overview of large scale experiments performed within the LBB project in the Czech Republic

    Energy Technology Data Exchange (ETDEWEB)

    Kadecka, P.; Lauerova, D. [Nuclear Research Institute, Rez (Czechoslovakia)

    1997-04-01

    During several recent years NRI Rez has been performing the LBB analyses of safety significant primary circuit pipings of NPPs in Czech and Slovak Republics. The analyses covered the NPPs with reactors WWER 440 Type 230 and 213 and WWER 1000 Type 320. Within the relevant LBB projects undertaken with the aim to prove the fulfilling of the requirements of LBB, a series of large scale experiments were performed. The goal of these experiments was to verify the properties of the components selected, and to prove the quality and/or conservatism of assessments used in the LBB-analyses. In this poster, a brief overview of experiments performed in Czech Republic under guidance of NRI Rez is presented.

  11. Building technology on construction site of nuclear power plant at Zaporozh'e

    Energy Technology Data Exchange (ETDEWEB)

    Dusek, R; Matyas, V [Vodni Stavby, Prague (Czechoslovakia)

    1981-12-01

    Basic data and technical and economic indexes are shown for a WWER 1000 nuclear power plant being built 120 km off Zaporozh'e (USSR). The schedule of construction and the choice of the means of mechanization used for building work are reported. Discussed are building machines used, the location of assembly cranes of the main unit, the design and the building technology of the reactor part, the engine house, deisel generator station, the special operations building, the use of concrete and steel units in the building, and the procurement of materials for the construction. The knowledge gained from the building of the power plant will be applied in the CSSR in the building of 1000 MW unit power plants.

  12. Seismic margin assessment and earthquake experience based methods for WWER-440/213 type NPPs

    International Nuclear Information System (INIS)

    Masopust, R.

    1996-01-01

    This report covers the review of the already completed studies, namely, safe shutdown system identification and classification for Bohunice NPP and the comparative study of standards and criteria. It contains a report on currently ongoing studies concerning seismic margin assessment and earthquake experience based methods in application for seismic evaluation and verification of structures and equipment components of the operating WWER-440/213 type NPPs. This is based on experiences obtained from Paks NPP. The work plan for the remaining period of Benchmark CRP and the new proposals are included. These are concerned with seismic evaluation of selected safety related mechanical equipment and pipes of Paks NPP, and the actual seismic issues of the Temelin WWER-1000 type NPP

  13. Fission product release from fuel of water-cooled reactors

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.; Klisinska, M.

    1997-01-01

    The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

  14. About the use of approximations, which ensure materials mass balance conservation by spatial meshes, in Sn full core calculations

    International Nuclear Information System (INIS)

    Voloshchenko, A.M.; Russkov, A.A.; Gurevich, M.I.; Olejnik, D.S.

    2008-01-01

    One analyzes a possibility to make use of the geometry approximations conserving the materials mass local balance in every mesh via adding of mixtures in the meshes containing several feed materials to perform the kinetic calculation of the reactor core neutron fields. To set the 3D-geometry of the reactor core one makes use of the combinatorial geometry methods implemented in the MCI Program to solve the diffusivity equations by the Monte Carlo method, to convert the combinatorial prescribing of the geometry into the mesh representation - the ray tracing method. According to the calculations of the WWER-1000 reactor core and the simulations of the spent fuel storage facility, the described procedure compares favorably with the conventional geometry approximations [ru

  15. Adaptation of GRS calculation codes for Soviet reactors

    International Nuclear Information System (INIS)

    Langenbuch, S.; Petri, A.; Steinborn, J.; Stenbok, I.A.; Suslow, A.I.

    1994-01-01

    The use of ATHLET for incident calculation of WWER has been tested and verified in numerous calculations. Further adaptation may be needed for the WWER 1000 plants. Coupling ATHLET with the 3D nuclear model BIPR-8 for WWER cores clearly improves studies of the influence of neutron kinetics. In the case of FBMK reactors ATHLET calculations show that typical incidents in the complex RMBK reactors can be calculated even though verification still has to be worked on. Results of the 3D-core model QUABOX/CUBBOX-HYCA show good correlation of calculated and measured values in reactor plants. Calculations carried out to date were used to check essential parameters influencing RBMK core behaviour especially dependence of effective voidre activity on the number of control rods. (orig./HP) [de

  16. To the choice of the regeneration system of the K-1000-68/1500 turbine plant for the NPP with a vertical-type steam generator

    International Nuclear Information System (INIS)

    Kuznetsov, N.M.; Piskarev, A.A.; Grinman, M.I.; Kruglikov, P.A.

    1985-01-01

    Several variants of the heat regeneration system for the NPP with WWER-1000 type reactors using vertical steam generator (SG) generating saturated steam at 7.2 MPa pressure and 200 deg C feed water temperature at the SG inlet are considered. The results of comparison of variants in water and steam circuits of turbine plants are greatly influenced by integral economy account, i.e. efficiency indexes account under variable conditions of power unit operation. From variants of water and steam circuits of the K-1000-68/1500 turbine plant considered preference is given to the variant with four low pressure heaters with increased up to 1.25 MPa pressure in a deacrator without high pressure heater with pumping intermediate steam superheater condensate into feedwater circuit

  17. District heating by the Bohunice nuclear power plant

    International Nuclear Information System (INIS)

    Metke, E.; Skvarka, P.

    1984-01-01

    Technical and economical aspects of district heating by the electricity generating nuclear plants in Czechoslovakia are discussed. As a first stage of the project, 240 MW thermal power will be supplied using bleeding lines steam from the B-2 nuclear power plant at Jaslovske Bohunice to heat up water at a central station to 130 grad C. The maximal thermal power that can be produced for district heating by WWER type reactors with regular condensation turbines is estimated to be: 465 MW for a WWER-440 reactor with two 220 MWe turbines and 950 MW for a WWER-1000 reactor with a Skoda made 1000 MWe turbine using a three-stage scheme to heat up water from 60 grad C to 150 grad C. The use of satelite heating turbines connected to the steam collector is expected to improve the efficiency. District heating needs will de taken into account for siting of the new power plants

  18. Nuclear fuel licensing procedures in Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    Harizanov, Y [Komitet za Mirno Izpolzuvane na Atomnata Energiya, Sofia (Bulgaria). Komysia za Biologichni i Selskostopanski Nauki

    1994-12-31

    A brief description of the structure and role of the Committee on the Use of Atomic Energy for Peaceful Purposes (BG) as a main governmental institution responsible for safety atomic energy management is presented. The main documentation and licensing procedures adopted for import, export, operation, storage and transportation of nuclear material including nuclear fuel for NPP are outlined. The corresponding institutions are facing now the urgent need for changing some regulations to meet the requirements of the international treaties and conventions recently signed by Bulgarian authorities. A new version of Atomic Law disposed at the Parliament for adoption is essential for updating the management of NPP in Bulgaria where four WWER-440 and two WWER-1000 are under operation.

  19. Nuclear Power Plant Control and Instrumentation activities in Czechoslovakia

    International Nuclear Information System (INIS)

    Stirsky, P.; Karpeta, C.; Rubek, J.

    1990-01-01

    After giving a survey of the Czechoslovak nuclear power plants a description of I and C systems of the operating plants is presented together with a brief outlook for future developments to be implemented at plants which are under construction. Special attention is paid to the adopted techniques for power distribution investigation and control in the WWER 1000 reactor core in the case of load changes. Basic futures of the in-core measurement systems are outlined. Measures implemented in the I and C systems of the operating units to improve their performance are described. Information on the country's approach to NPP personnel training and training aids usage as well as information on development work in the area of surveillance and monitoring systems completes the paper. (author). 9 refs, 1 tab

  20. Protective containment behaviour under exceeded design loads

    International Nuclear Information System (INIS)

    Holub, I.; Stepan, J.; Maly, J.; Schererova, K.

    2003-01-01

    The contribution describes the calculation results of the behaviour of containment structure if loaded in excess of its design load. The Temelin NPP comprises two WWER 1000 blocks and containment consists of a pre-stressed reinforced concrete structure with a system of unbonded cables. The objective of the calculations was to determine the level of load caused by the internal pressure and temperature at which the containment protective function would fail. In the first step, the maximum overpressure was determined, which may be transferred by the containment structure. In further steps analyses were made of various combinations of simultaneous pressure and temperature loads. The contribution presents relevant calculation results, including the evaluation of containment structure behaviour including liner under loads that exceed its design parameters. (author)

  1. Main problems of increasing labour productivity in the power plant construction

    International Nuclear Information System (INIS)

    Falaleev, P.P.

    1984-01-01

    The reserve for labour productivity growth in power-, industrial- and civil engineering in the USSR Minenergo system are discussed. Such reserve comprises: introduction of effective designs, increase of technological readiness of structures; a higher mechanization level in construction, improvement of industrial organization, economical and social aspects. Decrease of labour inputs in NPP construction will be attained by using unified designs of serial WWER-1000, RBMK-1000- and RBMK-1500 reactors as well as by developing nuclear power construction complexes-industrial-construction enterprises for manufacturing and transport of special structures as well as for performing civil engineering and installation work on the ground part of the reactor building and special structure. Other potentialities for increasing labour productivity in NPP construction are considered

  2. Development and mastering of production of dysprosium hafnate as absorbing material for control rods of promising thermal neutron reactors

    International Nuclear Information System (INIS)

    Zakharov, A.V.; Risovany, V.D.; Muraleva, E.M.; Sokolov, V.F.

    2011-01-01

    The main advantages of dysprosium hafnate as an absorbing material for LWR control rods are the following: -) unlimited radiation resistance; - two absorbing components, Dy and Hf, increasing physical efficiency of the material compared to Dy 2 O 3 -TiO 2 and alloy 80% Ag - 15% In - 5% Cd; -) variability of physical efficiency by changing a composition, but maintaining other performance characteristics of the material; -) high process-ability due to the absence of phase transients and single-phase structure (solid solution); -) production of high density pellets. Lab-scale mastering of dysprosium hafnate pellets production showed a possibility of material synthesis using a solid-phase method, as well as of dysprosium hafnate pellets production by cold pressing and subsequent sintering. Within a whole range of examined compositions (23 mol% - 75 mol% Dy 2 O 3 ), a single-phase material with a highly radiation resistant fluorite-like structure was produced. Experiments on cold pressing and sintering of pellets confirmed a possibility of producing high quality dysprosium hafnate pellets from synthesized powder. A pilot batch of dysprosium hafnate pellets with standard sizes was produced. The standard sizes corresponded to the absorbing elements of the WWER-1000 control rods and met the main requirements to the absorbing element columns. The pilot batch size was approximately 6 kg. Acceptance testing of the pilot batch of dysprosium hafnate pellets was conducted, fulfillment of the requirements of technical conditions was checked and preirradiation properties of the pellets were examined. High quality of the produced pellets was confirmed, thus, demonstrating a real possibility of producing large batches of the dysprosium hafnate pellets. The next step is the production of test absorbing elements and cluster assemblies for the WWER-1000 control rods with their further installation for pilot operation at one of the Russian nuclear power plants

  3. Examination of VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Markulin, K.

    2008-01-01

    The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

  4. ASTEC and MELCOR comparison for a VVER-1000 60 mm small break LOCA

    International Nuclear Information System (INIS)

    Georgieva, J.; Stefanova, A.; Groudev, P.; Tusheva, P.; Mladenov, I.; Dimov, D.; Passalacqua, R.

