WorldWideScience

Sample records for wwer nauchno-tekhnicheskie osnovy

  1. Imenovanje jednonitnih homopolimera i kopolimera na osnovi podrijetla (I. dio

    Directory of Open Access Journals (Sweden)

    Vida Jarm

    2018-04-01

    Full Text Available Prethodne IUPAC-ove preporuke imenovanja (nomenklature jednonitnih polimera na osnovi podrijetla (NOP odnosile su se uglavnom na kopolimere, nelinearne polimere, združene polimere i generičke polimere. Pravila navedena u ovim preporukama omogućuju jasnije i preciznije imenovanje polimera na osnovi podrijetla, kako homopolimera tako i kopolimera. Prikazani sveobuhvatni sustav imenovanja polimera na osnovi podrijetla prihvatljiva je alternativa sustavu imenovanja polimera na osnovi strukture. Zbog raširene i česte uporabe dodatno su opisane i preporuke za uporabu uvriježenih imena polimera.

  2. 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

  3. 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

  4. 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.

  5. Expert system GIP-WWER for verification of seismic adequacy of WWER equipment

    International Nuclear Information System (INIS)

    Masopust, R.

    1999-01-01

    The aim of this report is to describe the modified Generic Implementation Procedure (GIP) titled GIP-WWER which can be used to verify seismic adequacy of the safe shutdown mechanical and electrical equipment and distribution systems of operating or constructed WWER NPPs, namely WWER-440/213 type. The WWER-GIP procedure was prepared using available information contained in GIP and the experience taken from various seismic inspections and evaluations of WWER type NPPs performed in the last five years

  6. Vrednotenje rešitve za prostoročno uporabo računalnika na osnovi naprave Emotiv EPOC++

    OpenAIRE

    Špindler, Matic

    2017-01-01

    Naprave in načini njihove uporabe za upravljanje računalnika se ne spreminjajo zelo pogosto. Z napredkom tehnologije se pojavljajo nove rešitve, ki so sedaj bolj dostopne in ponujajo nove načine interakcije z računalnikom. V tej magistrski nalogi bomo ovrednotili rešitev za prostoročno uporabo računalnika na osnovi možganskega komunikacijskega vmesnika (BCI) Emotiv EPOC+, ki nam omogoča upravljanje računalnika s pomočjo obrazne mimike in misli (EEG). Z raziskavo želimo preveriti uporabnost te...

  7. 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)

  8. 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.

  9. 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

  10. 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)

  11. Thermophysical aspects of WWER safety

    International Nuclear Information System (INIS)

    Kolochko, Vladimir N.

    1999-01-01

    The paper presents a review of the main thermophysical aspects of NPPs safety and efficiency increase applied to WWERs. Improvement of operating WWER units is the main short-term and medium-term tasks of the utilities in Ukraine. The new generation of reactors for increasing the reactor facilities efficiency should utilize achievements of thermo physics research results. The thermophysical aspects of NPPs safety and efficiency are envisaged in the context of the atomic energy development strategy. The analysis of thermophysical processes occurring in the core shows that a number of problems concerning boiling crisis and heat exchange intensification during transient and accidents are not solved in spite of numerous calculations and experimental research. The up-to-date safety conception includes severe accidents consideration for safety assessment. Review of severe accidents management is presented. Additional validation of the West advanced thermal hydraulic codes which are applied for WWERs reassessment calculations is required. Contribution of the processes which might occur in WWER containment to safety problems solving is considered as well. (author)

  12. Surveillance of WWER-440 fuel performance

    International Nuclear Information System (INIS)

    Simko, J.; Urban, P.

    1999-01-01

    In this lecture next problems of surveillance of WWER-440 fuel performance are presented: surveillance of WWER-440 fuel performance at Mochovce NPP; basic data of WWER-440 reactor; in-core reactor measuring system 'SVRK'; basic level of SVRK; information output of basic level of SVRK; surveillance of fuel performance; table of permissible operation conditions of the reactor; limitation of the unit 1 power at the beginning of the operation; cyclic changes of power; future perspectives

  13. Test model of WWER core

    International Nuclear Information System (INIS)

    Tikhomirov, A. V.; Gorokhov, A. K.

    2007-01-01

    The objective of this paper is creation of precision test model for WWER RP neutron-physics calculations. The model is considered as a tool for verification of deterministic computer codes that enables to reduce conservatism of design calculations and enhance WWER RP competitiveness. Precision calculations were performed using code MCNP5/1/ (Monte Carlo method). Engineering computer package Sapfir 9 5andRC V VER/2/ is used in comparative analysis of the results, it was certified for design calculations of WWER RU neutron-physics characteristic. The object of simulation is the first fuel loading of Volgodon NPP RP. Peculiarities of transition in calculation using MCNP5 from 2D geometry to 3D geometry are shown on the full-scale model. All core components as well as radial and face reflectors, automatic regulation in control and protection system control rod are represented in detail description according to the design. The first stage of application of the model is assessment of accuracy of calculation of the core power. At the second stage control and protection system control rod worth was assessed. Full scale RP representation in calculation using code MCNP5 is time consuming that calls for parallelization of computational problem on multiprocessing computer (Authors)

  14. SCORPIO - WWER core surveillance system

    International Nuclear Information System (INIS)

    Hornaes, Arne; Bodal, Terje; Sunde, Svein; Zalesky, K.; Lehman, M.; Pecka, M.; Svarny, J.; Krysl, V.; Juzova, Z.; Sedlak, A.; Semmler, M.

    1998-01-01

    The Institut for energiteknikk has developed the core surveillance system SCORPIO, which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety, as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. The system has been implemented on western PWRs, but the basic concept is applicable to a wide range of reactors including WWERs. The main differences between WWERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a WWER version of SCORPIO has been done in co-operation with the Nuclear Research Institute Rez, and industry partners in the Czech Republic. The first system is installed at Dukovany NPP, where the Site Acceptance Test was completed 6. March 1998.(Authors)

  15. 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.

  16. 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

  17. Water chemistry in WWER reactors

    International Nuclear Information System (INIS)

    Yurmanov, V.A.; Mamet, V.A.; Shestakov, Yu.M.; Amosov, M.M.

    1997-01-01

    In this paper ''Water Chemistry in WWER Reactors'', are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH T values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ''soft decontamination'' involving changing the KOH concentration and, hence, the pH T before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs

  18. Water chemistry in WWER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yurmanov, V A; Mamet, V A; Shestakov, Yu M; Amosov, M M [All-Russian Scientific Research Inst. for Nuclear Power Plants Operation, Moscow (Russian Federation)

    1997-02-01

    In this paper ``Water Chemistry in WWER Reactors``, are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH{sub T} values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ``soft decontamination`` involving changing the KOH concentration and, hence, the pH{sub T} before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs.

  19. WWER-440 type reactor core

    International Nuclear Information System (INIS)

    Mizov, J.; Svec, P.; Rajci, T.

    1987-01-01

    Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

  20. Transmutation potential of reactor WWER-440

    International Nuclear Information System (INIS)

    Darilek, P.; Sebian, V.; Necas, V.

    2001-01-01

    Theoretical evaluation of WWER-440 transmutation potential by HELIOS - code is presented. Transmutation method proposal comprising special transmutation pins, combined FA and simple reprocessing is described. Transmutation efficiency of the method is characterized (Authors)

  1. 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

  2. Radioactive waste management at WWER type reactors

    International Nuclear Information System (INIS)

    1993-05-01

    This report was prepared within the framework of the Technical Assistance Regional Project on Advice on Waste Management at WWER Type Reactors, which was initiated by the IAEA in 1991. The Regional Project is an integral part of the IAEA's activities directed towards improvement of the safety and reliability of nuclear power plants with WWER type reactors (Soviet designed PWRs). Forty-five WWER type units are currently in operation and twenty-five are under construction in Bulgaria, Czechoslovakia, Finland, Hungary and the former USSR. The idea of regional collaboration between eastern European countries under the auspices of the IAEA was discussed for the first time during the last meeting of the Council for Mutual Economic Assistance (CMEA) on spent fuel and radioactive waste management, held in Rez, Czechoslovakia, in October 1990. Since then, the CMEA and some of its former Member States have ceased to exist. However, there are many reasons for eastern European countries to continue their regional collaboration at a higher level. The USSR, the designer and supplier of WWER type reactors in eastern European countries, participated in the first phase of the project. The majority of WWER type reactors are situated in States of the former USSR (Russia and Ukraine). The main results of the first phase of the Regional Project are: (i) Re-establishment of communication channels among eastern European countries operating WWER type reactors by incorporating the IAEA's technical assistance; (ii) Identification of common waste management problems (administrative and technical) requiring resolution; (iii) Familiarization with radioactive waste management systems at nuclear power plants with WWER type reactors - Paks (Hungary), Loviisa (Finland), Jaslovske Bohunice (Czechoslovakia) and Novovoronezh (Russian Federation). Tabs

  3. 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)

  4. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Casadei, A.L.; Doshi, P.K.

    1993-01-01

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

  5. WWER reactor fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    Stefanova, S.; Chantoin, P.; Kolev, I.

    1994-01-01

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

  6. WWER reactor fuel performance, modelling and experimental support. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Chantoin, P; Kolev, I [eds.

    1994-12-31

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: (1) WWER Fuel Performance and Economics: Status and Improvement Prospects: (2) WWER Fuel Behaviour Modelling and Experimental Support; (3) Licensing of WWER Fuel and Fuel Analysis Codes; (4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items.

  7. Operational behaviour of WWER nuclear power units after Chernobyl accident

    International Nuclear Information System (INIS)

    Milivojevic, S.; Spasojevic, D.

    2000-01-01

    The indicators of effectiveness of WWER operation, in 1987-1998 were analyzed. For three groups of nuclear units (WWER, NPP Kozloduy, NPP Paks), the trends of Indicators flow were established. The comparative analysis of forced outage rate, and load factor of WWERs and nuclear units all in the world was carried out; it gives the general picture of accident influence on the states and the relations of these indicators in considered period (author)

  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. Comparative study WWER-440/230 reactors and the Juragua-based WWER-440/318 project

    International Nuclear Information System (INIS)

    1994-01-01

    The work consists in a comparative analysis between WWER-440/230 and WWER-440/318 type reactors on the basis of the major findings and recommendations given in the extra budgetary Programme of the International Atomic Energy Agency on the safety of these reactors and summarized in the document 'The safety of WWER-440 model 230 nuclear power plant'. The fundamental emphasis was done in the design concept of these reactors

  10. 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)

  11. Partially closed fuel cycle of WWER-440

    International Nuclear Information System (INIS)

    Darilek, P.; Sebian, V.; Necas, V.

    2002-01-01

    Position of nuclear energy at the energy sources competition is characterised briefly. Multi-tier transmutation system is outlined out as effective back-end solution and consequently as factor that can increase nuclear energy competitiveness. LWR and equivalent WWER are suggested as a first tier reactors. Partially closed fuel cycle with combined fuel assemblies is briefed. Main back-end effects are characterised (Authors)

  12. Problems of WWER-440 dynamic changes

    International Nuclear Information System (INIS)

    Rydzi, S.

    1986-01-01

    The data processing capability of the DYNAMIKA program is presented and demonstrated for the calculation of coolant parameters and heat transfer variables of the WWER reactor in accident and transient modes of operation. An experimental outage is described of the TG 11 turbogenerator at the V-1 Jaslovske Bohunice nuclear power plant. The measured values were compared with values calculated using the DYNAMIKA program. The graphic representation makes it evident that the mathematical model comes very close to reality. (J.B.)

  13. Physics measurements on WWER-440 diagnostic assemblies

    International Nuclear Information System (INIS)

    Dach, K.; Jirousek, V.; Kott, J.; Horak, J.; Teren, S.; Nemec, J.

    1980-01-01

    The aims of physics measurements using diagnostic assemblies are the development of neutron noise diagnostics methods, the improvement of knowledge of the physical properties of the WWER reactor cores, the testing of computer programs, and the specification of nuclear safety criteria and the obtaining of information allowing the optimum nuclear fuel economy. The instrumentation of diagnostic assemblies is briefly described, including miniature fission chambers, SPN detectors and calorimeters. The method of evaluating and experimental testing is shown. (M.S.)

  14. Fracture analyses of WWER reactor pressure vessels

    International Nuclear Information System (INIS)

    Sievers, J.; Liu, X.

    1997-01-01

    In the paper first the methodology of fracture assessment based on finite element (FE) calculations is described and compared with simplified methods. The FE based methodology was verified by analyses of large scale thermal shock experiments in the framework of the international comparative study FALSIRE (Fracture Analyses of Large Scale Experiments) organized by GRS and ORNL. Furthermore, selected results from fracture analyses of different WWER type RPVs with postulated cracks under different loading transients are presented. 11 refs, 13 figs, 1 tab

  15. Fracture analyses of WWER reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Sievers, J; Liu, X [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    1997-09-01

    In the paper first the methodology of fracture assessment based on finite element (FE) calculations is described and compared with simplified methods. The FE based methodology was verified by analyses of large scale thermal shock experiments in the framework of the international comparative study FALSIRE (Fracture Analyses of Large Scale Experiments) organized by GRS and ORNL. Furthermore, selected results from fracture analyses of different WWER type RPVs with postulated cracks under different loading transients are presented. 11 refs, 13 figs, 1 tab.

  16. 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

  17. 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

  18. Ammonia role in WWER primary circuit water chemistry optimization

    International Nuclear Information System (INIS)

    Kritskij, V.G.; Stjagkin, P.S.; Chvedova, M.N.; Slobodov, A.A.

    1999-01-01

    Ammonia influence on iron crud's solubility at 300 deg. C and different relations of boric acid and alkaline cation sum are considered. Reduction of dose rate on WWER-440 steam generators at average ammonia concentration increasing is empirically explained. Practical recommendations on optimization of WWER primary circuit water chemistry are given. (author)

  19. Reliability of operating WWER monitoring systems

    International Nuclear Information System (INIS)

    Yastrebenetsky, M.A.; Goldrin, V.M.; Garagulya, A.V.

    1996-01-01

    The elaboration of WWER monitoring systems reliability measures is described in this paper. The evaluation is based on the statistical data about failures what have collected at the Ukrainian operating nuclear power plants (NPP). The main attention is devoted to radiation safety monitoring system and unit information computer system, what collects information from different sensors and system of the unit. Reliability measures were used for decision the problems, connected with life extension of the instruments, and for other purposes. (author). 6 refs, 6 figs

  20. 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

  1. 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

  2. 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.