    2005-01-01

    In this paper a comparison between severe accident calculations performed for a WWER 1000 with the ASTEC1.1v0 and MELCOR 1.8.5 computer codes for a small break LOCA (ID 60 mm) without intervention of hydro accumulators is presented. This investigation has been performed in the framework of the SARNET project under the EURATOM 6th framework program. Once the accident sequence scenario is specified, both codes (MELCORE and ASTEC) are able to determine the core and containment damaged states, to estimate the release of radionuclides from the fuel as well as from the primary circuit and containment. Theses results are used to estimate the maximum period of the time during which the personnel could still take particular decisions in order to mitigate such an accident. The aim of the performed analysis is to estimate the discrepancy between ASTEC and MELCORE 1.8.5 calculations. Such discrepancies will be studied, if the case, proposal for ASTEC improvements will be made. Also the ASTEC capability to simulate specific reactor accident scenarios and/or particular safety systems will be tested. The final target is to propose severe accident management procedure for WWER 1000 reactors. In conclusions, the analysis for a small break LOCA (ID 60 mm without hydroelectricities) has shown some discrepancies between ASTEC and MELCORE especially during the degradation of the core. Further analyses are planed in which the MELCORE temperature 'set point' for core degradation (2520 K) will be progressively increased to approach the ASTEC one (which has been estimated to be about 3200 K). The comparison of the new results will allow a better evaluation of the in-vessel models implemented in ASTEC

  5. Safeguards at Kozloduy NPP - Experience and expectations

    International Nuclear Information System (INIS)

    Elenkov, Todor

    2001-01-01

    Bulgaria is a party of Non Proliferation Treaty since 5 September 1969. The agreement between IAEA and Bulgaria - INFCIRC 178 - has been in force since 29 February 1972. At that time Bulgaria had one research reactor IRT-2000 in Sofia and two power reactors of WWER-440 type under construction. Now at Kozloduy NPP site there are 4 facilities, which consist of 4 WWER-440 and 2 WWER-1000 type power reactors, producing almost 50% of the electricity in Bulgaria and 1 wet away from reactor spent fuel storage. In 1991 under the green movements and social pressure, the research reactor in Sofia was closed and the construction of the second NPP in Belene with 2 WWER-1000 type reactors was halted. After the transfer in 1994 of the fresh fuel from the research reactor to Kozloduy due to security reasons practically NPP Kozloduy remains the only significant (from safeguards point of view) nuclear site in Bulgaria. In 1972 a 'Nuclear Fuel' group was formed at the Physicists Department in NPP Kozloduy with responsibilities to carry out for safeguards records and reports, fresh and spent fuel transport and control. In 1990 this group was transferred to the Safety Section and since 1992 it exists as 'Control and Accounting for of the Nuclear Materials' - a section in the Safety Department. Currently the section serves all four facilities in NPP Kozloduy and has four people: section head, chief inspector and two inspectors. The main activities of the section include: a) Control of the nuclear fuel location as well as meeting the storage and transport conditions regulations; b) Control of the conditions for normal operations of the installed IAEA surveillance systems; c) Preparation of documents for licensing of fresh and spent nuclear fuel transport; d) Preparation of the official information on nuclear materials location and quantity; e) Preparation of accounting records and the reports for IAEA (ICR, PIL, MBR); f) Co-ordination of the IAEA safeguards inspection activities at NPP

  6. Power distribution gradients in WWER type cores and fuel failure root causes

    Energy Technology Data Exchange (ETDEWEB)

    Mikuš, Ján M., E-mail: JanMikus.nrc@hotmail.com

    2014-02-15

    Highlights: • Power (fission rate) distribution gradients can represent fuel failure root causes. • Positions with above gradients were investigated in WWER type cores on reactor LR-0. • Above gradients were evaluated near core heterogeneities and construction materials. • Results can be used for code validation and fuel failure occurrence investigation. - Abstract: Neutron flux non-uniformity and gradients of neutron current resulting in corresponding power (fission rate) distribution changes can represent root causes of the fuel failure. Such situation can be expected in vicinity of some core heterogeneities and construction materials. Since needed data cannot be obtained from nuclear power plant (NPP), results of some benchmark type experiments performed on light water, zero-power research reactor LR-0 were used for investigation of the above phenomenon. Attention was focused on determination of the spatial power distribution changes in fuel assemblies (FAs): Containing fuel rods (FRs) with Gd burnable absorber in WWER-440 and WWER-1000 type cores, Neighboring the core blanket and dummy steel assembly simulators on the periphery of the WWER-440 standard and low leakage type cores, resp., Neighboring baffle in WWER-1000 type cores, and Neighboring control rod (CR) in WWER-440 type cores, namely (a) power peak in axial power distribution in periphery FRs of the adjacent FAs near the area between CR fuel part and butt joint to the CR absorbing part and (b) decrease in radial power distribution in FRs near CR absorbing part. An overview of relevant experimental results from reactor LR-0 and some information concerning leaking FAs on NPP Temelín are presented. Obtained data can be used for code validation and subsequently for the fuel failure occurrence investigation.

  7. Code systems for effective and precise calculation of the basic neutron characteristics, core loading optimization, analysis and estimation of the operation regimes of WWER type reactors

    International Nuclear Information System (INIS)

    Apostolov, T.; Ivanov, K.; Prodanova, R.; Manolova, M.; Petrova, T.; Alekova, G.

    1993-01-01

    Two directions for investigations are suggested: 1) Analysis and evaluation of the real loading patterns and operational regimes for Kozloduy NPP WWER-440 and WWER-1000 in the frame of the recent safety criteria and nuclear power plant operating limits. 2) Development of modern code system for WWER type reactor core analysis with advanced features: new design and materials for fuel and control rods, increasing the fuel enrichment, using the integral and discrete burnable absorbers etc. The fuel technology design evolution maximizes the fuel utilization efficiency, improves operation performance and enhances safety margins. By the joint efforts of specialists from INRNE, Sofia (BG) and KAB, Berlin (GE), the codes NESSEL-IV-EC, PYTHIA and DERAB have been developed and verified. In the frame of the PHARE programme the joint project ASPERCA has been proposed intended for reactor physics calculations with PHYBER-WWER code for safety enhancement and operation reliability improvement. In-core fuel management benchmarks for 4 cycles of unit 2 (WWER-440) and 2 cycles of unit 5 (WWER-1000) have been performed. The coordination of burnable absorber design implementation, low leakage loadings usage, reloading enrichment increase and steel content reduction in the core have made the reactor core analysis more demanding and the definition of loading patterns - more difficult. This complexity requires routine use of three-dimensional fast accurate core model with extended and updated cross section libraries. To meet the needs of WWER advanced loading patterns and in-core fuel management improvements the HEXANES code systems is being developed and qualified. Some test calculations have been carried out by the HEXANES code system investigating the influence of Gd in the fuel on the main reactor physics parameters. For reevaluation of the core safety-related design limits forming the basis of licensing procedure, the code DYN3D/M2 is used. 16 refs., 3 figs. (author)

  8. WWER safety investigations on LR-0 reactor

    International Nuclear Information System (INIS)

    Mikus, J.

    2001-01-01

    A set of the measurement needed for the WWER-440 and WWER-1000 reactor lifetime assessment, verification of the methods, codes and input cross section libraries for the WWER reactor pressure vessel exposure evaluation has been performed on the LR-0 experimental reactor. The WWER Mock-ups (engineering benchmarks) has been carried out on the reactor, with the aim to investigate differential neutron spectra for reactor dosimetry purposes. Critical experiments have also been performed to determine the perturbation of the fission density distribution caused by the WWER-440 control assembly. Such assembly, partially inserted in the core, has significant influence on the space power distribution. A wide research program for sub-criticality investigations of the spent nuclear fuel storage has been realized on the LR-0 reactor. A benchmark experiment is realized on the reactor in corresponding geometry for CASTOR 440/84 container for storage and transportation of spent fuel. Critical experiments with new fuel assemblies including various burnable absorbers and different enrichments are performed. A set of critical experiments is performed using the fuel assemblies with 3,6% and 4,4% enrichment, arranged in the WWER-440 type cores with various lattice pitch. The critical high of the moderator level and the moderator level coefficient of reactivity are measured and the effect of the fuel assembly, placed in a hexagonal tube of stainless steel containing boron absorber (ATABOR - STANDARD) is investigated. The obtained results are used for the validation of the codes (MCNP, KENO and SCALE) in the frame of the contract 'Burn-up credit implementation for the storage and transport containers of the spent fuel'. Combined neutron-gamma spectra measurements in the WWER-1000 Mock-up are carried out during 2001

  9. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3I. Kozloduy NPP units 5/6: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  10. Current state and perspectives of spent fuel storage in Russia

    International Nuclear Information System (INIS)

    Kurnosov, V.A.; Tichonov, N.S.; Makarchuk, T.F.

    1999-01-01

    Twenty-nine power units at nine nuclear power plants, having a total installed capacity of 22 GW(e), are now in operation in the Russian Federation. They produce approximately 12% of the generated electricity in the country. The annual spent fuel arising is approximately 790 tU. The concept of the closed fuel cycle was adopted as the basis for nuclear power development in the Russian Federation, but until now this concept is only implemented for the fuel cycles of WWER-440 and BN-600 reactors. The WWER-1000 spent fuel is planned to be reprocessed at the reprocessing plant RT-2 which is under construction near Krasnoyarsk. The RBMK-1000 spent fuel is not reprocessed. It is meant to be stored in intermediate storage facilities at the NPP sites. The status of the spent fuel (SF) stored in the storage facilities is given in the paper. The principal characteristics of the fuel cycles of the Russian NPPs in the period up to 2015 is also given in the report. The key variant of the current spent fuel management at RBMK-1000 NPPs is storage in at-reactor and in away-from-reactor wet storage facilities at the power plant site with a capacity of 2,000 W. The storage capacity at the operating RBMKs (including the increase due to denser fuel assembly arrangement) will provide SF reception from the NPPs only up to 2005. For RBMK spent fuel, intermediate dry storage is foreseen at power plant sites in metallic concrete casks and thereafter transportation to the central storage facility at the RT-2 plant for long-term storage. The SF will be reprocessing after completion of the reprocessing plant at RT-2. In the Programme of Nuclear Power Development in the Russian Federation for the period 1998 to 2005 and for the period until 2010 year, provisions are made for the construction of a central dry storage facility before 2010. The facility will have a design capacity of 30,000 tU for WWER-1000 and RBMK-1000 spent fuel and is part of the reprocessing plant RT-2. The paper considers

  11. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3H. Kozloduy NPP units 5/6: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  12. Co-ordinated research programme on benchmark study for the seismic analysis and testing of WWER-type nuclear power plants. V. 3G. Kozloduy NPP units 5/6: Analysis and testing. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    In August 1991, following the SMiRT-11 Conference in Tokyo, a Technical Committee Meeting was held on the 'Seismic safety issues relating to existing NPPs'. The Proceedings of this TCM was subsequently compiled in an IAEA Working Material. One of the main recommendations of this TCM, called for the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing NPPs. Twenty four institutions from thirteen countries participated in the CRP named 'Benchmark study for the seismic analysis and testing of WWER type NPPs'. Two types of WWER reactors (WWER-1000 and WWER-440/213) selected for benchmarking. Kozloduy NPP Units 5/6 and Paks NPP represented these respectively as prototypes. Consistent with the recommendations of the TCM and the working paper prepared by the subsequent Consultants' Meeting, the focal activity of the CRP was the benchmarking exercises. A similar methodology was followed both for Paks NPP and Kozloduy NPP Unit 5. Firstly, the NPP (mainly the reactor building) was tested using a blast loading generated by a series of explosions from buried TNT charges. Records from this test were obtained at several free field locations (both downhole and surface), foundation mat, various elevations of structures as well as some tanks and the stack. Then the benchmark participants were provided with structural drawings, soil data and the free field record of the blast experiment. Their task was to make a blind prediction of the response at preselected locations. The analytical results from these participants were then compared with the results from the test. Although the benchmarking exercises constituted the focus of the CRP, there were many other interesting problems related to the seismic safety of WWER type NPPs which were addressed by the participants. These involved generic studies, i.e. codes and standards used in original WWER designs and their comparison with current international practice; seismic analysis

  13. Limit the effects of secondary circuit water or steam piping breaks in the reactor building

    International Nuclear Information System (INIS)

    Nachev, N.

    2001-01-01

    The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

  14. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  15. Radiobiological significance of radioactive contamination - summary assessment based on great number of measurements

    International Nuclear Information System (INIS)

    Angelov, V.; Bonchev, Ts.; Mavrodiev, V.; Kyrdzhilov, N.

    1995-01-01

    In order to facilitate quantitative and qualitative characterisation of radioactive contamination the authors introduce a relative estimate of radionuclide activity by setting as a reference the most abundant element -Co-60 in the case of the Kozloduy NPP. The ratio η i of the mean annual permissible concentration in air for each radionuclide (RPC-92) to that of Co-60 is calculated. It is found that η i has the same or close values for groups of radionuclides, e.g. η i = 2.10 -4 for 238 Pu, 239 Pu, 240 Pu, 241 Am, 244 Cm; η i = 5 for 89 Sr, 91 Y; 93 Nb, 134 Cs, 137 Cs; η i = 50 for 55 Fe, 63 Ni, 95 Zr, 95 Nb, 140 Ba, 140 La. Then it is compared to the experimentally measured values of the same quantity η iexp , derived from surface contamination data. The ratio η iexp /η i is plotted against log η i . The resulting nomograms give graphic representation of the radiobiological significance of various radionuclide groups. Data from different locations at the Kozloduy NPP are presented. It is found that the alpha emitter contamination has highest values in the Unit 1 (WWER-440) control rooms after repair. The Unit 5 (WWER-1000) has lower alpha contamination compared to WWER-440 units. 1 ref., 5 figs., 1 tab

  16. The World Bank mission in Bulgaria with the objective of estimating the possibility to increase the safety of the Kozloduy NPP

    International Nuclear Information System (INIS)

    Laaksonen, J.

    1992-01-01

    This is a report of the World Bank mission in Bulgaria held from 22 March to 31 April 1992. The author considers the present state of the safety level of all Kozloduy NPP reactors with emphasis of the units 1-4 which are of the WWER-440/230 type. Three main factors are discussed in detail: 1) design inherent safety; 2) equipment protection from accidental events connected with the equipment quality; 3) quality and culture of operation and quantitative evaluation of the safety. The areas requiring special attention in view of the safety enhancing are: safety goal policy; creation of national safety culture; planning and implementation of the improvement programme; assuring the equipment reliability; updating and protection the existing systems; furnishing the plant with new systems. A review of all on-going and proposed projects of the national authorities, IAEA, EC, WANO and BEQE is made. Some remarks are given also about the WWER-1000 type units 5 and 6. The estimated cost of all necessary measures for equipment improvement is $100-200 million and of the annual investment after the main reconstruction and innovation of the reactors - about $10 million which should be included in the normal operational expenses of the plant. A new approach to the responsibility role of the Bulgarian specialists should be created. A new and well grounded proposal for modifications in the equipment should be made from the National Electrical Company including economical evaluations. Another expert assessment will be requested from the foreign banks before financing

  17. Future prospects of nuclear power plant simulators in Czechoslovakia

    International Nuclear Information System (INIS)

    Panek, J.

    1983-01-01

    The article is devoted to the innovation of the WWER-440 simulator which is being updated because of the obsolete design of the RPP 16 S system whose production has been discontinued. The innovation will replace the greater part of the computer system. Technical requirements and software of the new computer are described. The innovation will proceed in three stages: the acquisition of hardware, modification and complementation of mathematical models and the development of software. No innovation will be made in interface unit DASIO-600 which is conceptually new and has been practically proven. This is also true of the reactor instrumentation simulation which is based on the modular microprocessor system SM 50-40 and meets all requirements. The development of the WWER-1000 simulator has been approved. The development and construction of the simulator proceeds in the following stages: functional specification of the simulator, mathematical model, project of unit control room, assembly and debugging of the computer system, processing of user programs, production and activation of the unit control room and assembly, activation and adjustment of the whole simulator. (B.S.)

  18. Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mohammadi, Ashgar; Omidvari, Nima [Iran Radioactive Waste Management Company, Tehran (Iran, Islamic Republic of); Hassanzadeh, Mostafa [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2017-12-15

    Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k{sub eff}) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

  19. Scientific-technical cooperation with foreign (esp. Europe and INSC partner countries) nuclear regulatory authorities and their technical support organizations in the fields of nuclear safety of operating nuclear power plants and on the concept evaluation of generation 3+ plants. Final report

    International Nuclear Information System (INIS)

    Wolff, Holger

    2016-09-01

    The BMUB/BfS-Project 3614I01512 forms the frame of the GRS for the scientific-technical cooperation with Technical Support Organisations and Nuclear Regulatory Authorities in the field of nuclear safety in operating NPPs and for the concept evaluation of generation 3 + plants in Europe and INSC Partner Countries. In the present final project report results are described which were gained within the project duration 15.10.2014 up to the 30.09.2016 in the following working packages: Investigations following the catastrophe of Fukushima Daiichi, Evaluation of selected National Action Plans, DBA and severe accident analyses for NPP with PWR (WWER-440, WWER-1000), cooperation with INSC partner countries on DBA, BDBA and severe accident analyses for WWER plants of generation 3 + and building NRA and safety evaluation capacities and decommissioning of nuclear facilities and disposal of radioactive waste. The results are preceded by an outline on the activities related to the project management and to the planning of the bilateral work.

  20. Construction of the Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Jicha, J.

    1989-01-01

    The Temelin nuclear power plant with four WWER-1000 reactors is designed to supply electricity in an amount of 23 TWh/yr and heat in an amount of 8000 TJ/yr in the first stage. The maximum heat extraction should be 922 MW. The plant construction includes the building of 10 buildings, the total cost being 52 thousand million Czechoslovak crowns. Another 41 investment items are associated with the plant construction. The most important of them include constructions for leading out the electric power, for standby electricity supply for the power plant, and for the extraction of heat from the plant and its supply to the town of Ceske Budejovice. The first unit should be started up for test performance in November 1992, the second in 1994 and the whole power plant should be complete by 1998. The state of the construction by February 1989 is described in detail. Attention is also paid to the preparatory activity for the operation and to social welfare of the personnel. (Z.M.)

  1. Analysis of the State of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Bergunker, Olga

    2008-01-01

    The problem of safe operation of SG heat exchanging tubes, of both economical and effective control of their state is still important these days. Issues connected with peculiarities of methods of SG tubes inspection, automated analysis of the inspection results, tubes state analysis and development of algorithms of forecasting their state are considered in this report. The need for effective use of extensive data arrays on SG operation has led to the necessity of creating software tools for collection, storage and analysis of these data. The data-analytical system 'NPP Steam Generators' meant for data systematization and visualization as well as various types of analyses of data on eddy current inspection of WWER-440 and WWER-1000 SG tubes is presented in this report. The main possibilities of the data-analytical system (DAS), the code current state and prospects of its development are shown. The main fields of DAS application are considered and some results of its practical use are mentioned, namely, in the field of forecasting SG tubes state. (authors)

  2. Fundamental principles of failed fuel detection concepts on nuclear power units of WWER type

    International Nuclear Information System (INIS)

    Lusanova, L.; Miglo, V.; Slavyagin, P.

    2001-01-01

    The subject of the paper is the Russian failed fuel detection concept in both operating and shut down reactors. The philosophy for detection of fission products released from defective fuel during operation and sipping tests and using of these results for regulation of the radiological situation at the NPP during the next cycle is widely spread. In presented work such philosophy is applied to the shut down rectors. An option for sipping test performed in a mast of Refueling Machine (RM) using a wet-gas version of sipping test is briefly described. The use of the FFD method in RM mast allows combining the procedure of Fuel Assemblies (FA) tightness test with transport operation during reloading of the fuel from the core into the cooling pool. This reduces the time for reloading and transport operation with FA and increases the safety of reactor operation. The FFD method in RM mast has passed successful tests on Unit 4 at Balakovskaja NPP and it is expected to apply in other NPP unit with WWER-1000 reactors

  3. Modernized accurate methods for processing of in-core measurement signals in WWER reactors

    International Nuclear Information System (INIS)

    Polak, T.