  3. WWER-type NPP spray ponds screen

    International Nuclear Information System (INIS)

    Nikolova, M.; Jordanov, M.; Denev, J.; Markov, D.

    2003-01-01

    The objective of this study is to develop a protection screen of WWER-type NPP spray ponds. The screen design is to ensure reduction of the water droplets blown by the wind and, if possible, their return back to the spray ponds. The cooling capacity of the ponds is not to be changed below the design level for safety reasons. Computational fluid dynamics analysis is used to assess the influence of each design variant on the behavior of the water droplets distribution. Two variants are presented here. The one with plants is found not feasible. The second variant, with steel screen and terrain profile modification is selected for implementation. (author)

  4. Failure analysis for WWER-fuel elements

    International Nuclear Information System (INIS)

    Boehmert, J.; Huettig, W.

    1986-10-01

    If the fuel defect rate proves significantly high, failure analysis has to be performed in order to trace down the defect causes, to implement corrective actions, and to take measures of failure prevention. Such analyses are work-consuming and very skill-demanding technical tasks, which require examination methods and devices excellently developed and a rich stock of experience in evaluation of features of damage. For that this work specifies the procedure of failure analyses in detail. Moreover prerequisites and experimental equipment for the investigation of WWER-type fuel elements are described. (author)

  5. 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

  6. Experience in WWER fuel assemblies vibration analysis

    International Nuclear Information System (INIS)

    Ovtcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.

    2003-01-01

    It is stated that the vibration studies of internals and the fuel assemblies should be conducted during the reactor designing, commissioning and commercial operation stages and the analysis methods being used should complement each other. The present paper describes the methods and main results of the vibration noise studies of internals and the fuel assemblies of the operating NPPs with WWER reactors, as an example of the implementation of the comprehensive approach to the analysis on equipment flow-induced vibration. At that, the characteristics of internals and fuel assemblies vibration loading were dealt jointly as they are elements of the same compound oscillating system and their vibrations have the interrelated nature

  7. WWER fuel: Results of post irradiation examination

    International Nuclear Information System (INIS)

    Markov, D.V.; Smirnov, V.P.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

    2006-01-01

    Experience in the field of fabrication, operation, testing and post-irradiation examinations (PIE) made it possible to settle the following requirements for a new generation of WWER nuclear fuel: - For WWER-1000 FA, the service life is no less than 5 years, 3 alternative fuel cycles (FC): 12 months x 4 FCs, 12 months x 5 FCs and 18 months x 3 FCs; - For WWER-440 FA, fuel cycle is 12 months x 5 FCs and a part of operating assembly is left for the 6. year; - High fuel burnup - up to 70 MWd/kgU; - Dimensional stability of FA and its components; - FA repairability; - Adaptability of fuel cycles; - Maintenance of maneuvering operating conditions at the NPP; - Reliability of control rod operation; - High serviceability level - FE leakage is no worse than 10-5 l/year. In order to provide the fulfillment of the above-given requirements, designers and production engineers have worked out cumulative measures and engineering solutions, which are introduced in development of a new generation fuel. Currently old design FA-M assemblies provided with steel skeleton are being operated in WWER-1000 reactors at Ukrainian and Bulgarian NPPs. As for Russian NPPs, new-type FAs are operated. These are advanced FAs (AFA), FA-A and FA-2 provided with zirconium alloy skeletons. A design of the second generation of WWER-440 operating assemblies was developed with respect to changes in some geometrical parameters, fastening of FEs in the lower grid (splinting was substituted for collet), usage of reinforcing rib under the lower grid, anti-debris filter and hafnium elements of junction unit as well as hafnium content decrease from 0.05 % mass down to 0.01% mass in zirconium materials. They are basic designs of FAs in order to be introduced in a five-year fuel cycle of WWER-440 NPPs in Czech Republic and Slovakia since 2005 and have got prospects for development. The operating experience of dismountable operating assemblies at the Loviisa NPP, vibration-proof operating assemblies at the

  8. Design of the WWER-440 pressurizer

    International Nuclear Information System (INIS)

    Bednarek, L.

    1978-01-01

    The main specifications are presented of a pressurizer for the WWER-440 reactor as are factors securing operating reliability and the required life of the equipment. Ferritic-pearlite steel 22K is the basic material for the manufacture of the shell, support and couplings. The internal parts and the heads are manufactured from the 08KH18N10T chromium-nickel austenitic steel, the bolts and nuts of manholes from chromium-molybdenum steels 25KH1MF and 25KH2MFA. Quality testing of basic materials and quality control during the process of manufacture are briefly described. (Z.M.)

  9. Temperature etalon of WWER-440 reactor

    International Nuclear Information System (INIS)

    Stanc, S.; Slanina, M.

    2001-01-01

    The presentation deals with the description, parameters and advantages of use of the temperature etalon. The system ensures temperature measurement of reactor outlet and inlet temperatures with high accuracy. Accuracy of temperature measurement is 0.18 deg C, accuracy of temperature difference measurement is 0.14 deg C, both with probability 0.95. Using the temperature etalon it is possible to increase accuracy of the standard temperature reactor measurements and to check their accuracy in the course of power reactor statuses in every measurement cycle. Temperature reactor etalon was installed in 12 WWER-440 units in Slovakia, Bohemia and Bulgaria. (Authors)

  10. Reliability of operating WWER monitoring systems

    Energy Technology Data Exchange (ETDEWEB)

    Yastrebenetsky, M A; Goldrin, V M; Garagulya, A V [Ukrainian State Scientific Technical Center of Nuclear and Radiation Safety, Kharkov (Ukraine). Instrumentation and Control Systems Dept.

    1997-12-31

    The elaboration of WWER monitoring systems reliability measures is described in this paper. The evaluation is based on the statistical data about failures what have collected at the Ukrainian operating nuclear power plants (NPP). The main attention is devoted to radiation safety monitoring system and unit information computer system, what collects information from different sensors and system of the unit. Reliability measures were used for decision the problems, connected with life extension of the instruments, and for other purposes. (author). 6 refs, 6 figs.

  11. 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

  12. Equipment and building structures ageing management for WWER type NPPs

    International Nuclear Information System (INIS)

    Mayboroda, O.

    2001-01-01

    This report presents the working group 'Equipment and building structures ageing management for WWER type NPPs' activities. The analysis of experience in ageing management, recommendations for regulatory guidelines on ageing management, investigation of case studies, definition suitable communication channels among regulators for ageing related data are given. Analyses of water chemistry, inspection data (safety margins criteria), plugging criteria, volume and time of ECT implementation in all WWER countries are presented. The results of Working group activity show that it is advisable to concentrate efforts on: set up the permanent communication channel among regulators, collection of regulatory criteria for WWER type NPP key components based on understanding of ageing mechanisms and data collection

  13. WWER water chemistry related to fuel cladding behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez (Czech Republic); Vrtilkova, V [Nuclear Fuel Inst., Prague (Czech Republic)

    1997-02-01

    Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. (author). 14 refs, 5 figs, 3 tabs.

  14. 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

  15. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  16. 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

  17. 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

  18. Systems for noise diagnostics of WWER nuclear power plants

    International Nuclear Information System (INIS)

    Por, G.

    1996-01-01

    The aim of this paper is to give a short overview of the noise diagnostics system developed by Hungarian firms which are in operation in WWER type NPP Units. Giving a list of systems developed for noise diagnostics of WWER reactors we present their main characteristics, their goal and some of their achievements. The second part deals with the problem of acceptance of noise system by NPP and regulations. (author). 24 refs

  19. Systems for noise diagnostics of WWER nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Por, G [Technical Univ. of Budapest, Budapest (Hungary)

    1997-12-31

    The aim of this paper is to give a short overview of the noise diagnostics system developed by Hungarian firms which are in operation in WWER type NPP Units. Giving a list of systems developed for noise diagnostics of WWER reactors we present their main characteristics, their goal and some of their achievements. The second part deals with the problem of acceptance of noise system by NPP and regulations. (author). 24 refs.

  20. 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

  1. 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.

  2. 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.

  3. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L.; Wilam, M. [Vitkovice NPP Services (Switzerland); Herman, M. [Vuje, Trnava (Slovakia)

    1997-12-31

    At operated Czech and Slovak nuclear power plants the 80 % criteria for crack or other bulk defect depth is used for steam generator heat exchanging tubes plugging. This criteria was accepted as the recommendation of designer of WWER steam generators. Verification of this criteria was the objective of experimental program performed by Vitkovice, J.S.C., UJV Rez, J.S.C. and Vuje Trnava, J.S.C .. Within this program the following factors were studied: (1) Influence of secondary water chemistry on defects initiation and propagation, (2) Statistical evaluation of corrosion defects progression at operated SG, and (3) Determination of critical pressure for tube rupture as a function of eddy current indications. In this presentation items (2) and (3) are considered.

  4. Some consideration on reviewing of WWER

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki [Yokohama National University, Mechanical Engineering and Material Sciences, Tokiwadai, Hodogaya, Yokohama (Japan)

    1993-07-01

    The main points of NPP safety are to prevent core melt-down and to mitigate the divergence of radioactive materials to the atmosphere in general. However, destroying of the reactor building may cause damages of various mechanical and other components and piping systems, and these failures may proceed to LOCA and other critical failures of safety related items. The main items that are considered in view of upgrading the existing WWER type reactors are: buildings, emergency power supply systems, anchoring devices and concrete structures, mechanical components, water storage, piping support, civil engineering structure. This pipe describes the method for improving seismic capacity of the operating NPPs and to keep their safety under seismic conditions.

  5. Some consideration on reviewing of WWER

    International Nuclear Information System (INIS)

    Shibata, Heki

    1993-01-01

    The main points of NPP safety are to prevent core melt-down and to mitigate the divergence of radioactive materials to the atmosphere in general. However, destroying of the reactor building may cause damages of various mechanical and other components and piping systems, and these failures may proceed to LOCA and other critical failures of safety related items. The main items that are considered in view of upgrading the existing WWER type reactors are: buildings, emergency power supply systems, anchoring devices and concrete structures, mechanical components, water storage, piping support, civil engineering structure. This pipe describes the method for improving seismic capacity of the operating NPPs and to keep their safety under seismic conditions

  6. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L; Wilam, M [Vitkovice NPP Services (Switzerland); Herman, M [Vuje, Trnava (Slovakia)

    1998-12-31

    At operated Czech and Slovak nuclear power plants the 80 % criteria for crack or other bulk defect depth is used for steam generator heat exchanging tubes plugging. This criteria was accepted as the recommendation of designer of WWER steam generators. Verification of this criteria was the objective of experimental program performed by Vitkovice, J.S.C., UJV Rez, J.S.C. and Vuje Trnava, J.S.C .. Within this program the following factors were studied: (1) Influence of secondary water chemistry on defects initiation and propagation, (2) Statistical evaluation of corrosion defects progression at operated SG, and (3) Determination of critical pressure for tube rupture as a function of eddy current indications. In this presentation items (2) and (3) are considered.

  7. Improvements of radioactive waste management at WWER nuclear power plants

    International Nuclear Information System (INIS)

    2006-04-01

    This report is part of a systematic IAEA effort to improve waste management practices at WWER plants and to make them consistent with the current requirements and standards for safe and reliable operation of nuclear power plants. The report reviews the wet and dry solid waste management practices at the various types of WWER nuclear power plants (NPP) and describes approaches and recent achievements in waste minimization. Waste minimization practices in use at western PWRs are reviewed and compared, and their applicability at WWER plants is evaluated. Radioactive waste volume reduction issues and waste management practices are reflected in many IAEA publications. However, aspects of waste minimization specific to individual WWER nuclear power plant designs and WWER waste management policies are not addressed extensively in those publications. This report covers the important aspects applicable to the improvement of waste management at WWER NPP, including both plant-level and country-level considerations. It is recognized that most WWER plants are already implementing many of these concepts and recommendations with varying degrees of success; others will benefit from the included considerations. The major issues addressed are: - Review of current waste management policies and practices related to WWERs and western PWRs, including the influence of the original design concepts and significant modifications, liquid waste discharge limits and dry solid waste clearance levels applied in individual countries, national policies and laws, and other relevant aspects affecting the nature and quantities of waste arisings; - Identification of strategies and methods for improving the radioactive waste management generated in normal operation and maintenance at WWERs. This report is a composite (combination) of the two separate initiatives mentioned above. The first draft report was prepared at the meeting 26-30 May 1997 by five consultants. The draft was improved during an

  8. 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

  9. Life management of SG for WWER plants

    International Nuclear Information System (INIS)

    Trunov, N. B.; Dragunov, Yu. G.; Banyuk, G. F.

    2004-01-01

    Nowadays, 252 steam generators (SG) of horizontal type are in operation at WWER plants constructed by the Russian designs. In connection with end of the specified service life of the reactor plant equal to 30 years the activities are performed on service life extension of the main equipment including the SG. At some Units, throughout the design service life of SG there were problems resulting in necessity of SG replacement. At the same time the SGs at some Units are in successful operation above the design service life. This report deals with the peculiarities of operation of the horizontal SGs and the problems to be highlighted as the most important for service life extension. The main component to determine possibility for SG service life extension is the SG tubing. As the operating experience shows it is water chemistry of the secondary circuit that is the main factor influencing operability of the SG tubing. Therefore, differences in water chemistry organization leads to significant differences in operability of the SG tubing at various Units and in some cases within one Unit. Owing to the fact that the cases of water chemistry disturbance and the process of tubes fouling with the corrosion products of the main condensate system are not excluded, the damages continue to occur. Tube integrity shall be inspected by eddy current method using the various instrument complexes. This method has certain disadvantages but allows to estimate the degree and direction of degradation processes. The results of eddy current test (ECT) can be used to determine the plugging criterion for defective tubes. The significant number of defective tubes at some Units makes a choice of the plugging criterion to be an important problem, on which solution the SG safety, reliability and service life depends. The report deals with directions of activities in service life management for the SG at WWER plants. Main activities are improvement of water chemistry and non-destructive tests.(author)

  10. 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

  11. An approach to WWER fuels with BaCo

    International Nuclear Information System (INIS)

    Marino, A.; Demarco, G.