    1996-01-01

    Utilization of the new accurate WIMS-KAERI library (WIMKAL-88) to generate the following characteristics for Rhodium SPND: Sensitivity depletion law by high (approx= 75%) burnup of emitter; influence of burnup-history on depletion law course; influence of neutron spectrum change on Rh-SPND sensitivity caused by change of fuel enrichment, fuel burnup, moderator temperature, concentration of boracid, central pin power rate and concentration of Xe 135 ; generating and experimental testing of Rh-SPND signal to linear pin power rate and signal to neutron flux conversion factors. Rh-SPND instrumentation optimization (reduction) related to safety and operational aspects as needed for 3D power surveillance in WWER-1000 reactors. Analysis of SPND reduction from 64x7 to 46x7 by method of Shannon information entropy optimization. Influence of reduction on accuracy of 3D power distribution reconstruction. Physical methods of 3D power distribution unfolding in new modernized on-line I and C system in NPP J. Bohunice with in-core measurements according to 210 thermocouples and 36x7 Rh-SPNDs. Program system TOPRE under QNX operating system network in FORTRAN 77, neutronic background calculations by macrocode MOBY-DICK. (author). 10 refs, 6 figs, 7 tabs

  4. Alternative method of manufacture and assembly of concrete reinforcement for cylindrical part of containment

    International Nuclear Information System (INIS)

    Chalus, Z.

    1982-01-01

    The problems are discussed of the reinforcement for the cylindrical part of the containment of the standard WWER 1000 nuclear power plant in the Temelin locality. Soviet 25 G 2 S bars 40 mm in diameter and Czechoslovak bars 39 mm in diameter of steels 10 216 and 10 425 are to be used. The possible technologies of laying are presented and as an alternative solution the concept of reinforcing cages is discussed. The production of cages is considered as an on-site project. The cages would be manufactured using horizontal or vertical fixtures and stored vertically close together in special stands. Labour consumption for the production of both types of reinforcing cages is discussed and a survey presented of the necessary equipment when using the alternative technologies. Labour consumption for the assembly of the cages is also given, this for the variant when the reinforcement is fixed to the assembled steel lining, i.e., either cages 12 by 12 m in dimension (inner, outer, including transverse reinforcement) or as spatial components, baskets, with regard to the possibility of transporting elements 3 by 12 m in dimension. (B.S.)

  5. Influence of the number of energy groups on the accuracy of neutron fluence calculations

    International Nuclear Information System (INIS)

    Barz, H.U.; Konheiser, J.

    1999-01-01

    The question how many groups are necessary to obtain all needed integral quantities for the neutron load of pressure vessels and detector positions outside the vessel with sufficient accuracy is of general interest. Until now, there are no systematic investigations on this question. In principle 3-dimensional consideration is required for such neutron load calculations. Therefore, an estimation of the needed number of groups can be of interest to minimize calculation time. One general problem is the P L -approximation of the angular distributions for the transfers between different groups. For elastic scattering this P L -approximation becomes poorer with increasing number of groups. As deterministic methods generally use the P L -approximation they cannot be used for investigations of the errors caused by the group approximation. We have investigated this problem applying group Monte-Carlo but nearly exact representation of this elastic slowing down without P L -approximation. The calculations were directed to assess the neutron fluence of a Russian WWER-1000 reactor. For that a simplified geometrical model of this reactor type has been used. (orig.)

  6. Hardening and stress relaxation during repeated heating of 15Kh2MFA and 15Kh2NMFA steels welded joints

    International Nuclear Information System (INIS)

    Zubchenko, A.S.; Suslova, E.A.

    1986-01-01

    Results of investigation of temperature-time conditions of hardening of welded joints of 15Kh2MFA and 15Kh2NMFA steels and their relaxation resistance, effect of metal structure of imitated heat affected zone (HAZ) on intensity of precipitation hardening at repeated heating are presented as well as the results of the process of relaxation of residual stresses at welded joints samples heating carried out by automatic welding under the flux with the use of adding materials and technology of manufacturing of vessels of WWER-440 and WWER-1000 reactors. Peculiarities of the hardening at repeated heating of the HAZ metal imitated at these steels. Precipitation hardening of overheated 15Kh2MFA steel is connected with precipitations at repeated heating of carbides of the M 7 C 3 , M 3 C and VC type. Stress relaxation in welded joints runs more intensively at the initial stage of repeated heating, i.e. during the same period of the process of dispersed carbide precipitations

  7. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  8. Benchmark study for the seismic analysis and testing of WWER type NPPs. Final report of a co-ordinated research project

    International Nuclear Information System (INIS)

    2000-10-01

    This report is the final results of a five-year IAEA Coordinated Research Project (CRP) launched in 1992. The main goal was the harmonization of criteria and methods used in Member States in seismic reassessment and upgrading of existing nuclear power plants. To this aim, most of the activities have been focused on a benchmarking exercise related to a mixed numerical and experimental dynamic analysis carried out on two reference units of WWER reactors (WWER-1000 and WWER-440/213): Kozloduy NPP Units 5/6 and PAKS NPP. Twenty-four institutions from 13 countries participated in the CRP, and two other institutions from Japan contributed to the CRP informally. The objective of this TECDOC is to provide a consistent and comprehensive summary of the results of the work performed in the CRP through the preparation of a ''self-standing'' report with the general conclusion of the programme: a great deal of information from the Background Documents has been included in this report with a set of recommendations for future work in this field

  9. Low level radioactive waste disposal in Kozloduy NPP in Bulgaria

    International Nuclear Information System (INIS)

    Stanchev, V.

    2001-01-01

    Kozloduy NPP is the biggest power plant in the Republic of Bulgaria. It is in operation since 1974 and for the past 25 years it has generated over 263 billion kWh electric power. The NPP share in the total electric production in 1998 was about 50%. It has six units in operation - four WWER 440 B-230 and two WWER 1000 B-320. In the nuclear reactor operation the generation of radioactive waste (RAW) is an inevitable process. The waste must be conditioned, stored and disposed of in a safe manner. There are no national radioactive waste disposal facilities, for waste generated by an NPP, in Bulgaria to the moment. This situation necessitates the storage of operational RAW to be carried out on site for a long period of time (30 to 50 years). Following the principle for protection of human health and environment now and in the future, Kozloduy NPP adopted the concept for conditioning the RAW to a stable solid form and placing the waste in a package which should keep its features for a sufficiently long term so that the package can be safely transported to the disposal site. (author)

  10. Nuclear safety assessment of nuclear power plants and nuclear risk in Eastern Europe and other regions. Scientific-technical cooperation with nuclear regulatory authorities and technical support organizations (TSOs); Einschaetzung der nuklearen Sicherheit von Kernkraftwerken sowie nuklearer Risiken in Osteuropa und anderen Regionen. Wissenschaftlich-technische Zusammenarbeit mit atomrechtlichen Behoerden und deren Sachverstaendigenorganisationen (TSO)

    Energy Technology Data Exchange (ETDEWEB)

    Wolff, Holger

    2014-09-15

    The BMUB/BfS project 3611I01512 formed the frame of the GRS for the scientifictechnical cooperation with Technical Support Organisations and Nuclear Regulatory Authorities in the field of nuclear safety of NPPs and for the evaluation of nuclear risks in Eastern Europe and other regions for the period from September 2011 till June 2014. In the present final report main results of the project are described. The project comprised the following technical topics: - Record status of NPP modernisation programs, Monitoring and evaluation of modernisation programs; - Design basis and severe accident analyses for NPP with PWR (WWER-440, WWER-1000); - Cooperation with INSC partner countries on DBA, BDBA and severe accident analyses for WWER plants of generation 3+ and on building NRA and safety evaluation capacities; - Decommissioning of nuclear facilities and disposal of radioactive waste; - Evaluation of new reactor concepts and their safety design; Panels at regulatory level. The work results are preceded by a summary on the activities related to the project management and to the planning of the bilateral work.

  11. Development of four-year fuel cycle based on the advanced fuel assembly with uranium-gadolinium fuel and its implementation to the operating WWER-440 units

    International Nuclear Information System (INIS)

    Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, P.; Filimonov, P.

    2000-01-01

    Over the past few years in Russia the investigations aimed at the increase of the reliability, safety and efficiency of operation of the WWER-1000 reactors as well as of its competitiveness in the world market were carried out. In the frame of these investigations the four-year fuel cycle, based on advanced fuel assemblies with zirconium alloy spacer grids and guide tubes and with fuel pellet having a reduced diameter of the central hole (1,5 mm), has been developed. For the compensation of a part of excess reactivity, Gd 2 O 3 integrated burnable absorbers are used. CPS absorbing rods contain a combine absorber (B 4 C + Dy 2 O 3 *TiO 2 ). A part of depleted fuel is located on the core periphery. The algorithms controlling the reactor power and power distribution have been updated. For checking of the solutions adopted and for verification of code package developed at the RRC 'Kurchatov Institute' the wide-scale experimental operation of advanced FA and its individual components is carried out. (Authors)

  12. Diagnostic and monitoring systems produced in Vuje, Okruzna 5, 918 64 Trnava, Slovak Republic

    International Nuclear Information System (INIS)

    Oksa, G.; Bahna, J.; Murin, V.; Kucharek, P.; Smutny, S.

    1996-01-01

    Based on the 20 years experience in on-line vibration diagnostics of mechanical components in the primary circuit of nuclear power plant PWR WWER-440, Vuje, Okruzna 5, 918 64 Trnava produces its own diagnostic and monitoring systems since 1990. The variety of diagnostic systems includes: loose part monitoring system (LPMS), monitoring system of main circulating pumps (MCPMS), vibration monitoring system (LVMS), leakage monitoring system (LMS). The emphasis in the hardware solution is put on the design modularity and versatility so that many subcomponents (circuit boards) are common or highly similar for all systems. Using exclusively digital data for processing enhances the reliability of the measurements and allows the easy data transportation from one computer to another (e.g., for more sophisticated analysis). Trends in the software development follow the similar path as for the hardware solution - namely, the modularity and versatility of software is the imperative goal. The modern operating systems also incorporate the ability of network communication, which is crucial for the integration of stand-alone diagnostic systems into nuclear power plants information system. So far a number of systems have been successfully installed: 6 LPMSs (Jaslovske Bohunice, Dukovany), 4 MCPMs (Jaslovske Bohunice) and 2 LVMSs (Jaslovske Bohunice), all systems operate in PWR WWER-440 environment. Another diagnostic systems are under construction: 2 LPMSs (Temelin, PWR WWER-1000), 2 MCMSs (Mochovce - PWR WWER-440) and 2 LMSs (Jaslovske Bohunice). (author). 1 fig

  13. Design of NPP of new generation being constructed at Novovoronezh NPP site

    International Nuclear Information System (INIS)

    Afrov, A.; Berkovich, V.; Generalov, V.; Dragunov, Yu.; Krushelnitskij, V.

    2000-01-01

    The design of a new-generation nuclear power plant with advanced WWER-1000 units, currently under construction at Novovoronezh, is dealt with in considerable detail. Information is given on the general layout of the power plant, and the monoblock comprising the four-loop reactor plant with double containment, the turbine hall with two turbine-driven feedwater pumps and safety and auxiliary systems is described. In a separate building at the power plant site, a new fuel storage facility is located designed for extreme external effects. The operations involved in fuel handling, in both the storage facility and the reactor department, are mentioned. The management of various types of radioactive waste is highlighted. The basic principles of the engineering solutions pertaining to the plant's electrical part are outlined. The concept of the instrumentation and control system is explained. Information is also given on the double containment system. A considerable part of the presentation is devoted to the philosophy and concept of the power plant safety, to the results of safety evaluation and to the research carried out in this respect. The paper is concluded with a discussion of the experiments and analyses performed with the aim to justify the design solutions as regards the passive heat removal system. (A.K.)

  14. Research reactors spent fuel management in the Nuclear Research Institute Rez

    International Nuclear Information System (INIS)

    Rychecky, J.

    2001-01-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15

  15. Research reactors spent fuel management in the Nuclear Research Institute Rez

    Energy Technology Data Exchange (ETDEWEB)

    Rychecky, J. [Nuclear Research Institute, 25068 Rez (Czech Republic)

    2001-07-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15.

  16. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  17. Modelling and simulation of process control systems for WWER

    Energy Technology Data Exchange (ETDEWEB)

    Pangelov, N [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    A dynamic modelling method for simulation of process control system is developed (method for identification). It is based on the least squares method and highly efficient linear uninterrupted differential equations. The method has the following advantages: there are no significant limitations in the type of input/output signals and in the length of data time series; identification at none zero initial condition is possible; on-line identification is possible; a high accuracy is observed in case of noise. On the basis of real experiments and data time series simulated with known computer codes it is possible to construct highly efficient models of different systems for solving the following problems: real time simulation with high accuracy for training purposes; estimation of immeasurable parameters important to safety; malfunction diagnostics based on plant dynamics; prediction of dynamic behaviour; control vector estimation in regime adviser. Two real applications of this method are described: in dynamic behaviour modelling for steam generator level, and in creating of a Process Control System Simulator (PCSS) based on KASKAD-2 for WWER-1000 units of the Kozloduy NPP. 6 refs., 8 figs.

  18. About reliability of WWER pressure vessel neutron fluence calculation

    Energy Technology Data Exchange (ETDEWEB)

    Belousov, S; Ilieva, K; Antonov, S [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs.

  19. Nuclear fuel utilization at the Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, Z [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Kharalampieva, Ts; Pejchinov, Ts

    1996-12-31

    Data on core loading and operation conditions during past fuel cycles of the Units 1 to 6 at the Kozloduy NPP are presented. The Units 1 and 2 have reached average discharge fuel burn-up of 31 MW d/kg U, the Unit 3 - 34 MW d/kg U and the Unit 4 - 36.5 MW d/kg U. By use of dummies and low-leakage core loading patterns for WWER-440 cores an effective reduction in reactor pressure vessel irradiation is obtained. By increasing the enrichment level and improving the characteristics of the Units 3 and 4, a design fuel cycles duration has been reduced by 5-10% in number of assemblies. Core loading design has been modelled using computer codes SPPS-1, BIPR-7, ALBOM, PROROC. A 3-year fuel cycle utilizing 4.4% enriched fuel proved to be more efficient for WWER-1000 by 15% reduction in fuel cost compared to the 2-year cycle. Future developments include improvements of the in-core monitoring system and process on-line simulation based on more accurate computer codes. 7 refs., 6 figs., 10 tabs.

  20. Nuclear fuel for WWER reactors. Current status and prospects

    International Nuclear Information System (INIS)

    Molchanov, V.

    2006-01-01

    In this paper the following results from the JSC TVEL post-irradiation studies of full-scale fuel assemblies are discussed: 1) Oxide layer on fuel rod (FR) cladding does not exceed 15 μm; 2) Fission Gas Release does not exceed 3%; 3) Satisfactory condition of cladding. Concerning the operating experience it is noticed that: 1) Design criteria are fulfilled; 2) 5 TVSA are stayed at Kalinin-1 for 6 years up to burnup of 59 MWxdays/kgU; 3) 12 working assemblies (WA) are operated at Kola-3 for 6 years up to burnup of 57 MWxdays/kgU. At the end the following conclusions are made: 1) The main task of further development of nuclear fuel for WWER-1000 reactors is the increasing fuel rod burnup up to 72 MWxdays/kgU and operating life up to 6 years through the increase in uranium load in FA and maximum use of potential built-in in the designs of TVSA and TVS-2; 2) In the nearest future R and D in the field of nuclear fuel for WWER will be directed at justification of maneuvering fuel characteristics as well as at introducing of optimized zirconium alloys

  1. New Solutions For Increasing Environmental Protection During Severe Accidents At Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kulyukhin, Sergei A.; Mikheev, Nikolai B.; Falkovskii, Leo N.; Reshetov, Leo A.; Zvetkova, Marianna Ya.; Yagodkin, Ivan V.; Osipov, Viktor P.; Skvortsov, Sergei S.; Berkovich, Viktor M.; Taranov, Gennadii S.; Grigor'ev, Mikhail M.; Meshkov, Vladimir M.; Noskov, Andrei A.; Mitrofanov, Mikhail I.

    2008-01-01

    This paper reports new solutions for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the inter-containment area, which includes the 'wet' stage (scrubbers, etc.), the 'dry' stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable. (authors)

  2. Main Results of Updated Decommission Conception of NPPs Operating in Ukraine

    International Nuclear Information System (INIS)

    Purtov, Oleg; Masko, Alexander; Vasilchenko, Victor

    2014-01-01

    Results of long-term planning analysis based on consideration of 6 possible scenarios for the nuclear energy development with 15 years and 20 years life time extensions of operation of nuclear power units beyond 30 year provided by original design are presented in the updated decommission conception of NPP's operating in Ukraine. These characteristics of the two main options for NPP decommissioning deferred or immediate dismantling, which is close to the level of acceptability with relative superiority variant of deferred dismantling, are presented. The best option for NPP unit decommissioning as comparative analysis results is the option with deferred dismantling with 30 years endurance time. It can be taken as a basis for optimal strategies for NPP unit decommission design development. Cost estimations for the decommissioning of WWER-440 and WWER-1000 reactor type units are presented in the updated conception. The updated cost assessment for required annual payments with uniform accumulation costs to the Decommission Fund corresponding deferred dismantling variant with 20 years life time extension operation time is 98,2 mln US$ per year. This value is 3.61% of the electricity generated by NPP's in Ukraine and supplied to the wholesale electricity market of Ukraine in 2012 base year. (authors)

  3. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1995-01-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

  4. Technical and economical prerequisites of special district-heating nuclear power plant development

    International Nuclear Information System (INIS)

    Baturov, B.B.; Boldyrev, V.M.; Losev, V.L.; Sigal, M.V.

    1983-01-01

    Results are presented of technical and economical analysis of advisability of constructing combined Nuclear Power and Heating Plants (NPHP) assuring the possibility of their location near the areas of heat power consumption in case of observing a given degree of radiation safety for population and personnel. Specific features determining the choice of turbine-driven units for such plants are analyzed. Conditions of competiveness of a specialized NPHP with alternative power units, NPHP based on the WWER-1000 reactor and district heating plants (NDHP), are determined. Analysis of design specifications of NPHP with two VK-500 reactor units and structures of capital investments in such a plant reveal that an increase in the total capital investments in the NPHP would not exceed 2% with account of the difference in costs for grid heaters and a corresponding change in dimensions of operation rooms as well as changes in costs of heat removal system (within one site) at the TK turbine replacement by the T turbine

  5. Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

    1996-01-01

    In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

  6. Heat extraction from turbines of Czechoslovak nuclear power plants for district heating

    International Nuclear Information System (INIS)

    Drahy, J.

    1985-01-01

    Two design are described of SKODA extraction turbines for Czechoslovak nuclear power plants with WWER-440 and WWER-1000 reactors. 220 MW steam turbines were originally designed as pure condensation turbines with uncontrolled steam extraction. Optimal ways are now being sought for their use for heating hot water for district heating. For district heating of the town of Trnava, the nuclear power plant at Jaslovske Bohunice will provide a two-step heating of water from 70 to 120 degC with a heat supply of 60 MW th from one turbine unit. The ratio of obtained heat power to lost electric power is 5.08. Investigations showed the possibility of extracting 85 MW th of heat from uncontrolled steam extraction, this at three-step water heating from 60 to 145 degC, the ratio of gained and lost power being 7.14. Information is presented on the SKODA 220 MW turbine with steam extraction for heat supply purposes and on the 1000 MW turbine with 893 MW th heat extraction. The specifications of both types are given. (Pu)

  7. Two-dimensional full-core transport theory Benchmarks for the WWER reactors

    International Nuclear Information System (INIS)

    Petkov, P.T.

    2002-01-01

    Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

  8. New Solutions For Increasing Environmental Protection During Severe Accidents At Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kulyukhin, Sergei A.; Mikheev, Nikolai B. [Institute of Physical Chemistry and Electrochemistry, Russian Academy of Sciences, Moscow (Russian Federation); Falkovskii, Leo N.; Reshetov, Leo A.; Zvetkova, Marianna Ya. [All-Russian Research Institute of Atomic Machine-Building, Moscow, Russia (Russian Federation); Yagodkin, Ivan V.; Osipov, Viktor P.; Skvortsov, Sergei S. [Institute of Physics and Power Engineering, Obninsk (Russian Federation); Berkovich, Viktor M.; Taranov, Gennadii S.; Grigor' ev, Mikhail M. [Institute ' Atomenergoproekt' , Moscow (Russian Federation); Meshkov, Vladimir M.; Noskov, Andrei A.; Mitrofanov, Mikhail I. [ROSENERGOATOM Concern, Moscow (Russian Federation)

    2008-07-01

    This paper reports new solutions for increasing environmental protection during severe accidents at NPPs. For NPPs with two protective shells and pressure release system such as WWER-1000 we suggest a new comprehensive, passive-mode environmental protection system of decontamination of the radioactive air-steam mixture from the containment and the inter-containment area, which includes the 'wet' stage (scrubbers, etc.), the 'dry' stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For Russian WWER-440/230 NPPs we suggest three protection levels: 1) a jet-vortex condenser; 2) the spray system; 3) a sorption module. For modern designs of new generation NPPs, which do not provide for pressure release systems, we proposed a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable. (authors)

  9. Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

    International Nuclear Information System (INIS)

    Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

    2013-01-01

    The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

  10. Impact of extreme load requirements and quality assurance on nuclear power plant costs

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1993-01-01

    Definitive costs, applicable to nuclear power plant concrete structures, as a function of National Regulatory Requirements, standardization, the effect of extreme load design associated with both design basis accidents and extreme external events and quality assurance are difficult to develop since such effects are interrelated and not only differ widely from country to country, project to project but also vary in time. Table 1 shows an estimate of the of the overall plant cost effects of external event extreme load design on nuclear power plant design for the U.S -and selected foreign countries for which experience with LWRs exist- Germany is the most expensive primarily due to a military aircraft crash resistance. However, the German requirement for 4 safeguards trains rather than 2 and the containment design requirement to consider one Steam Generator blowdown concurrent with a RCS blowdown. This presentation will concentrate on the direct current impact extreme load design and quality assurance have on concrete structures, systems and components for nuclear plants. This presentation is considered timely due to the increased interest in the c potential backfit of Eastern European nuclear power stations of the WWER 440 and WWER 1000 types which typically did not consider the extreme loads identified in Table 1 and accident loads in Table 3 and quality assurance in Table 5 in their original design. Concrete structures in particular are highlighted because they typically form the last barrier to radioactive release from the containment and other Safety Related Structures

  11. About reliability of WWER pressure vessel neutron fluence calculation

    International Nuclear Information System (INIS)

    Belousov, S.; Ilieva, K.; Antonov, S.