    2008-01-01

    BaCo is a code for the simulation of the behaviour of a nuclear fuel rod under operation conditions. BaCo, a quasi 2D code based on a finite differences scheme, has been used for simulating PHWR, CANDU, PWR, BWR, MOX, WWER, and experimental fuel rods. We improve the performance of BaCo with a set of tools based on the method of finite elements for 3D analysis of the stress-strain state. We can simulate any UO 2 pellet geometry. Standard WWER-440 fuel assemblies irradiated in the Kola-3 reactor of the CRP FUMEX II of the IAEA were the first WWER simulations with BaCo. We find a very good agreement among our calculations, the experimental results and other qualified fuel codes. We present the BaCo code and our results for PWR and WWER fuels of the CRP FUMEX II, the 3D analysis of WWER fuel pellet and the projections of these results with the Argentinean nuclear fuels development. (authors)

  12. 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

  13. 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

  14. 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

  15. 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.)

  16. New designs of medium power WWER reactor plants

    International Nuclear Information System (INIS)

    Ryzhov, S.B.; Mokhov, V.A.; Nikitenko, M.P.; Chetverikov, A.E.; Veselov, D.O.; Shchekin, I.G.; Petrov, V.V.

    2010-01-01

    The task of constructing NPPs as the objects of regional power industry is included into the Federal Target Program on nuclear power technologies of new generation for the period till 2020. Such NPPs are considered as perspective sources of energy for solution of the problems concerning provision of electric energy, household and industrial heat to the regions with limited capabilities of the power grid. OKB 'GIDROPRESS' present the conceptual study of RP design for the Unit of 600 MW (el.) power, taking into account their long-term experience in the field of development and operation of WWER reactor plants. Practical implementation of WWER-600 and WWER-300 RP designs seems to be feasible: practice in manufacturing the main equipment is available; cooperation of design, scientific organizations and manufacturers of equipment; is established; basic design solutions for equipment are of reference character

  17. 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

  18. Teaching WWERs at Hacettepe University Nuclear Engineering Department in Turkey

    International Nuclear Information System (INIS)

    Ergun, S.

    2011-01-01

    In this study, the challenges faced in the teaching WWER design for the reactor engineering course, which is taught in the Hcettepe University Nuclear Engineering Department are discussed. Since the course is designated taking a western reactor design into account, the computer programs and class projects prepared for the course include models and correlations suitable for these designs. The attempts for modifying the course and developing codes or programs for the course become a challenge especially in finding proper information sources on design in English. From finding proper material properties to exploring the design ideas, teaching WWER designs and using analysis tools for better teaching are very important to modify the reactor engineering course. With the study presented here, the reactor engineering course taught is described, the teaching tools are listed and attempts of modifying the course to teach and analyze WWER designs are explained

  19. 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.

  20. 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

  1. 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

  2. Project margins of advanced reactor design WWER-500

    International Nuclear Information System (INIS)

    Rogov, M.F.; Birukov, G.I.; Ershov, V.G.; Volkov, B.E.

    1994-01-01

    Project criteria for design of advanced WWER-500 reactor within design conditions are compared to the requirements of the Russian regulatory guides. Normal operation limits, safe operation limits for main anticipated operational occurrences and design limits accepted for design basis accidents are considered as in preliminary safety report. It is shown that the basic design criteria in the design of WWER-500 for the anticipated operational occurrences and for design basis accidents are more severe than required in the following regulatory guides General Safety Regulations for Nuclear Power Plants and Nuclear Safety Rules for Reactors of Nuclear Power Plants. This provides certain margins from safety point of view

  3. Seismic resistance of WWER equipment (systematization and generalization)

    International Nuclear Information System (INIS)

    Kaznovski, S.; Chechenov, H.

    1999-01-01

    Within the scope of the contract with IAEA, calculational and experimental investigations of equipment seismic resistance were carried out. Systematic seismic inspection was conducted for 21 WWER type reactors in Armenia, Bulgaria, Russia, Slovakia and Ukraine. On the base of generalisation of extensive real data accumulated during inspection of equipment seismic resistance at different WWER-type NPPs the classification of seismic instability reasons and working out of generalised recommendations were carried out. It gives the possibility to specialists in many cases to expose the seismic insufficiency of equipment and to choose the suitable measures without laborious experimental investigations, tests, and calculations

  4. Safety of WWER type nuclear power plants - viewing from Hungary

    International Nuclear Information System (INIS)

    Voeroess, L.

    1991-01-01

    An evaluation of WWER type nuclear power plants operating in Hungary is given, relative to the safety requirements accepted internationally; how safe can they be regarded and what can be done to assure a high level of safety in all case. After an overview of general safety criteria, an overall description of WWER-440 type nuclear reactors is presented. Design safety, operational safety issues are treated in detail. Safety inspection and safety-related research and development is discussed. Regarding the future, five different issues associated with nuclear reactor safety should be considered. (R.P.) 20 refs.; 12 figs.; 3 tabs

  5. 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)

  6. Power peak in vicinity of WWER-440 control assembly

    International Nuclear Information System (INIS)

    Mikus, J.

    2002-01-01

    This paper presents information concerning the WWER-440 local power peaking problem induced by a control assembly and corresponding investigation possibilities on the light-water zero-power reactor LR-O at the Nuclear Research Institute Rez plc. Brief description of the disposable CA model, experimental arrangement and conditions on the LR-O reactor, preparation of the relevant measurements in the WWER-440 type cores with CA model, as well as some preliminary results of the fission density distribution obtained in a core without boron and with fuel assemblies having profiled enrichment are mentioned too (Author)

  7. Development of codes for physical calculations of WWER

    International Nuclear Information System (INIS)

    Novikov, A.N.

    2000-01-01

    A package of codes for physical calculations of WWER reactors, used at the RRC 'Kurchatov Institute' is discussed including the purpose of these codes, approximations used, degree of data verification, possibilities of automation of calculations and presentation of results, trends of further development of the codes. (Authors)

  8. WWER 440 integrity assessment with respect to PTS events

    International Nuclear Information System (INIS)

    Hrazsky, M.; Mikua, M.; Hermansky, P.

    1997-01-01

    The present state of art in WWER 440 RPV integrity assessment with respect to PTS events utilized in Slovakia is reviewed briefly in this paper. Recent results of some PTS's (very severe) analyses, shortly described in our paper, have confirmed the necessity of elaboration of new more sophisticated procedures again. Such methodology should be based on prepared IAEA Guidance. (author). 13 refs, 1 fig

  9. 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)

  10. Operational results of WWER fuel fabricated by MSZ (Elektrostal, Russia)

    International Nuclear Information System (INIS)

    Asatiani, I.; Balabanov, S.; Beglov, A.; Khryashchev, D.

    2009-01-01

    The presentation brings forth a statistical analysis of the WWER fuel manufactured by OAO MSZ, operational experience. A necessity of such an analysis is determined by the fact that objective operational results prove the appropriateness of the solutions and decisions made by vendor, designer, manufacturer and utility, as well as motivates further fuel improvements. (authors)

  11. WWER in-core fuel management benchmark definition

    International Nuclear Information System (INIS)

    Apostolov, T.; Alekova, G.; Prodanova, R.; Petrova, T.; Ivanov, K.

    1994-01-01

    Two benchmark problems for WWER-440, including design parameters, operating conditions and measured quantities are discussed in this paper. Some benchmark results for infinitive multiplication factor -K eff , natural boron concentration - C β and relative power distribution - K q obtained by use of the code package are represented. (authors). 5 refs., 3 tabs

  12. 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

  13. 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.

  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. Review of RIA and LOCA criteria for WWER fuel

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    The RIA and LOCA fuel safety criteria are under revision in the international community of fuel suppliers, authorities and research organizations. The main criteria will be reviewed in the paper for WWER fuel. Experimental data on the fuel failure behaviour under reactivity-initiated accident (RIA) conditions produced in the last decade in French and Japanese test reactors indicated low failure enthalpy for high burnup fuel compared to fresh fuel. However the high burnup was not the only phenomenon influencing the fuel failure. The oxide scale on the external surface of the fuel rod, hydrogen content of the Zr cladding and the local hydriding seemed also be responsible for the failure at low enthalpy. Furthermore differences have been found between Western design fuel and Russian type WWER fuel. The burnup dependence of fuel failure for WWER fuel was found much less, probably due to the low oxidation during normal operational conditions compared to other PWRs. The recently published Vitanza and KAERI correlations for RIA failure enthalpy have been applied to 23 WWER tests. Experimental data from Russian IGR and BIGR reactors have been used. The calculations have shown that both burnup and cladding oxidation effects must be considered, however the pulse width dependence of failure enthalpy has not been confirmed. During loss of coolant accidents (LOCA) the peak cladding temperature and local oxidation criteria have to be met. The oxidation criterion is under discussion today in many laboratories. The AEKI carried out several experimental series with Zr1%Nb cladding used in WWER reactors. The paper will describe the main results of the tests and present the limit for ductile-brittle transition derived from ring compression test. The behaviour of Zr1%Nb (E110) and Zircaloy-4 claddings under LOCA conditions will be compared as well. (author)

  16. 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)

  17. 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)

  18. Osnovi ! gradniki kondicijske priprave boksarja:

    OpenAIRE

    Korsika, Žiga

    2014-01-01

    The purpose of this article is to introduce a view on boxing, based on biomechanical and physiological analysis of movement with the help of expert literature It highlights the more important aspects of load and effort, which include endurance, power and speed, that suggest a way in which boxers should supplement technical-tactical training in order to better their competitive ability. Namen tega članka je s pomočjo strokovne literature predstaviti pogled na obremenitev in napor v boksu, k...

  19. 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

  20. WWER NPPs fuel handling machine control system

    International Nuclear Information System (INIS)

    Mini, G.; Rossi, G.; Barabino, M.; Casalini, M.

    2001-01-01

    In order to increase the safety level of the fuel handling machine on WWER NPPs, Ansaldo Nucleare was asked to design and supply a new Control System. Two FHM Control System units have been already supplied for Temelin NPP and others supplies are in process for the Atommash company, which has in charge the supply of FHMs for NPPs located in Russia, Ukraine and China. The Fuel Handling Machine (FHM) Control System is an integrated system capable of a complete management of nuclear fuel assemblies. The computer-based system takes into account all the operational safety interlocks so that it is able to avoid incorrect and dangerous manoeuvres in the case of operator error. Control system design criteria, hardware and software architecture, and quality assurance control, are in accordance with the most recent international requirements and standards, and in particular for electromagnetic disturbance immunity demands and seismic compatibility. The hardware architecture of the control system is based on ABB INFI 90 system. The microprocessor-based ABB INFI 90 system incorporates and improves upon many of the time proven control capabilities of Bailey Network 90, validated over 14,000 installations world-wide. The control system complies all the former designed sensors and devices of the machine and markedly the angular position measurement sensors named 'selsyn' of Russian design. Nevertheless it is fully compatible with all the most recent sensors and devices currently available on the market (for ex. Multiturn absolute encoders). All control logic components were developed using standard INFI 90 Engineering Work Station, interconnecting blocks extracted from an extensive SAMA library by using a graphical approach (CAD) and allowing an easier intelligibility, more flexibility and updated and coherent documentation. The data acquisition system and the Man Machine Interface are implemented by ABB in co-operation with Ansaldo. The flexible and powerful software structure

  1. Use of the Benchmarking System for Operational Waste from WWER Reactors

    International Nuclear Information System (INIS)

    2017-06-01

    The focus of this publication is on benchmarking low and intermediate level waste generated and managed during the normal operating life of a WWER, and it identifies and defines the benchmarking parameters selected for WWER type reactors. It includes a brief discussion on why those parameters were selected and their intended benchmarking benefits, and provides a description of the database and graphical user interface selected, designed and developed, including how to use it for data input and data analysis. The CD-ROM accompanying this publication provides an overview of practices at WWER sites, which were to a large extent prepared using the WWER BMS.

  2. WWER steam generator tube structural and leakage integrity

    International Nuclear Information System (INIS)

    Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

    1998-01-01

    The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

  3. 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.)

  4. Armenian earthquake WWER-440 NNPs and Turkish early warning system

    International Nuclear Information System (INIS)

    Bektur, Y.

    1991-01-01

    On December 7, 1988 a severe earthquake occurred at Spitak, approximately 90-100 km far from the Armenian Nuclear Power Plant in Yerivan. Another one named Vrancea earthquake which occurred on 4 March, 1977. During this earthquake, the Kozloduj NPP (Bulgaria) was strongly damaged. Until this event, seismic loadings had received scant attention in the siting of WWER's. However after the Kozlodui damage Soviet designers changed their opinion. In this study, the seismicity of the Black Sea region and eastern Europe, seismic requirements for WWER's and the changes in plants for which to resistant against to the earthquake are given. During the earthquake radiation levels obtained by Turkish early warning system is also given

  5. 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

  6. [95/95] Approach for design limits analysis in WWER

    International Nuclear Information System (INIS)

    Shishkov, L.; Tsyganov, S.

    2008-01-01

    The paper discusses a well-known condition [95%/95%], which is important for monitoring some limits of core parameters in the course of designing the reactors (such as PWR or WWER). The condition ensures the postulate 'there is at least a 95 % probability at a 95 % confidence level that' some parameter does not exceed the limit. Such conditions are stated, for instance, in US standards and IAEA norms as recommendations for DNBR and fuel temperature. A question may arise: why can such approach for the limits be only applied to these parameters, while not normally applied to any other parameters? What is the way to ensure the limits in design practice? Using the general statements of mathematical statistics the authors interpret the [95/95] approach as applied to WWER design limits. (Authors)

  7. 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)

  8. 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

  9. Power peak in vicinity of WWER-440 control rod

    International Nuclear Information System (INIS)

    Mikus, J.

    2003-01-01

    The measurements of axial power (fission density) distribution have been performed by means of gamma activity determination (gamma scanning method) of the irradiated fuel pins detecting gamma quanta in the La peak area - 1596.5 keV of their selected parts having 20 mm length in the axial coordinates range 50-950 mm with a 10 mm step using a rectangular collimator (dimensions 20x10 mm). Two NaI(Tl) scintillation crystals (one as a monitor) with diameter of 40 mm were used, each of them in Pb shielding (thickness 150 mm). The obtained results enlarge the available 'power peaking database' and enable validation of the codes also in important case of zero-boron concentration that corresponds to the end of WWER-440 fuel cycle. This validation can improve the reliability of the calculation results of the power distribution in WWER-440 cores, re-loading schemes etc

  10. Review and comparison of WWER and LWR Codes and Standards

    International Nuclear Information System (INIS)

    Buckthorpe, D.; Tashkinov, A.; Brynda, J.; Davies, L.M.; Cueto-Felgeueroso, C.; Detroux, P.; Bieniussa, K.; Guinovart, J.