    1995-01-01

    This reliability study was carried out under a Research Contracts F111 and TH-324 of the Bulgarian Ministry of Higher Education and the IAEA. The effect of geometry approximation and the choice of neutron cross-sections on the calculation model is estimated. The neutron flux onto reactor pressure vessel at locations, important for metal embrittlement surveillance, has been calculated using the codes TORT and DORT. The flux values calculated for both WWER-440 and WWER-1000 show good consistency within the limits of solution accuracy. It is concluded that the synthesis method (DORT) can be used for calculations at a reasonable cost whenever metal embrittlement surveillance is considered. Using an iron sphere benchmark measurement, a comparison of an experimental leakage spectrum with spectrum calculated using multigroup neutron cross-sections based on ENDF/B-4 and ENDF/B-6 data is performed. In the energy region above 1 MeV the best agreement with the experiment is achieved for ENDF/B-6 in VITAMIN-E group structure. 7 refs., 1 fig., 4 tabs

  12. Gaseous swelling of B4C and UO2 fuel: similarities and differences

    International Nuclear Information System (INIS)

    Evdokimov, I.; Khoruzhii, O.; Kourtchatov, S.; Likhanskii, V.; Matweev, L.

    2001-01-01

    A major factor limiting the resource of control rods (CRs) for WWER-1000 reactors is their radiation damage. Radiation induced embrittlement of the CRs cladding, core swelling and gaseous internal pressure in CRs result in mechanical core-cladding interaction. This work is devoted to the physical analysis of processes that control the structural changes in neutron absorber elements with B 4 C under irradiation in water reactors. Particularly, the analysis of mechanisms of the helium porosity formation in B 4 C is undertaken. In view of the deficiency of experimental data on the subject, a fruitful approach to the problem is a comparative analysis of the swelling mechanisms in B 4 C absorber and UO 2 fuel. Using this similarity a phenomenological model of fission gas behavior in boron carbide is proposed. The model predictions for radial profile of 10 B burnup under influence of thermal and epithermal neutrons are compared with experimental results. The main results show that despite the external similarity of the process of fission gas accumulation in UO 2 and in B 4 C, phenomenology of gaseous swelling is much different for the fuel and the CR core. The reason for that difference is the distinction of physical conditions in irradiated fuel and CR core

  13. Long-term prospects of capital demands of nuclear power sources construction in Czechoslovakia

    International Nuclear Information System (INIS)

    Sladek, V.; Kysel, J.

    1988-01-01

    The capital demands are compared for four variants of nuclear power plant construction projects to be implemented in Czechoslovakia by the year 2030. By the year 2000, all variants are expected to achieve the total installed capacity of 11,280 MW. Construction in the following period then differs: by the year 2030 total installed nuclear capacity is to reach the following values: 1st variant - 30,000 MW; 2nd variant - 35,000 MW, 3rd variant - 36,500 MW and 4th variant - 27,760 MW. WWER-1000, WWER-1500 and WWER-500 reactors are expected to be installed in the variant projects which differ in speed of construction and life of the units. They are designed such as to provide for the smooth transfer of building workers from one site to the next. The comparison of capital demands shows that a suitable siting strategy and site preparation can save costs and preclude serious sociological problems. The possible extension of the life of nuclear power plants will provide for a better use of primary investments and improve the economic efficiency of nuclear power plants. (Z.M.). 2 figs., 4 tabs., 6 refs

  14. Different approaches to estimation of reactor pressure vessel material embrittlement

    Directory of Open Access Journals (Sweden)

    V. M. Revka

    2013-03-01

    Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

  15. General structure and functions of the OPAL optimization system

    International Nuclear Information System (INIS)

    Mikolas, P.; Sustek, J.; Svarny, J.

    2005-01-01

    Presented version of OPAL - the in core fuel management system is under development also for core loading optimization of NPP Temelin (WWER-1000 type reactor). Description of the algorithm of separate modules was presented in several AER papers. The optimization process of NPP Temelin loading patterns comprises problems like preparation input data for NPP SW, loading searching, fixing and splitting of fuel enrichments, BP-assignment, FA rotation and fuel cycle economics. In application for NPP Temelin has been used NPP Temelin code system (spectral code with macrocode). The objective of fuel management is to design a fuel-loading scheme that is capable of producing the required energy at the minimum cost while satisfying the safety constraints. Usually the objectives are: a) To meet the energy production requirements (loaded fuel should have sufficient reactivity that covers reactivity defects associated with startup as well as reactivity loss due to the fuel depletion); b) To satisfy all safety-related limits (loaded fuel should preserve adequate power peaking limits (given namely LOCA), shutdown margins and no positive Moderator Temperature Coefficient (MTC); c) To minimize the power generation cost ($/kWh(e)). Flow of optimization process OPAL management system is in detail described and application for NPP Temelin cores optimization presented (Authors)

  16. Criticality safety analysis of TK-13 cask in Bushehr nuclear power plant

    International Nuclear Information System (INIS)

    Mohammadi, Ashgar; Omidvari, Nima; Hassanzadeh, Mostafa

    2017-01-01

    Spent fuel production is one of the main problems of nuclear power plants that should be managed properly considering the strategy of each country. Today, in most of nuclear power owner countries, the interim storage has been selected as the temporary solution of spent fuel management because of absence of deep geological repositories and no tendency for reprocessing. On the other side, considering the merits of storage in dual purpose casks based on dry storage, this method was chosen for interim storage. By taking into account that the only operating reactor of Iran is of Water-Water Energetic Reactor (WWER)-1000 type, proposed TK-13 cask by Russia which is the manufacturer of these types of reactors has been considered. In this study, the calculation of basket holding spent fuel assembly criticality of this cask has been analyzed for two modes of fresh and spent fuel by ORIGEN2.1 and MCNPX2.6 nuclear codes. The criterion of the nuclear criticality safety for effective multiplication factor (k eff ) should be 0.95 and 0.98 for many ordinary and accident conditions, respectively. Therefore, the results show that a cylindrical basket with 66 cm diameter and 28 cm pitch with internal holding basket made of borated steel with 0.1% borate and steel free from borate would meet the criticality of cask, respectively.

  17. Scientific-technical cooperation with foreign (esp. Europe and INSC partner countries) nuclear regulatory authorities and their technical support organizations in the fields of nuclear safety of operating nuclear power plants and on the concept evaluation of generation 3+ plants. Final report; Wissenschaftlich-Technische Zusammenarbeit (WTZ) mit auslaendischen (insbesondere in Europa und INSC-Partnerstaaten) atomrechtlichen Behoerden und deren Sachverstaendigenorganisationen zur nuklearen Sicherheit in Betrieb befindlicher Kernkraftwerke und zur Konzeptbewertung von Generation-3+-Anlagen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Wolff, Holger

    2016-09-15

    The BMUB/BfS-Project 3614I01512 forms the frame of the GRS for the scientific-technical cooperation with Technical Support Organisations and Nuclear Regulatory Authorities in the field of nuclear safety in operating NPPs and for the concept evaluation of generation 3{sup +} plants in Europe and INSC Partner Countries. In the present final project report results are described which were gained within the project duration 15.10.2014 up to the 30.09.2016 in the following working packages: Investigations following the catastrophe of Fukushima Daiichi, Evaluation of selected National Action Plans, DBA and severe accident analyses for NPP with PWR (WWER-440, WWER-1000), cooperation with INSC partner countries on DBA, BDBA and severe accident analyses for WWER plants of generation 3{sup +} and building NRA and safety evaluation capacities and decommissioning of nuclear facilities and disposal of radioactive waste. The results are preceded by an outline on the activities related to the project management and to the planning of the bilateral work.

  18. Radiation damage in a high Ni weld

    International Nuclear Information System (INIS)

    Brumovsky, M.; Kytka, M.; Kopriva, R.

    2015-01-01

    WWER-1000 RPV weld metals are characterized by a high content of nickel, mostly about 1.7 mass % with content of manganese around 0.8 mass % with a very low copper content - about 0.05 mass %. In such material some late blooming phase effect should be observed during irradiation. Such typical weld material was irradiated in the experimental reactor LVR-15 in N RI Rez at the irradiation temperature 290 C degrees and at five neutron fluences from 1.5 to 9.5 *10 23 m -2 (E>1 MeV). Charpy V-notch impact tests, static fracture toughness tests, tensile and hardness measurement were performed to obtain effect of neutron fluence on radiation hardening as well as embrittlement. Neutron fluence dependences of all these property changes have monotonic character but with a high neutron embrittlement exponent around 0.8. Scanning electron microscope of fracture surfaces showed no or very small portion of intercrystalline fracture. Transmission electron microscopy was performed on specimens from all neutron fluences. Only low density of black-dot damage has been observed. It is assumed that most of defect are dislocation loops. The late blooming phase which may be observed from results of mechanical properties are probably below the resolution of the used JEM-2010, i.e. 1.5 nm. (authors)

  19. Operator support system for power unit control in abnormal modes of operation

    International Nuclear Information System (INIS)

    Kurka, J.

    1993-01-01

    I and C system technology, partly Soviet and partly Czechoslovakian, used on the NPP Dukovany units represents the control technology standard of late 70-ties and it becomes the weak part of the whole system. The modernization of the system, therefore, is necessary and it is already in preparation. The specification of both the scope and the depth of upgrading/replacement is being carried out within the framework of the PHARE program. The second phase of the program aimed at the final specifications of requirements on new I and C system is in progress. The output will serve as detailed specification for bid invitation for control system supplier. Parallely, the preparation of specification for WWER-440 full-scope plant simulator for operator training is in progress as well. In the case of two units with WWER-1000 MW reactors, the completion of construction of which was even threaten for a certain period of time, essential changes have taken place in the design of both the I and C systems and the reactor core. 7 figs

  20. Probabilistic analysis of strength and thermal-physic WWER fuel rod characteristics using START-3 code

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Khramtsov; Sokolov, F.

    2001-01-01

    During the last years probabilistic methods for evaluation of the influence of the fuel geometry and technology parameters on fuel operational reliability are widely used. In the present work the START-3 procedure is used to calculate the thermal physics and strength characteristics of WWER fuel rods behavior. The procedure is based on the Monte-Carlo method with the application of Sobol quasi-random sequences. This technique allows to treat the fuel rod technological and operating parameters as well as its strength and thermal physics characteristics as random variables. The work deals with a series of WWER-1000 fuel rod statistical tests and verification based on the PIE results. Also preliminary calculations are implemented with the aim to determine the design schema parameters. This should ensure the accuracy of the assessment of the parameters of WWER fuel rod characteristics distribution. The probability characteristics of fuel rod strength and thermal physics are assessed via the statistical analysis of the results of probability calculations

  1. Power supply for nuclear power plant home consumption using inverter sources

    International Nuclear Information System (INIS)

    Stepar, V.; Tvrdik, J.

    1988-01-01

    A description is presented of the system configuration of a nuclear power plant with WWER-440 and WWER-1000 units and of a system of guaranteed power supply for the plant home consumption network for a WWER-440 power plant. The power supply network consists of 6 inverter supplies for safety systems, 4 supplies for nonsystem facilities and of a minimum of three power supplies for facilities outside the units. A diagram is shown and the principle described of the currently used ABP 1500 power supply system of Soviet origin. Using the experience from its operation, a Czechoslovak inverter supply system was designed featuring parallel operation of the inverter units. The principles are given for the design of a power supply system of an output of 75 kVA, the diagram is shown and the principle described. The design allows implementing and using the system without redundancy or in redundant operation of the 2 from 3 type at an installed capacity of 66%, or of the 1 from 2 type at an installed capacity of 50% of the output. The system is undergoing tests that should verify its reliability. (J.B.). 3 figs., 1 tab

  2. Estimation of RPV material embrittlement for Ukrainian NPP based on surveillance test data

    International Nuclear Information System (INIS)

    Revka, V.; Chyrko, L.; Chaikovsky, Yu.; Gulchuk, Yu.

    2012-01-01

    The WWER-1000 RPV material embrittlement has been evaluated using the surveillance test data for the nuclear power plant which is under operation in Ukraine. The RPV materials after the neutron (E > 0,5 MeV) irradiation up to fluence of 22,9.10 22 m -2 have been studied. Fracture toughness tests were performed using pre-cracked Charpy specimens for the beltline materials (base and weld metal). The maximum shift of T 0 reference temperature is equal to 44 o C. A radiation embrittlement rate, A F , for the RPV materials was estimated using the standard and reconstituted specimens. A comparison of the A F values has shown a good agreement between the specimen sets before and after reconstitution both for base and weld metal. Furthermore it has been revealed there is no nickel effect for the studied materials. In spite of the high nickel content the radiation embrittlement rate for weld metal is not higher than for base metal with low nickel content. Fracture toughness analysis has shown the Master curve shape describes well a temperature dependence of K Jc values. However a higher scatter of K Jc values is observed in comparison to 95 % tolerance bounds. (author)

  3. Vibrations as a contributor to the cracking of PGV-1000 steam generator cold collector

    International Nuclear Information System (INIS)

    Verezemskij, V.G.

    1997-01-01

    The results of multiple investigations of cold collector ruptures at NPPs with WWER-1000 reactor as a complex and multi-parameter process are generalized. It is shown that the temperature of cold collector operation (280-290 deg C) at which environment corrosion effects are maximum has an important role for revealing the factors causing the damage. When the reactor plant operates under stationary and transient conditions the primary coolant circuit equipment, main circulation pipelines and main steam lines become involved into vibrations with different intensities as a result of pressure pulsations at reverse and multiple frequencies of the main circulation pumps connected with formation of standing pressure waves in the primary circuit and steam lines. The analysis made gives an opportunity to conclude that dynamic loads (vibrations) play the leading part in reaching the limits of cold collector metal cyclic strength and its cracking. It means that the measures for cold collector service life increasing should be directed on vibration amplitude lowering and cyclic stress decreasing

  4. Response of the main building, unit 5, Kozloduy NPP (Bulgaria) to an explosion input motion

    International Nuclear Information System (INIS)

    Kostov, M.; Boncheva, H.; Varbanov, G.; Kaneva, A.; Stefanov, D.; Koleva, N.

    1999-01-01

    The main building of Unit 5, NPP Kozloduy is one of the two selected prototypes of WWER-type plants included in the project 'Benchmark study for seismic analysis and testing of WWER-type nuclear power plants', co-ordinated by IAEA. The dynamic in-situ testing of the structure of Unit 5 WWER-1000 MW reactor containment building and the adjacent auxiliary structures is performed by specialists of ISMES (Italy), assisted by Hungarian consultants (GEOPARD) and Bulgarian specialist of NPP Kozloduy. The aim of the experiment is to assess the dynamic characteristics and behaviour of the structures, the soil-structure interaction effects, the equipment-structure interaction, etc. The dynamic excitation is a low level ground motion induced by underground explosions. The measuring of the response is done at different pre-selected structure points. Two experiments are performed. The registrations during the second experiment of 01.07.1996 are used in the present report. One set of free field records are given up to the participant institutions for the benchmarking. The purpose is to perform a dynamic analysis of the main building of Unit 5, to assess the response to the explosion ground motion and to compare the analytical results with the recorded motion. In this report are discussed soil and structure modelling and computation results consisting in acceleration response spectra obtained at foundation level and at different points of the structure

  5. Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2001-01-01

    This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

  6. Validation of thermal hydraulic computer codes for advanced light water reactor

    International Nuclear Information System (INIS)

    Macek, J.

    2001-01-01

    The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

  7. Hybrid district heating system with heat supply from nuclear source

    International Nuclear Information System (INIS)

    Havelka, Z.; Petrovsky, I.

    1987-01-01

    Several designs are described of heat supply from large remote power sources (e.g., WWER-1000 nuclear power plants with a 1000 MW turbine) to localities where mainly steam distribution networks have been built but only some or none networks for hot water distribution. The benefits of the designs stem from the fact that they do not require the conversion of the local steam distribution system to a hot water system. They are based on heat supply from the nuclear power plant to the consumer area in hot water of a temperature of 150 degC to 200 degC. Part of the hot water heat will be used for the production of low-pressure steam which will be compressed using heat pumps (steam compressors) to achieve the desired steam distribution network specifications. Water of lower temperature can be used in the hot water network. The hot water feeder forms an automatic pressure safety barrier in heat supply of heating or technological steam from a nuclear installation. (Z.M.). 5 figs., 9 refs

  8. Method of determination of thermo-acoustic coolant instability boundaries in reactor core at NPPs with WWER

    International Nuclear Information System (INIS)

    Skalozubov, Volodymyr; Kolykhanov, Viktor; Kovryzhkin, Yuriy

    2007-01-01

    The regulatory body of Ukraine, the National Atomic Energy Company and the Scientific and Production Centre have led team-works concerned with previously unstudied factors or phenomena affecting reactor safety. As a result it is determined that the thermo-acoustic coolant instability conditions can appear in the core at definite operating WWER regimes. Considerable cyclic dynamic loads affect fuel claddings over thermo-acoustic pressure oscillations. These loads can result in inadmissible cassette design damage and containment damage. Taking into account calculation and experimental research authors submit a method of on-line assessment of WWER core state concerning thermo-acoustic coolant instability. According to this method, the thermo-acoustic coolant instability appearance conditions can be estimated using normal registered parameters (pressure, temperature, heat demand etc.). At operative modes, a WWER-1000 core is stable to tracheotomies oscillations, but reduction of coolant discharge through the core for some times can result in thermo-acoustic coolant instability. Thermo-acoustic instability appears at separate transitional modes concerned with reactor scram and unloading/loading at all power units. When thermo-acoustic instability begins in transitional modes, core elements are under influence of high-frequency coolant pressure pulsations for a long time (tens of hours)

  9. Some aspects of nuclear power development in Czechoslovakia

    International Nuclear Information System (INIS)

    Simandl, S.; Stefec, V.

    1986-01-01

    Some technical and economic aspects of the development of nuclear power in Czechoslovakia are discussed. Specific conditions include the high population density of the territory and related factors as well as the shortage of raw materials for the construction of technological equipment and for the construction of buildings. It is stated that projects of future nuclear power plants should use reserves, such as are available in the standard project of a WWER-1000 nuclear power plant now being used in Temelin, as against foreign nuclear power plants. This mainly includes a bigger built-up area per installed megawatt of power and a greater number of personnel for operation and maintenance. Also discussed is the world-wide growth of capital costs, of construction time and of the number of workers needed for the construction of nuclear power plants. With the exception of the V-2 nuclear power plant the construction time of Czechoslovak nuclear power plants does not exceed world average. The maximum number of workers required for construction, however, compares unfavourably with advanced capitalist countries. Operating costs in Czechoslovakia are more favourable for nuclear power plants than for coal burning power plants and pumped-storage hydroelectric plants. (Z.M.)

  10. Complex program of advance in science and technology

    International Nuclear Information System (INIS)

    Sychev, V.V.

    1986-01-01

    A draft of the complex program of advance in science and technology of the CMEA member-countries is described in brief. The basis of the program includes five priority trends electronics development complex automatization, advanced development of nuclear energy, production of new materials and tecnologies of their production and processing, advanced developmen of biotechnologies. Development of nuclear energy will be based on WWER-440 and WWER-1000 type NPPs. Heat-only nuclear stations and power and heat nuclear stations will receive a large development effort, as well as sodium-cooled fast reactors of the BN type having 800 and 1600 MW capacity, high-temperature gas-cooled breeders of the BGR-300 type, gas-cooled reactors of the VG-400 type for high-temperature heat supply (500-1000 deg C). It is contemplated to design the TOKAMAK-15 research thermonuclear facility and a pilot thermonuclear reactor for power generation and plutonium production. The program also comprises works aimed at improving reliability and safety of the nuclear installations

  11. IAEA expert assessment of basic issues associated with the preparation and implementation of the Temelin nuclear power plant

    International Nuclear Information System (INIS)

    Horak, B.