    2003-01-01

    The results of work on a collaborative project on comparison of Codes and Standards used for safety related components of the WWER and LWR type reactors is presented. This work was performed on behalf of the European Commission, Working Group Codes and Standards and considers areas such as rules, criteria and provisions, failure mechanisms , derivation and understanding behind the fatigue curves, piping, materials and aging, manufacturing and ISI. WWERs are essentially designed and constructed using the Russian PNAE Code together with special provisions in a few countries (e.g. Czech Republic) from national standards. The LWR Codes have a strong dependence on the ASME Code. Also within Western Europe other codes are used including RCC-M, KTA and British Standards. A comparison of procedures used in all these codes and standards have been made to investigate the potential for equivalencies between the codes and any grounds for future cooperation between eastern and western experts in this field. (author)

  11. 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.

  12. Contribution to the WWER-440 Gd-assembly evolution

    International Nuclear Information System (INIS)

    Cudrnak, P.; Darilek, P.; Necas, V.

    2009-01-01

    Development activities, presented in the paper, contribute to the evolution of gadolinium fuel for the WWER-440 reactors. Short overview of existing and already proposed Gd-assemblies is given. Two new assembly types with improved features are described. Multiplication coefficients and pin power peaking factors of all mentioned gadolinium assemblies are compared. Animations of power distribution behaviour in selected Gd-assemblies are shown. (Authors)

  13. Investigation of WWER fuel behaviour under MIR power ramps

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Novikov, V.V.; Agafonov, S.N.

    1996-01-01

    The paper discusses results of experimental WWER fuel investigation under power ramps. Specificity of using the research reactor ''MIR'' to accomplish scheduled power rating of fuel is considered. The paper presents the methodology of experiments using irradiation facility ''TEST''. Reactor experiments were performed at burn-up ∼ 10000 MW.day/t UO 2 using standard fuel pins and the ones having backfitted fuel and cladding. (author). 7 figs, 1 tab

  14. Assessment of WWER fuel condition in design basis accident

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

    1994-01-01

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

  15. First generation WWER 440 - What is the 'reasonable' price?

    International Nuclear Information System (INIS)

    Pishtuhina, K.

    2004-01-01

    A chronological review of the international nuclear concerns and Three Mile Island and Chernobyl accidents effects on the international nuclear organizations' policies are presented. The IAEA, G 7 and EU approaches for Soviet design plants' safety and the main results from their implementation are described. The premature closure of the first generation WWER type reactors operated in Bulgaria, Lithuania and Slovakia and their unupgradability at a reasonable price are also discussed in this paper

  16. Criticality safety and shielding analysis of WWER-440 fuel configurations

    International Nuclear Information System (INIS)

    Christoskov, I.

    2008-01-01

    An overview is made of some studies performed on the criticality safety and radiation shielding analysis of irradiated WWER-440 fuel storage and handling configurations. The analytical tools are based on the SCALE 4.4a code system, in combination with the TORT discrete ordinates transport code and the BUGLE-96 cross-sections library. The accuracy of some important results is assessed through comparison with independent evaluations and with measurement data. (author)

  17. Performance analysis of WWER-440/230 nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

    This report examines one particular design, the WWER-440/230, the first generation of commercial WWERs, essentially comparable to the western PWR. This design was installed widely in eastern Europe with a total of 16 unites being completed in what are now Armenia, Bulgaria, Germany (the former German Democratic Republic) the Slovak Republic and Russia. The plants in Armenia and Germany (the former German Democratic Republic) have been closed down, but particularly in Bulgaria and to a lesser extent the Slovak Republic the remaining plants supply a significant proportion of the electricity of the country and decisions to close them could not be taken lightly. The aim of this report is twofold: first to determine whether the impression given by these good overall performance indicators is confirmed using more detailed indicators covering a wide range of factors; second, to see to what extent good performance can be attributed to the industrial and institutional environment in which these plants were designed, built and operated. Particular attention is paid to identifying factors that may impact the quality of the service provided, especially those factors under management control which can be strongly influenced by current and future policy changes and those factors that are beyond the plant management control but could have influenced the performance of the power plants. Issues concerning the safety of these plants are of considerable importance, but they remain outside the scope of this report. Conclusions and recommendations formulated by the IAEA related to WWER safety are contained in the series of reports prepared in the framework of the Extrabudgetary Programme on WWER Safety. A programme progress report was published in 1994 (IAEA-TECDOC-773). Refs, figs, tabs

  18. 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%

  19. Assessment of WWER fuel condition in design basis accident

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

  20. Development of a new WWER-440 fuel design

    International Nuclear Information System (INIS)

    Coucil, D.; Totev, T.

    1998-01-01

    In March 1996 British Nuclear Fuel Limited signed a contract with Imatran Voima and Paks Nuclear Power Plant to design, develop, license and supply 5 Lead Test Assemblies to the WWER-440 reactor at Loviisa in Finland. In June 1998 the manufacture of these 5 assemblies (4 fixed assemblies and 1 follower assembly) was completed. The fuel is expected to be loaded into Loviisa Unit 2 reactor during the shutdown scheduled for September of this year. (Authors)

  1. WWER 440 integrity assessment with respect to PTS events

    Energy Technology Data Exchange (ETDEWEB)

    Hrazsky, M; Mikua, M; Hermansky, P [Nuclear Power Plant Research Inst. Trnava Inc., Trnava (Slovakia)

    1997-09-01

    The present state of art in WWER 440 RPV integrity assessment with respect to PTS events utilized in Slovakia is reviewed briefly in this paper. Recent results of some PTS`s (very severe) analyses, shortly described in our paper, have confirmed the necessity of elaboration of new more sophisticated procedures again. Such methodology should be based on prepared IAEA Guidance. (author). 13 refs, 1 fig.

  2. Performance analysis of WWER-440/230 nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-01-01

    This report examines one particular design, the WWER-440/230, the first generation of commercial WWERs, essentially comparable to the western PWR. This design was installed widely in eastern Europe with a total of 16 unites being completed in what are now Armenia, Bulgaria, Germany (the former German Democratic Republic) the Slovak Republic and Russia. The plants in Armenia and Germany (the former German Democratic Republic) have been closed down, but particularly in Bulgaria and to a lesser extent the Slovak Republic the remaining plants supply a significant proportion of the electricity of the country and decisions to close them could not be taken lightly. The aim of this report is twofold: first to determine whether the impression given by these good overall performance indicators is confirmed using more detailed indicators covering a wide range of factors; second, to see to what extent good performance can be attributed to the industrial and institutional environment in which these plants were designed, built and operated. Particular attention is paid to identifying factors that may impact the quality of the service provided, especially those factors under management control which can be strongly influenced by current and future policy changes and those factors that are beyond the plant management control but could have influenced the performance of the power plants. Issues concerning the safety of these plants are of considerable importance, but they remain outside the scope of this report. Conclusions and recommendations formulated by the IAEA related to WWER safety are contained in the series of reports prepared in the framework of the Extrabudgetary Programme on WWER Safety. A programme progress report was published in 1994 (IAEA-TECDOC-773). Refs, figs, tabs.

  3. Control assembly ejection accident analysis for WWER-440 (Armenian NPP)

    International Nuclear Information System (INIS)

    Bznuni, S.; Malakyan, Ts.; Amirjanyan, A.; Ghasabyan, L.

    2007-01-01

    Control Assembly ejection in WWER-440 initiated by the loss of integrity of the Control Assemblies drive housing has been analyzed. This event causes a very rapid reactivity insertion to the core and small break LOCA which potentially could lead to rapid power increase and redistribution of heat release in the core resulting in a fuel, cladding and coolant temperature rise; primary pressure increase, radiological consequences due to loss of primary coolant and potential loss of cladding integrity and fuel disintegration (if applicable). Methodology of the analysis is based on conservative assumptions as well as on deterministic approach for selection of functioning logic of systems and equipment's to maximize reactor core power and minimize power decreasing reactivity feedback. Computational analyses were performed by 3D kinetics PARCS-RELAP coupled code. WWER-440 fuel cross-section libraries, diffusion coefficients and kinetics parameters were calculated by HELOS code. In this paper analysis of accident for Hot Full Power was presented. Results of analysis show that ANPP WWER-440 reactor design meets acceptance criteria prescribed for RIA type design based accidents (Authors)

  4. Consideration of severe accidents in design of advanced WWER reactors

    International Nuclear Information System (INIS)

    Fedorov, V.G.; Rogov, M.F.; Podshibyakin, A.K.; Fil, N.S.; Volkov, B.E.; Semishkin, V.P.

    1998-01-01

    Severe accident related requirements formulated in General Regulations for Nuclear Power Plant Safety (OPB-88), in Nuclear Safety Regulations for Nuclear Power Stations' Reactor Plants (PBYa RU AS-89) and in other NPP nuclear and radiation guides of the Russian Gosatomnadzor are analyzed. In accordance with these guides analyses of beyond design basis accidents should be performed in the reactor plant design. Categorization of beyond design basis accidents leading to severe accidents should be made on occurrence probability and severity of consequences. Engineered features and measures intended for severe accident management should be provided in reactor plant design. Requirements for severe accident analyses and for development of measures for severe accident management are determined. Design philosophy and proposed engineered measures for mitigation of severe accidents and decrease of radiation releases are demonstrated using examples of large, WWER-1000 (V-392), and medium size WWER-640 (V-407) reactor plant designs. Mitigation of severe accidents and decrease of radiation releases are supposed to be conducted on basis of consistent realization of the defense in depth concept relating to application of a system of barriers on the path of spreading of ionizing radiation and radioactive materials to the environment and a set of engineered measures protecting these barriers and retaining their effectiveness. Status of fulfilled by OKB Gidropress and other Russian organizations experimental and analytical investigations of severe accident phenomena supporting design decisions and severe accident management procedures is described. Status of the works on retention of core melt inside the WWER-640 reactor vessel is also characterized

  5. 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)

  6. Verification of RRC Ki code package for neutronic calculations of WWER core with GD

    International Nuclear Information System (INIS)

    Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.

    2001-01-01

    The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)

  7. Upgrading of seismic capacity of WWER equipment including reactor control rods and distribution systems

    International Nuclear Information System (INIS)

    Kostarev, V.

    1993-01-01

    This paper concerns the experience of CKTI VIBROSEISM firm (CVS) in seismic capacity upgrading of the WWER equipment and pipings. During the last 15 years CVS has accumulated a lot of experimental and analysis results on many of WWER units 'in former Soviet Union. That might be very useful in seismic protecting of East European Nuclear Power Plants. (author)

  8. 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

  9. 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)

  10. 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

  11. 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

  12. Surveillance extension experience at WWER-440 type reactors

    International Nuclear Information System (INIS)

    Gillemot, F.; Uri, G.; Oszwald, F.; Trampus, P.

    1993-01-01

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs

  13. Data presentation in the WWER-440 Basic Principle Simulator

    International Nuclear Information System (INIS)

    Bota, J.; Gossanyi, A.; Vegh, E.

    1986-09-01

    At the Central Research Institute for Physics, Budapest, Hungary, a Basic Principle Simulator for WWER-440 nuclear power plants is under development. The input/output subsystem of the simulator is described. This subsystem is a control desk-shaped special periphery representing each main parameter of the model. During development special attention was paid to the hardware support of the controllers as this part of the model is regarded as most time consuming. A short summary on the hardware configuration of the simulator is also presented. (author)

  14. Radiation stability and recovery of WWER-440 materials

    International Nuclear Information System (INIS)

    Amaev, A.; Kryukov, A.; Levit, V.; Platonov, P.; Sokolov, M.

    1993-01-01

    The main results of a complex investigation of radiation embrittlement of WWER-440 reactor vessel materials, carried out in Russia, are presented. The effect of the annealing temperature and annealing time, neutron fluence, and phosphorous and copper impurity contents on the recovery of the ductile-to-brittle transition temperature are studied. It is shown that the recovery of the transition temperature depends mainly on the annealing temperature. At an annealing temperature of 420 and 460 C, residual post-annealing embrittlement does not depend on neutron fluence. 14 figs., 3 tabs

  15. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  16. Automatic accounting of nuclear materials at WWER type reactor NPPs

    International Nuclear Information System (INIS)

    Babaev, N.S.; Poznyakov, N.L.; Strelkov, D.F.

    1978-01-01

    The possibilities of automatic accounting of nuclear materials at NPPs based on WWER reactors are considered. Organizational and technical principles of an automated system of accounting that takes into consideration IAEA requirements in conducting accounting documentation are proposed. A program is described for accounting materials using a BESM-6 computer. Operation of the program requires that all accounting data be recorded on conventional carriers of computer information (magnetic tapes, discs, perforated cards), which constitute the basic NPP accounting documents and may be directly used as initial data for a corresponding information program

  17. Temperature and pressure instrumentation in WWERs and their testing

    International Nuclear Information System (INIS)

    Por, G.

    1998-01-01

    A description of WWER model V-213 reactors of second generation is presented and compared to analogous NPPs including description of temperature and pressure instrumentation which was tested at Paks NPP. From the experimental results it was concluded that measured response of in core neutron detector to bubbles strongly depends on the relative position of detector and point bubble injection. Neutron noise spectra show characteristic sink when the origin of bubbles is close to the detectors. Dependence of phase behaviour on the boiling conditions is included as well

  18. 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.

  19. 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.

  20. Radiation stability and recovery of WWER-440 materials

    Energy Technology Data Exchange (ETDEWEB)

    Amaev, A; Kryukov, A; Levit, V; Platonov, P; Sokolov, M

    1994-12-31

    The main results of a complex investigation of radiation embrittlement of WWER-440 reactor vessel materials, carried out in Russia, are presented. The effect of the annealing temperature and annealing time, neutron fluence, and phosphorous and copper impurity contents on the recovery of the ductile-to-brittle transition temperature are studied. It is shown that the recovery of the transition temperature depends mainly on the annealing temperature. At an annealing temperature of 420 and 460 C, residual post-annealing embrittlement does not depend on neutron fluence. 14 figs., 3 tabs.

  1. Surveillance extension experience at WWER-440 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gillemot, F; Uri, G [Budapesti Mueszaki Egyetem, Budapest (Hungary); Oszwald, F; Trampus, P

    1994-12-31

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs.