    1990-01-01

    An independent IAEA team composed of foreign experts was invited to Temelin with the aim to review the plant with respect to nuclear safety of the site, organization of construction activities, implementation of the construction including quality assurance aspects, preparation of operation, radioactive waste management, reactor core physics, and safety system design. Final reports conclude that the site has a low seismicity and favorable meteorological, hydrogeological and demographic conditions, and thus is well suited to the construction of a nuclear power plant. It has been suggested that within CEZ (the Czech Power Company, the utility operating all power plants in the Czech Republic), all activities associated with nuclear power should be concentrated in one division which would be responsible for safety supervision, technical assistance, and overall construction management. Particular efforts must be devoted to quality assurance. The expert team stated that in its design, the WWER-1000 reactor approaches state-of-the-art pressurized water reactors which are being launched in other countries. (M.D.)

  12. Some remarks about engineering factor determination

    International Nuclear Information System (INIS)

    Svarny, J.

    2010-01-01

    The problem of the power distribution uncertainties is in general a multidimensional problem of random vector and problem of multidimensional Probability Density Function. The standard methodology of derivation WWER engineering factors is based on representation of analyzed power peaking in the linear form of random factors and on presumption about their normal Probability Density Function. In this paper is presented the derivation of locally defined engineering factors and for mechanical factors has been performed their reformulation. Final formulation of engineering factors as a statistics of relative deviations involves new parameter-mean. Engineering factors definition from so called endpoints of uncertainty tolerance interval is recommended. Approach (95%/95%) for normal Probability Density Function is discussed in detail, the relation to present standard uncertainty methodology of power distribution is found and problem of optimality in tolerance factor finding including limitation of sample size is discussed. On the bases of statistically based uncertainty kinf analysis for linear model has been shown that multivariate outputs vector of power peaking has nearly normal Probability Density Function independently on the Probability Density Function character of input multivariate vector from under its small dimension (lower than number of fuel assemblies in 1/6 core symmetry). Finally the development of the methodological part of the engineering factors for WWER-1000 design macrocode MOBY-DICK is described and their calculation direct on the bases of self-powered detector experimental data of Temelin NPP and Volgodonsk NPP has been performed with inclusion the variability of detector (self-powered detector). (Author)

  13. Temelin safety monitor

    International Nuclear Information System (INIS)

    Mlady, O.

    2000-01-01

    Temelin NPP is a WWER-1000/320 two unit plant under construction, originally designed according to the standards of the former Soviet Union. After a series of reviews in the 80s, a decision was taken to upgrade the design of Temelin, including the supply of fuel and instrumentation and instrumentation and control system (I and C). Details on the current design and other related safety matters were presented to the nuclear community in a meeting organized by the IAEA in November 1994. Based upon recommendations of IAEA OSART missions, post TMI requirements and Temelin Risk Audit recommendations it was decided to perform a Probabilistic Safety Assessment within the Temelin PSA Project. The general purpose of this project was to perform systematic examination of the Temelin Unit 1 NPP for severe accident vulnerabilities by performance of a Level 1 and 2 PSA study. In addition to the completion of Temelin documented living PSA model, the decision was to develop and implement a PSA based software tool able to analyze real and scheduled plant conditions for determining the risk impact of plant configurations and on-line maintenance activities. This paper provides an overview of the key features of the Temelin Safety Monitor, describes its development activities and its current status and intended use at Temelin NPP for PSA applications. (author)

  14. National Report of Bulgaria [National practice at WWER sites

    International Nuclear Information System (INIS)

    2017-01-01

    Kozloduy NPP Plc. is located in north-western Bulgaria at about 3 km from the Danube River, which serves as a source of cooling water and a receiver of liquid released from the plant (Fig. I.3). Six nuclear units were constructed on Kozloduy NPP site. Units 1 to 4 are WWER-440/230 reactors and Units 5 and 6 are WWER-1000/320 reactors. Units 1 to 4 were commissioned from 1974 to 1982. Units 5 and 6 were commissioned in 1987 and 1991, respectively. On December 31, 2002 Units 1 and 2 were shut down and on December 31, 2006 Units 3 and 4 were shut down as well. Spent Nuclear Fuel was removed from the pre-reactor pools and the reactors have been prepared for decommissioning. Units 1 and 2 (in 2008) and Units 3 and 4 (in 2012) were declared as Decommissioning Facilities by Ministerial Decree and outsourced from the plant. A separate complex consisting of facilities for treatment and conditioning of wet and dry solid radioactive waste (RAW) and a facility for storage of conditioned RAW was commissioned in 2001 on site. The complex is situated next to the Auxiliary Building of Units 5 and 6. In 2005 this complex was outsourced as a separate enterprise called the Specialized Enterprise for Radioactive Waste (SE RAW). There are two Spent Fuel Storage Facilities (SFSF) for wet and dry storage located on the plant site as well.

  15. Revision of nuclear power plants safety systems' routine testing assigned periodicity during the design extension period

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Kozlov, Yi.L.; Chulkyin, O.O.

    2017-01-01

    When nuclear power plants safety systems thermal equipment operation extending, a necessary requirement shall rely on revising the scheduled equipment tests frequency to optimize those tests schedule taking into account the equipment remained lifespan. On the one hand, there exists a need for tests frequency increase to detect ''hidden'' failures, and on the another, frequent tests cause a premature wear of the equipment. Proposed is an original method for optimizing the frequency of NPPs safety systems thermal engineering equipment testing. Essential in the proposed method is the optimization criterion chosen: index of security system failure probability non-exceedance during the beyond-design operating period as referred to the failure probability expected considering the equipment residual resource during the design operating period. The developed method implementation when applied to NPPs safety systems operated beyond the design service life at nuclear power plants with WWER-1000 series reactors, allowed to establish that the optimal tests frequency makes half the designed one when the equipment service life is extended by five years and three times less that the designed frequency when subject lifespan extended by 10 years.

  16. Nuclear power in Czechoslovakia and its development within the CMEA collaboration framework

    International Nuclear Information System (INIS)

    Gavel, S.

    1984-01-01

    The state of and prospects for the development of nuclear pwer and energy machine building in Czechoslovakia are considered. In 1980 the NPPs generated 1.5% of the country's energy. It is envisaged that by 1990 the share of NPPs will rise to 9.2% and by 2000 to 15%. Since 1974, when an agreement with the Soviet Union was signed on cooperation and specialization in producing equipment for NPPs, a new industry, nuclear power machine building, has been successfully developed in Czechoslovakia. The production of steam generators, volume compensators, main gate valves, pipelines, special fittings, special pumps and auxiliary systems of the primary circuit has been mastered. Presently, the Czechoslovakian industry is capable of manufacturing above 80% of the whole servicing equipment of WWER-440 power units. Alongside with the USSR, Czechoslovakia has become the second greatest supplier of equipment for NPPs among the COMECON member-states. By 2000 it is planned to put into operation 12 WWER-440 power unit and 6 WWER-1000 power units with the total capacity of 11280 MW. An experimental agro-nuclear complex will soon be constructed in which NPP heat will be utilized for growing vegetables, fruit, mushrooms and for breeding fish

  17. Methodology for qualification of in-service inspection systems for WWER nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1998-03-01

    Program was initiated by IAEA in 1990 with the aim to assist the countries of Central and Eastern Europe and former Soviet Union in evaluating the safety of their first generation WWER-440/230 nuclear power plants. The main objectives were: to identify major design and operational safety issues; to establish international consensus on priorities for safety improvements; and to provide assistance in the review of the competence and and adequacy of safety improvement programs. The scope was extended in 1992 to include RBMK, WWER-440/312 and WWER-1000 plants in operation and under construction. Integrity of primary circuit is fundamental for the safe operation of any nuclear power plant. In-service inspection (ISI) in general terms and in particular, non-destructive tests (NDT) play a key role in maintaining primary circuit integrity. This report provides a methodology for qualification of ISI systems which might be used by WWER operating countries as a commonly accepted basis for further development of the necessary qualification related infrastructures. It also provides several qualification principles defining the administrative framework needed for the practical implementation of the methodology, a description of the process of qualification of an inspection system, specifying its minimum technical and documentation related requirements, as well as specific requirements with regard to the NDT procedures, equipment and personnel to be qualified and the test specimen to be used in practical trials. Finally, the report suggests an appropriate distribution of responsibilities among all the parties involved in a qualification process, based on international practice

  18. East/west steels for reactor pressure vessels

    International Nuclear Information System (INIS)

    Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

    1997-01-01

    The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

  19. Approach to evaluation and prediction of lifetime characteristics of NPP valve on the basis of operation data

    International Nuclear Information System (INIS)

    Emelyanov, V.; Kamyshnikov, O.; Dovgalyuk, V.; Plying, B.

    1994-01-01

    The report contains brief description of the main activity stages for testing, evaluation and prediction of reliability factors (including characteristics and factors of longevity) for NPP operating equipment. Valves equipped with electric drive that are installed in level control system of steam generator in WWER-1000 reactor are taken as an example. Main emphasis is made on classification of failures which had taken place during operation, on detection of prevailing mechanisms of ageing and on assessment of operation factors of reliability and methods of their testing, assessment and prediction. Principles of product ageing parameters selection are briefly described as well as mathematic methods used for quantitative assessment of products reliability factors according to its operation data. The report includes considerations on procedure of operating evaluation, testing and prediction of complex unique equipment based on testing of state vectors path, probabilities of defining parameters to be tested characterizing operability of set components within the assumed boundaries written in design and operation documentation are components of the vectors mentioned above. 9 refs, 4 figs

  20. Advanced in-core monitoring system for high-power reactors

    International Nuclear Information System (INIS)

    Mitin, V.I.; Alekseev, A.N.; Golovanov, M.N.; Zorin, A.V.; Kalinushkin, A.E.; Kovel, A.I.; Milto, N.V.; Musikhin, A.M.; Tikhonova, N.V.; Filatov, V.P.

    2006-01-01

    This paper encompasses such section as objective, conception and engineering solution for construction of advanced in-core instrumentation system for high power reactor, including WWER-1000. The ICIS main task is known to be an on-line monitoring of power distribution and functionals independently of design programs to avoid a common cause error. This paper shows in what way the recovery of power distribution has been carried out using the signals from in-core neutron detectors or temperature sensors. On the basis of both measured and processed data, the signals of preventive and emergency protection on local parameters (linear power of the maximum intensive fuel rods, departure from nucleate boiling ratio peaking factor) have been automatically generated. The paper presents a detection technology and processing methods for signals from SPNDs and TCs, ICIS composition and structure, computer hardware, system and applied software. Structure, composition and the taken decisions allow combining class IE and class B and C tasks in accordance with international standards of separation and safety category realization. Nowadays, ICIS-M is a system that is capable to ensure: monitoring, safety, information display and diagnostics function, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and power units. Meanwhile, it reduce negative influence of human factor on thermal technical reliability in the operational process (Authors)