  2. WWER type reactors used as multipurpose nuclear power sources

    International Nuclear Information System (INIS)

    Fiala, J.; Mulak, J.

    1976-01-01

    Safety aspects are assessed of the siting of nuclear power installations in the vicinity of large housing estates and in areas with a high population density, mainly the aspect of the liquidation of the consequences of the maximum credible accident, i.e., the transversal rupture of the primary coolant circuit. The application of WWER type reactors as multipurpose nuclear power sources in Czechoslovakia is justified. It is shown that such a multipurpose nuclear power source differs from a purely condensation nuclear power plant mainly in the design of the secondary stage. The possibilities of such projects are indicated with a view to power and heat operation. (F.M.)

  3. Burnup credit implementation in WWER spent fuel management systems: Status and future aspects

    International Nuclear Information System (INIS)

    Manolova, M.

    1998-01-01

    This paper describes the motivation for possible burnup credit implementation in WWER spent fuel management systems in Bulgaria. The activities being done are described, namely: the development and verification of a 3D few-group diffusion burnup model; the application of the KORIGEN code for evaluation of WWER fuel nuclear inventory during reactor core lifetime and after spent fuel discharge; using the SCALE modular system (PC Version 4.1) for criticality safety analyses of spent fuel storage facilities. Future plans involving such important tasks as validation and verification of computer systems and libraries for WWER burnup credit analysis are shown. (author)

  4. Modernization of the WWER 440/230 nuclear power plant environmental protection system

    Energy Technology Data Exchange (ETDEWEB)

    Mikheev, N.V.; Kamenskaya, A.N.; Kulyukhin, S.A.; Novichenko, V.L.; Rumer, I.A. [Russian Academy of Sciences, Institute of Physical Chemistry, Moscow (Russian Federation); Antonov, B.V.; Kornienko, A.G.; Meshkov, V.M.; Rogov, M.F. [Rosenergoatom Concern, Moscow (Russian Federation)

    2001-07-01

    The papers reports a new approach to the problem of increasing environmental protection during severe accidents at WWER 440/230 nuclear power plants. The environmental protection system that we propose has three, not two protection levels, and can be introduced with minor modernization of the equipment available at WWER 440/230 nuclear power plants: 1. a jet-vortex condenser; 2. the sprinkler system; 3. a sorption module. The proposed modernization not only makes it possible to avoid emergency discharge of radioactive air and steam mix into the environment under any accident scenario, but also would substantially contribute to the safety of WWER 440/230 nuclear power plants. (author)

  5. A new neutron spectrometry system for in-core and out-core application at WWER

    International Nuclear Information System (INIS)

    Mehner, H.C.; Boehmer, B.; Hagemann, U.; Nagel, S.; Schoene, M.; Stephan, I.

    1985-01-01

    An automated measuring system based on a newly-developed multi-component wire activation detector (MWAD) is presented which has been applied for neutron spectrometry inside WWER. The MWAD consisting of Au, Mn, Mo, Ni, and W is designed to determine the neutron spectrum in the thermal, epithermal and fast energy region by a single gamma-ray spectrum measurement. Investigations were carried out within an ordinary fuel assembly of the Greifswald NPS (WWER-440) as well as in experimental fuel assembly of the Rheinsberg NPS (WWER-70). The evaluation of the measurements was done with the newly-developed adjustment code COSA based on the generalized least square method. (author)

  6. 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

  7. 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

  8. Power peak in vicinity of WWER-440 control rod at end of fuel cycle

    International Nuclear Information System (INIS)

    Mikus, J.

    2003-01-01

    This paper presents some results of the axial power distribution measurements carried out in a WWER-440 type core on the light-water, zero-power reactor LR-O in the vicinity of the WWER-440 control rod model at zero boron concentration in moderator. Further presented information concern the description of the control rod model, LR-0 core arrangement, specification of the fuel assemblies and measurement conditions. The aim of performed experiment is enlargement of the available 'power peaking database' to enable the calculation codes validation also by means of data that correspond to the end of WWER-440 fuel cycle (Authors)

  9. 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.

  10. 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)

  11. Structural and leakage integrity assessment of WWER steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K.; Otruba, J. [Nuclear Research Inst., Rez (Switzerland)

    1997-12-31

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

  12. 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.

  13. WWER Steam Generators Tubing Performance and Aging Management

    International Nuclear Information System (INIS)

    Trunov, Nikolay B.; Davidenko, Stanislav E.; Grigoriev, Vladimir A.; Popadchuk, Valery S.; Brykov, Sergery I.; Karzov, Georgy P.

    2006-01-01

    At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned. (authors)

  14. 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

  15. 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

  16. 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

  17. Use of Main Loop Isolating Valves (GZZS) in WWER 440

    International Nuclear Information System (INIS)

    Stefanova, A.E.; Gencheva, R.V.; Groudev, P.P.

    2002-01-01

    This paper discusses the usage of Main Loop Isolation Valves in case of Steam Generator Tube Rupture accident in WWER440/V230. A double-ended single pipe break in SG-6 was chosen as representative. In the paper are investigated two cases. In the first one the operator isolates the affected loop by Main Loop Isolation Valves closing and after primary depressurization re-opens them to cooldown the damaged Steam Generator. The second case treats the situation, where Main Loop Isolation Valves fail to close with the necessary operator actions for managing plant recovery. RELAP5/MOD3.2 computer code has been used to simulate the Steam Generator Tube Rupture accident in WWER440 NPP model. This model was developed and validated at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences. The results of analyses presented in this report demonstrate that in the both cases (with or without Main Loop Isolation Valves usage) the operator could bring the plant to stable and safety conditions (Authors)

  18. Bounding approach in BUC implementation in pool at WWER-440

    International Nuclear Information System (INIS)

    Havluj, F.

    2006-01-01

    As new fuel designs (with higher enrichment) are introduced, spent fuel storage facilities might not fulfill criticality safety criteria when using fresh fuel approach to the criticality analyses. Since optimum moderation conditions evaluation even in wet storage systems is required by some regulatory bodies, any credit for soluble boron cannot be taken. Thus, the only suitable way to prove subcriticality of the given spent fuel system with higher enriched/burnt fuel is burnup credit implementation. This paper outlines burnup credit implementation methodology as demonstrated on criticality evaluation of WWER-440 reactor pool at NPP Dukovany. Operational history effects, isotopic set choice, as well as computational issues (SCALE 4.4a was used both for depletion and criticality calculations) are discussed, maintaining strictly conservative approach. Bounding approach in operational history treatment was carefully examined. Criticality evaluation using selection of (as expected) conservative values of operational parameters (specific power, fuel and moderator temperatures, boron content in moderator,..) was compared to criticality evaluation of real fuel assemblies from the NPP database. Therefore, bounding approach was justified and it was shown that it is not excessively conservative. Presented methodology can be applied on any similar spent fuel facility. Suggestions for future research are noted (mainly end-effect evaluation and consideration of profiled fuel) and urgent need of validation of depletion codes for WWER systems is emphasized (Authors)

  19. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.

    1997-01-01

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction

  20. Structural and leakage integrity assessment of WWER steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Splichal, K; Otruba, J [Nuclear Research Inst., Rez (Switzerland)

    1998-12-31

    The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

  1. 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.

  2. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Proselkov, V.; Saprykin, V.; Scheglov, A.

    2003-01-01

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  3. Software in support of fuel operation in WWERS

    International Nuclear Information System (INIS)

    Evdokimov, I.A; Novikov, V.V; Ugrumov, A.V; Shishkin, A.A

    2013-01-01

    A software package comprising computer codes and fuel monitoring tools is under development in Russia in support of WWER fuel operation. The software package includes an expert computer system designed for failure diagnosis in course of reactor operation, prediction of activity evolution in primary coolant and express analysis of pellet-to-cladding mechanical interaction (PCMI) on rod-by-rod basis under normal and transient modes of operation. Coupled with the expert system, the first version of a graphical interface computer program is developed for NPP operating bodies. One of the features of this program is to launch automatically a fuel performance code for a series of detailed calculations for fuel rods with severe PCMI. The particular rods for calculations are determined by the expert system during the express core analysis. A greater attention is paid to recent results in prediction of fuel behavior after a primary failure has occurred. One of the major risks to further operation of leaking fuel comes from secondary fuel degradation due to massive cladding hydriding. Threshold conditions for initiation of secondary hydriding have been found on the basis of physical modeling. Final criteria of secondary failure occurrence were deduced by applying the model to analysis of post-irradiation examinations of leaking WWER fuel. (authors)

  4. Strategy for assessment of WWER steam generator tube integrity. Report prepared within the framework of the coordinated research project on verification of WWER steam generator tube integrity

    International Nuclear Information System (INIS)

    2007-12-01

    Steam generator heat exchanger tube degradations happen in WWER Nuclear Power Plant (NPP). The situation varies from country to country and from NPP to NPP. More severe degradation is observed in WWER-1000 NPPs than in case of WWER-440s. The reasons for these differences could be, among others, differences in heat exchanger tube material (chemical composition, microstructure, residual stresses), in thermal and mechanical loadings, as well as differences in water chemistry. However, WWER steam generators had not been designed for eddy current testing which is the usual testing method in steam generators of western PWRs. Moreover, their supplier provided neither adequate methodology and criteria nor equipment for planning and implementing In-Service Inspection (ISI). Consequently, WWER steam generator ISI infrastructure was established with delay. Even today, there are still big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment (plugging criteria for defective tubes vary from 40 to 90% wall thickness degradation). Recognizing this situation, the WWER operating countries expressed their need for a joint effort to develop methodology to establish reasonable commonly accepted integrity assessment criteria for the heat exchanger tubes. The IAEA's programme related to steam generator life management is embedded into the systematic activity of its Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP). Under the advice of the TWG-LMNPP, an IAEA coordinated research project (CRP) on Verification of WWER Steam Generator Tube Integrity was launched in 2001. It was completed in 2005. Thirteen organizations involved in in-service inspection of steam generators in WWER operating countries participated: Croatia, Czech Republic, Finland, France, Hungary, Russian Federation, Slovakia, Spain, Ukraine, and the USA. The overall objective was to

  5. 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

  6. 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

  7. 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

  8. Code Package to Analyze Parameters of the WWER Fuel Rod. TOPRA-2 Code - Verification Data

    International Nuclear Information System (INIS)

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

    2009-01-01

    Presented are the data for computer codes to analyze WWER fuel rods, used in the WWER department of RRC 'Kurchatov Institute'. Presented is the description of TOPRA-2 code intended for the engineering analysis of thermophysical and strength parameters of the WWER fuel rod - temperature distributions along the fuel radius, gas pressures under the cladding, stresses in the cladding, etc. for the reactor operation in normal conditions. Presented are some results of the code verification against test problems and the data obtained in the experimental programs. Presented are comparison results of the calculations with TOPRA-2 and TRANSURANUS (V1M1J06) codes. Results obtained in the course of verification demonstrate possibility of application of the methodology and TOPRA-2 code for the engineering analysis of the WWER fuel rods

  9. Seismic upgrading of WWER 440-230 structures, units 1/2, Kozloduy NPP

    International Nuclear Information System (INIS)

    Stafanov, D.; Kostov, M.; Boncheva, H.; Varbanov, G.

    1995-01-01

    The purpose of this paper is to present final results from a big amount of computational work in connection with the investigations of the possibilities for upgrading of WWER 440-230 structures, units 1/2, Kozloduy NPP. (author)

  10. 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)

  11. 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.

  12. 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.

  13. 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

  14. Experience of developing the imitators of the fuel element for the WWER reactors

    International Nuclear Information System (INIS)

    Balashov, S.M.; Boltenko, Eh.A.; Vinogradov, V.A.

    1998-01-01

    Peculiarities of designs of fuel elements imitators for the WWER-type reactors of nominal capacity and with single-ended current feed positioning are considered. The data on the filler heat conductivity and the results of tests and application of the fuel elements imitators at various testing facilities are presented. The possibility of equipping one of the non operating WWER reactors with the fuel element imitators for conduct of large-scale experiment is indicated

  15. Results of experimental investigations for substantiation of WWER cermet fuel pin performance

    International Nuclear Information System (INIS)

    Popov, V.V.; Karpin, A.D.; Isupov, I.A.; Rumyantsev, V.N.; Troyanov, V.M.; Subonyaev, V.N.; Melnichenko, N.A.

    1997-01-01

    The out-of-pile experiment results on interaction of the cladding and matrix materials and uranium dioxide at cermet fuel temperature for normal operating conditions of the WWER-440 reactor are analyzed. Cermet fuel element behaviour under the maximum designed damage of the WWER-440 reactor is considered. In the AM reactor loop a fission product output from the unsealed cermet fuel elements have been studied. (author). 6 figs, 3 tabs

  16. Assessment of the influence of design limits to the economics of WWER fuel cycle

    International Nuclear Information System (INIS)

    Dementiev, V.G.; Shishkov, L.K.

    2010-01-01

    The paper discusses the influence of the reactor parameters limits for normal operation on the economical performance of WWER fuel cycles. It is shown for the typical WWER fuel cycles that decreasing the limits for the main power distribution parameters to 10% leads to decreasing the fuel components of the electricity cost price up to 4-5%. As the nowadays limitations are reached the dependence becomes weaker. (Authors)

  17. Problems of control of WWER-type pressurized water reactors (PWR's)

    International Nuclear Information System (INIS)

    Drab, F.; Grof, V.

    1978-01-01

    The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

  18. Guidelines for prediction of irradiation embrittlement of operating WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC has been developed under an International Atomic Energy Agency Coordinated Research Project (CRP) entitled Evaluation of Radiation Damage of WWER Reactor Pressure Vessels (RPV) using Database on RPV Materials to develop the guidelines for prediction of radiation damage to WWER-440 PRVs. The WWER-440 RPV was designed by OKB Gidropress, Russian Federation, the general designer. Prediction of irradiation embrittlement of RPV materials is usually done in accordance with relevant codes and standards that are based on the large amounts of information from surveillance and research programmes. The existing Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than twenty years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. Nevertheless, it is still in use and generally consistent with new data. The present publication presents the analyses using all available data required for more precise prediction of radiation embrittlement of WWER-440 RPV materials. Based on the fact that it contains a large amount of data from surveillance programmes as well as research programmes, the IAEA International Database on RPV Materials (IDRPVM) is used for the detailed analysis of irradiation embrittlement of WWER RPV materials. Using IDRPVM, the guideline is developed for assessment of irradiation embrittlement of RPV ferritic materials as a result of degradation during operation. Two approaches, i.e. transition temperatures based on Charpy impact notch toughness, as well as based on static fracture toughness tests, are used in RPV integrity evaluation. The objectives of the TECDOC are the analysis of irradiation embrittlement data for WWER- 440 RPV materials using IDRPVM database, evaluation of predictive formulae depending on chemical composition of the material, neutron fluence, flux, and

  19. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    International Nuclear Information System (INIS)

    Vitkova, M.; Manolova, M.; Stefanova, S.; Simeonova, V.; Passage, G.; Lassmann, K.

    1994-01-01

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: 1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; 2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; 3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs

  20. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Vitkova, M; Manolova, M; Stefanova, S; Simeonova, V; Passage, G [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Kharalampieva, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Lassmann, K [European Atomic Energy Community, Karlsruhe (Germany). European Inst. for Transuranium Elements

    1994-12-31

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: (1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; (2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; (3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs.

  1. Review of WWER fuel and material tests in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.; Kolstad, E.

    2006-01-01

    A review of the tests with WWER fuels and materials conducted in HBWR over the years of cooperation with Russia is presented. The first test with old generation WWER-440 fuel and PWR specification fuel was carried out from 1995 to 1998. Some differences between these fuels regarding irradiation induced densification and pellet design as well as similar fuel thermal behaviour, swelling and FGR were revealed during the test. The data from this test are reviewed and compared with PIE recently performed to confirm the in-pile measurements. The second test was started in March 1999 with the main objective to study different modified WWER fuels also in comparison with PWR fuel. The results indicated that all these modified WWER fuels exhibit improved densification properties relative to earlier tested fuel. In-pile data on fuel densification have been analysed with respect to as fabricated fuel microstructure and can be used for verification of fuel behaviour models. Corrosion and creep tests in the Halden reactor encompass WWER cladding alloys and some results are given. Prospective WWER fuel and material tests foreseen within the frame of the joint program of OECD HRP are also presented. (authors)

  2. VUJE's experience in the field of thermal-hydraulic behaviour of WWER

    International Nuclear Information System (INIS)

    Klepach, J.

    1995-01-01

    The thermal-hydraulic behavior (THB) of NPP coolant system and its consequences to nuclear safety of WWER reactors in previous Czechoslovakia has been studied in the VUJE (Nuclear Power Plants Research Institute, Trnava, SK). The institute takes part in the development and verification of its own (SLAP, LENKA, PUMKO, SICHTA, TRACO etc.) and international (DYNAMIKA5) codes for thermal-hydraulic analysis. The verification efforts are concentrated on the WWER specific features such as horizontal steam generators, control and safety system functioning, etc. The whole range of NPP accident analyses is covered by the VUJe staff. The author outlined briefly the WWER specific features as design and implemented improvements in Bohunice V-1 and Mochovce V-1 (WWER 230 model). The pros and cons of the WWER design compared against western type PWR are described. It is believed that although the WWERs are designed under the rules and standards of 1960s, their safety and operational performance can be improved to acceptable level by thorough analysis and appropriate measures. 5 figs

  3. 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

  4. Safety and upgrading of nuclear power plants with WWER-440/V-230; Voprosy bezopasnosti i modernizatsii AEhS s WWER-440 (V-230)

    Energy Technology Data Exchange (ETDEWEB)

    Melnikov, N [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1996-12-31

    Modernization of WWER-440/V-230 reactors is considered according to the WANO project `Proposal for reconstruction of WWER-440/V-230 reactors` (1990). The main technical measures to be taken relate to: 1. Expanding the considered spectrum of design basis accidents; 2. Provision of first loop integrity; 3. Safety system structure and components reliability enhancement; 4. Reduction o initiating events probability; 5. Enhancement of hermeticity and integrity of the area surrounding the first loop; 6. Fire fighting system upgrade; 7. Radiation control system upgrade; 8. Provision of seismic resistance. Details of the recommended technical steps for some NPPs are presented. 1 tab.

  5. Risk assessment basis for WWER-440 spent nuclear fuel

    International Nuclear Information System (INIS)

    Lascek, M.; Necas, V.; Darilek, P.

    2000-01-01

    The most problematic part of nuclear fuel cycle is its back end. Various high level waste management are available or under development (final disposal of spent assemblies in deep repository, reprocessing, partitioning, transmutation,...). Application of any method is connected with production of characteristic high level waste (amount, radio-toxicity, form,...) as well as various risk level for the environment and mankind. Strategy selection should be based on risk analysis also. The paper deals with assessment of risk, that is associated with WWER-440 spent fuel inventory. In order to evaluate the risk, the accumulated amount of the radioactive inventory is calculated and the decay of the long-lived radionuclides is computed by ORIGEN code. Analysis is oriented on calculation of hazard indexes for assessing the relative hazards of actinides, toxic and long-lived radionuclides. (Authors)

  6. Accident transient processes at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    1982-01-01

    Thermal-physical and nuclear-physical transient processes at NPPs with the WWER type reactors during accidents with the main technological equipment failures and the accidents with loss of coolant in the primary and secondary coolant circuits are considered. Mathematical methods used for these processes modelling is described. Examples of concrete calculations for accidents with different failures are given. Comparative analysis of the results of dynamic tests at the Novo-Voronezh-3 reactor is presented. It is concluded that the modern NPP design is impossible without application of mathematical modelling methods. The mathematical modelling of transients is also necessary for proper and safe NPP operation. Mathematical modelling of accidents at NPPs is a comparatively new method of investigation. Its success and development are completely based on the progress in modern computer development. With their improvement the mathematical models will become more complicate and adequacy of real physical process representation by their means will increase

  7. Improved anchorage of steel linings at WWER nuclear power plants

    International Nuclear Information System (INIS)

    Bouda, M.; Cepicky, J.

    1987-01-01

    Three variants were tested of anchoring the steel lining of special structures of nuclear power plants with WWER reactors, such as the containment, viz.: point anchorage with pins according to a Czechoslovak design, and two methods of linear anchorage according to a Soviet design and a design by the Institute for Industrial Construction of the GDR. Graphically shown are some results of the tests of the point anchorage with pins that showed the reliability of the method with respect to deformation and strain of the lining and of the anchoring elements at a lining temperature of up to 300 degC, which significantly exceeded the design-basis accident temperature of 150 degC. The advantages of the variant include a lower content of labour and an increased quality of welding jobs thanks to the application of the Nelson semi-automatic welding technique. (Z.M.). 10 figs., 7 refs

  8. 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)

  9. Inovation of the computer system for the WWER-440 simulator

    International Nuclear Information System (INIS)

    Schrumpf, L.

    1988-01-01

    The configuration of the WWER-440 simulator computer system consists of four SMEP computers. The basic data processing unit consists of two interlinked SM 52/11.M1 computers with 1 MB of main memory. This part of the computer system of the simulator controls the operation of the entire simulator, processes the programs of technology behavior simulation, of the unit information system and of other special systems, guarantees program support and the operation of the instructor's console. An SM 52/11 computer with 256 kB of main memory is connected to each unit. It is used as a communication unit for data transmission using the DASIO 600 interface. Semigraphic color displays are based on the microprocessor modules of the SM 50/40 and SM 53/10 kit supplemented with a modified TESLA COLOR 110 ST tv receiver. (J.B.). 1 fig

  10. Simulator of nuclear power plant with WWER-440 units

    International Nuclear Information System (INIS)

    Krcek, V.

    1985-01-01

    The use is discussed of simulators in the training of qualified personnel for the construction and operation of nuclear power plants. Simulators are used for training all activities and thinking processes related to the control of a nuclear reactor in the course of quasi-steady and non-steady states. The development and implementation is summed up of the construction of such a simulator for WWER-440 nuclear power plants. The main parts of the simulator include the unit control room, the computer system, the teacher's workplace and the interface system. The possibility of simulating the functions of the unit for personnel training is based on the description of the behaviour of the simulated object in form of mathematical models of its basic technological subsystems and their interrelations within the range of operating patterns. (J.C.)

  11. Probabilistic safety analysis second level of WWER-TOI

    International Nuclear Information System (INIS)

    Chekin, A.A.; Bajkova, E.V.; Levin, V.N.; Shishina, E.S.

    2015-01-01

    Probabilistic safety assessment (PSA) of Level-1 and Level-2 gives a comprehensive qualitative and quantitative evaluation of the safety of the project. The operation of the unit at rated power is considered. As sources of radioactivity in the development of the second-level PSA, nuclear fuel in the core of the reactor is considered. As initiating events, internal initiating events (including de-energizing) are considered, which may arise due to failures of NPP systems, equipment or components, or due to erroneous actions of personnel. In general, an assessment of the level of project safety shows that the WWER-TOI project complies with the requirements of the TOR, as well as all the requirements of modern Russian and foreign regulatory documents in the field of security [ru

  12. Basic safety principles and practice of WWERs in Hungary

    International Nuclear Information System (INIS)

    Voeroess, L.

    1992-01-01

    The nuclear safety is the actual subject of this presentation and it is considered to be the most important issue, and its permanent improvement is the key responsibility. We share the opinion, that everybody who works in the field of nuclear power generation has to be at such a high level, both in respect of the professional and the moral aspects, which would practically exclude occurrence of accidents causing adverse environmental effects. We are aware that another severe accident occurring in any country of the world would put the whole nuclear industry into a hopeless situation, which - as we have already seen - would seriously influence the Hungarian energy system as well. I try to illustrate in my presentation how can our WWER reactors be evaluated in the highlight of the internationally accepted safety requirements, how safe can they be considered and what can we do in order to ensure at every time the appropriate level of safety. 22 refs, 15 figs

  13. 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)

  14. Development of the code package KASKAD for calculations of WWERs

    International Nuclear Information System (INIS)

    Bolobov, P.A.; Lazarenko, A.P.; Tomilov, M.Ju.

    2008-01-01

    The new version of software package for neutron calculation of WWER cores KASKAD 2007 consists of some calculating and service modules, which are integrated in the common framework. The package is based on the old version, which was expanded with some new functions and the new calculating modules, such as: -the BIPR-2007 code is the new one which performs calculation of power distribution in three-dimensional geometry for 2-group neutron diffusion calculation. This code is based on the BIPR-8KN model, provides all possibilities of BIPR-7A code and uses the same input data; -the PERMAK-2007 code is pin-by-pin few-group multilayer and 3-D code for neutron diffusion calculation; -graphical user interface for input data preparation of the TVS-M code. The report also includes some calculation results obtained with modified version of the KASKAD 2007 package. (Authors)

  15. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  16. 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

  17. WWER nuclear waste management regulatory experience in Finland

    International Nuclear Information System (INIS)

    Varjoranta, Tero

    2000-01-01

    About 30% of all electricity produced in Finland is generated by nuclear power. Four reactors, with a total capacity of 2 656 MW e (net), are currently in operation. At Loviisa, there are two 488 MW e WWER units (recently upgraded 440-units) and at Olkiluoto two 840 MW e BWR units. At the Loviisa plant conditioning, storage and final disposal of low-and intermediate-level wastes from reactor operation will take place at the NPP sites. Intermediate level ion exchange resins and evaporation concentrates are currently stored in tanks. However, a license application for constructing a solidification plant based on cementation is currently under STUKs regulatory review. The construction of the final repository for I/LLW at the Loviisa site was started in 1993 and the Government granted the operating license in 1998. The nuclear legislation requires disposal of spent fuel into the Finnish bedrock. (Authors)

  18. WWER core pattern enhancement using adaptive improved harmony search

    Energy Technology Data Exchange (ETDEWEB)

    Nazari, T. [Nuclear Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Aghaie, M., E-mail: M_Aghaie@sbu.ac.ir [Nuclear Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Norouzi, A. [Nuclear Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer The classical and improved harmony search algorithms are introduced. Black-Right-Pointing-Pointer The advantage of IHS is demonstrated in Shekel's Foxholes. Black-Right-Pointing-Pointer The CHS and IHS are compared with other Heuristic algorithms. Black-Right-Pointing-Pointer The adaptive improved harmony search is applied for two cases. Black-Right-Pointing-Pointer Two cases of WWER core are optimized in BOC FA pattern. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Core performance analysis constitutes an essential phase in core fuel management optimization. Finding an optimum core arrangement for loading of fuel assemblies, FAs, in a nuclear core is a complex problem. In this paper, application of classical harmony search (HS) and adaptive improved harmony search (IHS) in loading pattern (LP) design, for pressurized water reactors, is described. In this analysis, finding the best core pattern, which attains maximum multiplication factor, k{sub eff}, by considering maximum allowable power picking factors (PPF) is the main objective. Therefore a HS based, LP optimization code is prepared and CITATION code which is a neutronic calculation code, applied to obtain effective multiplication factor, neutron fluxes and power density in desired cores. Using adaptive improved harmony search and neutronic code, generated LP optimization code, could be applicable for PWRs core with many numbers of FAs. In this work, at first step, HS and IHS efficiencies are compared with some other heuristic algorithms in Shekel's Foxholes problem and capability of the adaptive improved harmony search is demonstrated. Results show, efficient application of IHS. At second step, two WWER cases are studied and then IHS proffered improved core patterns with regard to mentioned objective functions.

  19. Assesment of integrity of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Brozova, A.; Zdarek, J.

    1992-01-01

    Full text: The leak rates measurement project was held to give experimental data enabling the Czechoslovak Atomic Agency Inspection to decree the change in the Technical Specification allowable limit of steam generator activity release on secondary side. The WWER types of nuclear power plants in Czechoslovakia have horizontal steam generators. The tubes studying in frame of the project belong to steam generator WWER- 440 type, the diameter of tube is 16 mm, the wall thickness 1.4 mm. The subject of the project was the measurement of service leak rates of typical in service cracks. Secondary side stress corrosion cracks were determined as the typical crack created in service condition. These cracks were prepared in tubes artificially by exposition in chloride environment accompanied by an internal stress. The experimental device consisted of a pressure vessel connected with pressure water loop, a cooling vessel for leakage medium and a measuring vessel. The leak rates were determined as a slope of plots the leakage volume - time. Inside the pressure vessel the steam generator operation environment was simulated. It means: primary side of tube 12.5 MPa, Z90 deg. C, secondary side -4.6MPa, 250 deg. C, water service quality. We observed reduce of leak rate in course of time in each experiment. We suppose the tubes were stopped up by deposits formed in manufacturing of crack and in experiment. Our opinion has been proved by fractography. Project results in recommendation for in service leak rate limit based on safety factors with respect to critical crack lengths and for determination of tube plugging criteria. (author)

  20. Analysis of WWER-440 fuel performance under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  1. WWER core pattern enhancement using adaptive improved harmony search

    International Nuclear Information System (INIS)

    Nazari, T.; Aghaie, M.; Zolfaghari, A.; Minuchehr, A.; Norouzi, A.

    2013-01-01

    Highlights: ► The classical and improved harmony search algorithms are introduced. ► The advantage of IHS is demonstrated in Shekel's Foxholes. ► The CHS and IHS are compared with other Heuristic algorithms. ► The adaptive improved harmony search is applied for two cases. ► Two cases of WWER core are optimized in BOC FA pattern. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Core performance analysis constitutes an essential phase in core fuel management optimization. Finding an optimum core arrangement for loading of fuel assemblies, FAs, in a nuclear core is a complex problem. In this paper, application of classical harmony search (HS) and adaptive improved harmony search (IHS) in loading pattern (LP) design, for pressurized water reactors, is described. In this analysis, finding the best core pattern, which attains maximum multiplication factor, k eff , by considering maximum allowable power picking factors (PPF) is the main objective. Therefore a HS based, LP optimization code is prepared and CITATION code which is a neutronic calculation code, applied to obtain effective multiplication factor, neutron fluxes and power density in desired cores. Using adaptive improved harmony search and neutronic code, generated LP optimization code, could be applicable for PWRs core with many numbers of FAs. In this work, at first step, HS and IHS efficiencies are compared with some other heuristic algorithms in Shekel's Foxholes problem and capability of the adaptive improved harmony search is demonstrated. Results show, efficient application of IHS. At second step, two WWER cases are studied and then IHS proffered improved core patterns with regard to mentioned objective functions.

  2. 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)

  3. 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)

  4. 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)

  5. Ensuring the operational safety of nuclear power plants with WWER reactors

    International Nuclear Information System (INIS)

    Shasharin, G.A.; Veretennikov, G.A.; Abagyan, A.A.; Lesnoj, S.A.

    1984-01-01

    At the start of 1983, 27 nuclear power producing units with reactor facilities of the WWER type were in operation in the Soviet Union and other countries. In 1982 the average load factor for nuclear power plants with WWER reactors was 73 per cent. There was not a single nuclear accident or even damage with any significant radiation consequences in the WWER reactors during the entire period of their operation. The most modern nuclear power plants with WWER-440 and WWER-1000 reactors meet all present-day international requirements. Safe operation of the plants is achieved by a variety of measures, the most important of which include: procedures for increasing the reliability of plant equipment and systems; ensuring exact compliance with plant operating instructions; ensuring reliable operation of plant safety systems; action directed towards maintaining the skills of plant personnel at a level adequate to ensure the taking of proper action during transient processes and accident situations. The paper discusses concrete steps for ensuring safe nuclear power plant operation along these lines. In particular, measures such as the following are described: the use of a system for collecting and processing information on equipment failures and defects; the development and introduction of methods of early defect diagnosis; the performance of complex testing of safety systems; the training of highly skilled personnel for nuclear power plants at educational combines and at teaching and training centres making use of simulators; arranging accident-prevention training and special instruction for personnel. (author)

  6. 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

  7. 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)

  8. Assessment of nuclear data needs for broad-group SCALE library related to WWER spent fuel applications

    International Nuclear Information System (INIS)

    Zalesky, K.; Markova, L.

    1999-12-01

    A preliminary study aimed at the issue of feasibility to generate a broad-group SCALE library related to WWER spent fuel applications was made. The SCALE code system has been installed and is being used in many countries operating WWER-type reactors for criticality and shielding analyses as well as spent fuel isotopic inventory calculations but still without an extensive validation and verification for the WWER environment. This study should be a contribution to QA connected with the SCALE code system application for the WWER calculations as a basis on which the generation of the specific WWER SCALE library can be prepared. Possible ways of the broad-group library development are described. (author)

  9. 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

  10. 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)

  11. 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

  12. Results of noise analysis in the WWER 440 - type nuclear power plant Dukovany

    International Nuclear Information System (INIS)

    Runkel, J.; Stegemann, D.; Fiedler, J.; Hrosso, P.; Figedy, S.; Sadilek, J.; Hulin, J.

    1996-01-01

    Results of a common work in the frame of scientific and technological cooperation between the Slovak Republic and Germany in the field of development of diagnostic methods to improve the safety and availability of WWER-type reactors are presented. Signals of the standard diagnostic instrumentation of the WWER-type reactors of units 1,2 and 4 at the nuclear power plant Dukovany (EDU) in Czech Republic were analyzed. Mechanical vibrations like pendular and vertical movements of the reactor pressure vessel and its internals as well as thermohydraulic fluctuations like fluid resonances or standing pressure waves could be identified in all the three units. Incipient changes of the mechanical and/or thermohydraulic conditions of the core can be detected by periodical analysis of the signals of the standard reactor instrumentation of WWER-types reactors. (authors)

  13. Integration of post-irradiation examination results of failed WWER fuel rods

    International Nuclear Information System (INIS)

    Smirnov, A.; Markov, D.; Smirnov, V.; Polenok, V.; Perepelkin, S.

    2003-01-01

    The aim of the work is to investigate the causes of WWER fuel rod failures and to reveal the dependence of the failed fuel rod behaviour and state on the damage characteristics and duration of their operation in the core. The post-irradiation examination of 12 leaky fuel assemblies (5 for WWER-440 and 7 for WWER-1000) has been done at SSC RF RIAR. The results show that the main mechanism responsible for the majority of cases of the WWER fuel rod perforation is debris-damage of the claddings. Debris fretting of the claddings spread randomly over the fuel assembly cross-section and they are registered in the area of the bundle supporting grid or under the lower spacer grids along the fuel assembly height. In the WWER fuel rods, the areas of secondary hydrogenating of cladding are spaced from the primary defects by ∼2500-3000 mm, as a rule, and are often adjacent closely to the upper welded joints. There is no pronounced dependence of the distance between the primary and secondary cladding defects neither on the linear power, at which the fuel rods were operated, nor on the period of their operation in the leaky state. The time period of the significant secondary damage formation is about 250 ± 50 calendar days for the WWER fuel rods with slight through primary defects (∼0.1 - 0.5 mm 2 ) operated in the linear power range 170-215 W/cm. Cladding degradation, taking place due to the secondary hydrogenating, does not occur in case of large through debris-defects during operation up to 600 calendar days

  14. 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.

  15. 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)

  16. The ames network and the task group on WWER's

    International Nuclear Information System (INIS)

    Davies, L.M.; Duysen, J.C. van; Estorff, U. von; Sycamore, D.

    1997-01-01

    The European Network on 'Ageing Materials Evaluation and Studies' (AMES) was created in 1993. Its main objectives are (a) to provide information and understanding on neutron irradiation effects in reactor materials in support of designers, operators, regulators and researchers and (b) to establish and discharge projects in the above areas. The Steering Committee is composed of at least one participant from each nuclear European Union country. The JRC's Institute for Advanced Materials of the European Commission plays the role of Operating Agent and Manager of the AMES Network. This paper describes the structure, objectives, and major projects of the AMES network. Particular emphasis is placed upon the work it is intended to perform within the Task Group on 'WWER's of the first AMES project (AMES1) on 'Validation of surveillance practice and mitigation methods'. EC DGXVII is addressing the question of how to facilitate contacts between EU and Russian industries in the framework of nuclear Industrial co-operation, and this project may provide a suitable starting point upon which to develop a basis for further work of mutual interest. (author)

  17. 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.)

  18. Technically feasibility solution for WWER-440/B230. Confinement upgrading

    International Nuclear Information System (INIS)

    Balabanov, E.; Sartmandgiev, A.

    2000-01-01

    A concept for modernization of the units 1 to 4 of Kozloduy NPP ( WWER-440/V230) localization system to cope with a large spectrum of accidents, including 2x100% LOCA is proposed. This concept in foreseeing the installation of several systems, which altogether would form a complex of beyond original design basis accidents mitigation systems. These systems are: Pressure Suppression System (PSS) - with implementation of VNIIAES vortex condenser - supposed to cope with the first pressure peak and to protect the confinement integrity, hydrogen control system-with implementation of Passive Autocatalic Recombiners (PAR) - supposed to cope with hydrogen and to mitigate the possible uncontrolled release of radiation products, active Filter Venting System (FVS) - to maintain underpressure in the confinement and to retain radioactive products, as well as to control radioactive releases during a long period of time. This concept is ensuring: integrity of the localization system for the full spectrum of primary LOCA, non violation of the dose limits, defined by BNSA, for a large spectrum of accidents including severe accidents, fulfillment of IAEA recommendation (INSAG - 3) - decrease the probability for large radioactive release by a factor of 10 outside of NPP side

  19. 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)

  20. 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)

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

    International Nuclear Information System (INIS)

    Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

    1993-12-01

    Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

  6. 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.

  7. Evaluation of efficiency of axial profiling in WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Ananjev, Yu. A.; Kurakin, K. Yu.; Artemov, V.G.; Ivanov, A.S.

    2005-01-01

    The present report deals with consideration of fuel enrichment axial profiling in WWER-440 assemblies. The study is performed on improving the effectiveness of fuel utilization using the example of implementing the axial profiling in the assemblies of the second generation. For simulation of fuel loadings the computer code package SAPFIR 9 5 and RC is used that allows for correct consideration of specific features of assemblies design changes. The methodical approach to assessment of effectiveness of implementing the axial profiling is considered with the use of capabilities of the mentioned code package. In conclusion the recommendations are given on using the fuel enrichment axial profiling in WWER-440 assemblies (Authors)

  8. Nuclear safety improvement activities related to WWER-440 units in Bulgaria

    International Nuclear Information System (INIS)

    Gantchev, T.

    1998-01-01

    The systematic evaluation of the deficiencies of the original design of the WWER reactors brought to the development of a Short Term Programme for Safety Upgrading and Modernisation of Kozloduy WWER-440 units. The implementation of this Programme was completed in 1997. The strive for continuos improvement of Kozloduy Nuclear Power Plant (NPP) safety level, the new requirements of the Bulgarian Nuclear Safety Authority and the public concern initiated the development of new Complex Programme for Safety Improvement (PRG'97), now in a process of implementation. (author)

  9. 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)

  10. 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.

  11. 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

  12. 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)

  13. Study on steam separation in steam generators of a NPP with the WWER-440 reactors

    International Nuclear Information System (INIS)

    Dmitriev, A.I.; Kolzov, Yu.V.; Titov, V.F.; Dubrovin, A.V.; Ilyushin, V.F.; Volkov, A.P.

    1977-01-01

    The separation characteristics as well as the actual level position in steam generators with and without a submerged holy sheet have been determined at a WWER-440 reactor nuclear power plant. It has been shown, that without changing the design of steam generators their load at the WWER-440 reactor nuclear power plant can be increased by about 10%. In this case the vapour humidity does not exceed the permissible value equal to 0.25%. The submerged holy sheet considerably decreases load irregularity and swelling of the water-steam mixture layer

  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. 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

  16. Decommissioning strategy of the operating WWER type units in the Ukraine

    International Nuclear Information System (INIS)

    Litvinsky, L.L.; Lobach, Yu.N.; Skripov, A.E.

    2002-01-01

    At present in Ukraine, 13 WWER type units are in operation and two other ones are in the final stage of construction. Decommissioning of these units is expected after the year 2010. General planning of their decommissioning is developed in the framework of the decommissioning strategy of operating WWER type units. The strategy contains the objectives, principles and main tasks of the decommissioning as well as the activities at each phase of decommissioning. It is considered a broad range of factors important for the planning and implementation of decommissioning. (author)

  17. 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

  18. Development of anti-debris filter for WWER-440 working fuel assembly

    International Nuclear Information System (INIS)

    Kolosovsky, V.; Aksyonov, P.; Kukushkin, Y.; Molchanov, V.; Kolobaev, A.

    2006-01-01

    Mechanical damaging of the fuel rod claddings caused by debris is one of the main reasons for fuel assembly failures. The paper focuses on the program and results of experimental and design activities carried out by Russian organizations relating to the development and investigation of operational characteristics of anti-debris filters for WWER-440 working fuel assemblies. Lead working fuel assemblies equipped with anti-debris filters have been loaded in the core of Kola-2 NPP. The results obtained can be used for making the decision concerning the application of anti-debris filter for WWER-440 working fuel assemblies with the purpose of enhancing their debris-resistance properties. (authors)

  19. Report of a consultants meeting on backfittings and safety enhancement measures in NPPs with WWER 440/213 reactors. Extrabudgetary programme on the safety of WWER NPPS

    International Nuclear Information System (INIS)

    1994-01-01

    The purpose of this Consultants' Meeting held by the IAEA in Vienna from 11-15 April 1994 within the framework of the Extrabudgetary Programme on WWER Safety was to review and analyze safety issues revealed during operation and through analyses of NPPs with WWER 440/213 reactors. The initial list of safety issues based on the available reports from various studies had been prepared by the IAEA secretariat before the meeting, together with indications of safety enhancement measures proposed in various NPP units. During the meeting, the underlying safety concerns and actual technical status of the plants were discussed and the ranking of the safety issues was considered. 58 refs, 1 tab

  20. Anticipated transients without scram for WWER reactors. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1999-12-01

    Anticipated transients without scram (ATWS) are anticipated operational occurrences followed by the failure of one reactor scram function. Current international practice requires that the capability of pressurized water reactors (PWRs) to cope with ATWS be demonstrated following a systematic evaluation of plants' defence in depth. Countries operating PWRs require design consideration of ATWS events on a deterministic basis. The regulatory requirements may concern either specific mitigating systems or acceptable plant performance during these events. The prevailing international practice for performing transient analysis of ATWS for licensing is the best estimate approach. Available transient analyses of ATWS events indicate that WWER reactors, like PWRs, have the tendency to shut themselves down if the inherent nuclear feedback is sufficiently negative. Various control and limitation functions of the WWER plants also provide a degree of defence against ATWS. However, for most WWER plants, complete and systematic ATWS analyses have yet to be submitted for rigorous review by the regulatory authorities and preventive or mitigative measures have not been established. In addition, it has also been recognized that plant behaviour in case of ATWS also relies on certain system functions (use of pressurizer safety valves for liquid discharge, availability of steam dump valve to both the condenser (BRU-K) and the atmosphere (BRU-A) for secondary side pressure control, and others) which have been identified as safety issues and need to be qualified for accident conditions. In all countries operating WWERs, the need for ATWS investigations is recognized and reflected in the safety improvement programmes. ATWS analysis for WWERs is not required for the licensing process in Bulgaria, the Czech Republic (with the exception of the Temelin nuclear power plant) and Russia. Design consideration of ATWS is required if expert assessments of probabilistic safety assessment (PSA) results

  1. Guidelines for evaluation of anchorage adequacy for safety-related equipment typically used on WWER-type NPPs

    International Nuclear Information System (INIS)

    Masopust, R.

    1999-01-01

    This report describes the criteria which should be met when the capacity evaluation of anchorage of safety related equipment is performed for the WWER type NPPs. It should be noted that these criteria were developed specifically for anchorage of WWER type equipment and components to the concrete or steel building structures and they cover different types of anchor bolts and other anchorage details which are typical just for the existing, constructed or reconstructed WWER type NPPs. The screening approach for verifying of equipment anchorage presented in this report is based on a combination of inspections, calculations, and engineering judgement

  2. 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

  3. A simplified approach to WWER-440 fuel assembly head benchmark

    International Nuclear Information System (INIS)

    Muehlbauer, P.

    2010-01-01

    The WWER-440 fuel assembly head benchmark was simulated with FLUENT 12 code as a first step of validation of the code for nuclear reactor safety analyses. Results of the benchmark together with comparison of results provided by other participants and results of measurement will be presented in another paper by benchmark organisers. This presentation is therefore focused on our approach to this simulation as illustrated on the case 323-34, which represents a peripheral assembly with five neighbours. All steps of the simulation and some lessons learned are described. Geometry of the computational region supplied as STEP file by organizers of the benchmark was first separated into two parts (inlet part with spacer grid, and the rest of assembly head) in order to keep the size of the computational mesh manageable with regard to the hardware available (HP Z800 workstation with Intel Zeon four-core CPU 3.2 GHz, 32 GB of RAM) and then further modified at places where shape of the geometry would probably lead to highly distorted cells. Both parts of the geometry were connected via boundary profile file generated at cross section, where effect of grid spacers is still felt but the effect of out flow boundary condition used in the computations of the inlet part of geometry is negligible. Computation proceeded in several steps: start with basic mesh, standard k-ε model of turbulence with standard wall functions and first order upwind numerical schemes; after convergence (scaled residuals lower than 10-3) and near-wall meshes local adaptation when needed, realizable k-ε of turbulence was used with second order upwind numerical schemes for momentum and energy equations. During iterations, area-average temperature of thermocouples and area-averaged outlet temperature which are the main figures of merit of the benchmark were also monitored. In this 'blind' phase of the benchmark, effect of spacers was neglected. After results of measurements are available, standard validation

  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. Gamma-radiation effect on diamond and steel during their irradiation in WWER type reactors

    International Nuclear Information System (INIS)

    Nikolaenko, V.A.; Karpukhin, V.I.; Amaev, A.D.; Vikhrov, V.I.; Korolev, Yu.N.; Krasikov, E.A.

    1996-01-01

    A study is made into the influence of reactor gamma radiation on expansion of crystal lattice in diamond. The data obtained are compared to those on radiation embrittlement of reactor vessel steels. The necessity of taking into consideration gamma radiation effects on WWER reactor vessel radiation resistance during long-term operation is shown [ru

  6. Safety assessment of unit 5 (WWER-440/W-213) of the Greifswald nuclear power station

    International Nuclear Information System (INIS)

    1992-02-01

    The report represents the common results of the program of German-Soviet cooperation in reactor safety and radiation protection. The technical plant and features of type WWER-440/W-213 nuclear power plants, basic legal licensing principles, reactor core and pressurized components, load resulting from accidents, systems engineering, spreading impacts, civil engineering aspects, and the evaluation of operating experience are described. (DG)

  7. Method for determining the outlet temperature of fuel assemblies unsupplied with thermometer in WWER-440 reactors

    International Nuclear Information System (INIS)

    Miko, S.; Kalya, Z.; Hamvas, I.

    1987-09-01

    The paper outlines a method for the evaluation of the outlet temperatures of fuel assemblies unsupplied with thermometer in WWER-440 reactors. The process is based on interpolation of directly measured assembly temperatures. A quantitative comparison of the errors of described algorithm to those of standard plant-computer interpolation rutine is also presented. (author)

  8. The interaction between bitumen matrix and chemical components of radioactive wastes of WWER type

    International Nuclear Information System (INIS)

    Selucky, P.; Sazavsky, P.; Peka, V.; Krupka, M.

    2000-01-01

    The interaction between bitumen matrix and chemical components of WWER type radioactive wastes was studied. So called ''cold'' model bitumen products were prepared and compared with real products using macroDTA method. On the basis of obtained curves, the evaluation of bitumen product fire risks was performed with the aim to minimize risks of bituminization process. (authors)

  9. Influence of temperature measurement accuracy and reliability on WWER-440 reactor operation

    International Nuclear Information System (INIS)

    Petenyi, V.; Ricany, J.

    2001-01-01

    The WWER-440 reactor power is controlled by coolant heat-up measurements installed on hot and cold circulation loops (enthalpy rise). For power distribution determination the thermocouples installed in reactor vessel above the fuel assemblies are mainly utilised. The paper shortly presents some interesting observations of temperature measurements influencing the reactor power operation of revealed changes in reactor core behaviour. (Authors)

  10. Results of the safety analyses for the Greifswald and Stendal WWER nuclear power plants

    International Nuclear Information System (INIS)

    Milhem, J.L.

    1993-03-01

    Following a brief introduction of the design features of the three types of the WWER reactors, the paper deals with the main issues of the safety-related design and the most important recommendations which have been derived for upgrading measures. Furthermore some operational safety aspects of the VVER-1000 will be discussed in some detail

  11. 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

  12. 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

  13. Spent fuel storage practices and perspectives for WWER fuel in Eastern Europe

    International Nuclear Information System (INIS)

    Takats, F.

    1999-01-01

    In this lecture the general issues and options in spent fuel management and storage are reviewed. Quantities of spent fuel world-wide and spent fuel amounts in storage as well as spent fuel capacities are presented. Selected examples of typical spent fuel storage facilities are discussed. The storage technologies applied for WWER fuel is presented. Description of other relevant storage technologies is included

  14. 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

  15. Solution and implementation of project ''Simulator of WWER-440 nuclear power plant''

    International Nuclear Information System (INIS)

    Koukal, V.

    1985-01-01

    The time data are given of the development and construction of the simulator of a WWER-440 nuclear power plant unit. The individual tasks are summed up which are related to the project implementation, and cooperating institutions and enterprises are listed. (J.C.)

  16. Part-task simulator for a WWER-440 nuclear power plant subsystem

    International Nuclear Information System (INIS)

    Szabo, B.K.

    1988-07-01

    PC-based part-task simulators for simulating subsystems of nuclear power plants are low cost tools for operator training. In the Central Research Institute for Physics, Budapest, a simulator system has been developed to facilitate fast development of such simulators. The first application simulates the Neutron Flux Monitoring System of WWER-440 nuclear power plants. (author) 9 refs.; 2 figs

  17. Developing new calculational and experimental techniques for studying the WWER reactor cores and characteristics

    International Nuclear Information System (INIS)

    Zakharko, Yu.A.; Proshkin, A.A.

    1986-01-01

    Necessity of analytical approaches alongside with existing numerical methods of fuel element calculation is discussed. Analytical solutions of viscoelastic equations describing mechanical fuel-cladding interaction have been obtained. At that universal temperature dependence of creep characteristics is suggested. Dependence of behaviour of the WWER fuel element fuel and cladding on absolute temperature level and gradients is analysed

  18. Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

    International Nuclear Information System (INIS)

    Ilieva, K.; Belousov, S.; Apostolov, T.

    1998-01-01

    The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

  19. 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)

  20. Experience of developments and implementation of advanced fuel cycles of WWER-440 reactors

    International Nuclear Information System (INIS)

    Gagarinski, A.A.; Lizorkin, M.P.; Novikov, A.N.; Proselkov, V.N.; Saprykin, V.V.

    2000-01-01

    The paper presents the experience of development and implementation of advanced four- and five-year fuel cycles in the WWER-440 reactors, the results of experimental operation of the new fuel design and the main neutronic characteristics of the core. (Authors)

  1. Lifetime forecasting of a WWER NPP steam generator tube bundle from stress corrosion conditions

    International Nuclear Information System (INIS)

    Sereda, E.V.; Gorbatykh, V.P.

    1984-01-01

    An approach is outlined to the description of corrosion cracking of austenitic stainless steels in hot chloride solutions to predict the failure of WWER NPP steam generator heat exchange tubes. The dependence of the corrosion cracking development rate on the chloride concentration and characteristic electrochemical potentials is suggsted. The approach permits also to determine the quantity of damaged tubes versus the operation parameters

  2. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  3. 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)

  4. 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

  5. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  6. Experimental investigation of power peak in vicinity of WWER-440 control rod at end of fuel cycle

    International Nuclear Information System (INIS)

    Mikus, J.

    2004-01-01

    This paper presents some results of the axial power (fission density) distribution measurements carried out on the light-water, zero-power reactor LR-0 in a WWER-440 type core in vicinity of the WWER-440 control rod model at zero boron concentration in moderator, modelling the conditions at the end of the WWER-440 fuel cycle. Further information concerns the control rod model description, specification of the LR-0 core, fuel assemblies and measurement conditions. The aim of performed experiment is enlargement of the available power peaking database to enable the validation of the calculation codes by means of the measured data that correspond to the end of WWER-440 fuel cycle. (author)

  7. Simulation of the dynamic behaviour of the secondary circuit of a WWER-440 type nuclear power plant Pt. 2

    International Nuclear Information System (INIS)

    Doorenbos, J.; Gacs, A.; Kiss, Zs.

    1987-12-01

    This report describes the dynamic simulation models of the most important controllers of the secondary circuit of a WWER-440 type nuclear power plant, i.e., the hydraulic turbine controller and the level controls of the condenser hotwell and that of the feedwater tank. Simulation results are also presented. (For dynamic simulation models of the primary circuit of WWER-440 type reactors see Reports KFKI--1983-127 and KFKI--1985-08.) (author) 15 figs

  8. 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.)

  9. 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.

  10. Comparison of fuel cycles characteristics for nuclear energy systems based on WWER-TOI and BN-1200 reactors

    International Nuclear Information System (INIS)

    Kagramanyan, V.S.; Kalashnikov, A.G.; Kapranova, Eh.N.; Puzakov, A.Yu.

    2014-01-01

    Authors determine the characteristics of the fuel cycle (FC) based on stationary nuclear power system based on WWER-TOI and BN-1200 reactors with fuel of different composition. Characteristics of reactor systems with partial or complete spent nuclear fuel reprocessing and recycling of plutonium are compared to those of the reference system consisting only of WWER-TOI with uranium oxide fuel, operating in an open FC [ru

  11. Simulation of the dynamic behaviour of the secondary circuit of a WWER-440 type nuclear power plant Pt. 1

    International Nuclear Information System (INIS)

    Gacs, A.; Janosy, J.S.; Kiss, Zs.

    1987-07-01

    This report describes the simulation model of the secondary circuit of a WWER-440 type nuclear power plant. The goal of this modelling is to simulate normal and small abnormal transients in a Basic Principles Simulator. The earlier reports describing the dynamic simulation of primary circuit of a WWER-440 nuclear power plant are KFKI--1983-127 and KFKI--1985-08. At present the controllers of the secondary circuit are not simulated. Finally, some simulation results are presented. (author)

  12. Western and WWER materials investigations - past lessons, present achievements and future trends for fuel rod cladding and fuel assembly structure

    International Nuclear Information System (INIS)

    Weidinger, H.

    2001-01-01

    The paper gives a detailed overview of Western and WWER materials used in nuclear fuel manufacturing industry. The status of technical experience with regard to design, fabrication and particular in-pile behavior is described and compared for material of major importance for PWR and WWER fuel. In particular Zr-base alloys for cladding tubes, spacer grids and guide thimbles are considered. In addition spacer spring materials are also discussed. The paper shows that during the last decade a considerable diversification of different Zr materials occurred in Western PWR fuel, while for WWER fuel the focus is still on the classical Zr1%Nb material. Corrosion and hydrogen uptake as well as the dimensional behavior (creep and growth) of all presently relevant Zr-based materials is described in detail. For spacer springs Zr-based and Ni-based materials are considered. For this application spring force relaxation is the most important issue. The paper shows that the focus of consideration is typically different for PWR and WWER fuel materials. While for PWR fuel mainly corrosion and hydrogen uptake is most important and design limiting, for WWER fuel the focus of interests is with mechanical strength. The main reason for this significant difference is that the corrosive environment is typically different for PWR and WWER cores

  13. 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)

  14. 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.

  15. Analysis of differences in fuel safety criteria for WWER and western PWR nuclear power plants

    International Nuclear Information System (INIS)

    2003-11-01

    In 2001 the OECD issued a report of the NEA/CSNI (Committee on the Safety of Nuclear Installations) Task Force on the existing safety criteria for reactor fuel for western LWR nuclear power plants (both for PWRs and BWRs) under new design elements. Likewise in 2001, the IAEA released a report by a Working Group on the existing safety criteria for reactor fuel for WWER nuclear power plants under new design requirements. However, it was found that it was not possible to compare the two sets of criteria on the basis upon which they had been established. Therefore, the IAEA initiated an assessment of the common features and differences in fuel safety criteria between plants of eastern and western design, focusing on western PWRs and eastern WWER reactors. Between October 2000 and November 2001, the IAEA organized several workshops with representatives from eastern and western European countries in which the current fuel safety related criteria for PWR and WWER reactors were reviewed and compared. The workshops brought together expert representatives from the Russian Federation, from the Ukraine and from western countries that operate PWRs. The first workshop focused on a general overview of the fuel safety criteria in order for all representatives to appreciate the various criteria and their respective bases. The second workshop (which involved one western and one eastern expert) concentrated on addressing and explaining the differences observed, and documenting all these results in preparation for a panel discussion. This panel discussion took place during the third workshop, where the previously obtained results were reviewed in detail and final recommendations were made. This report documents the findings of the workshops. It highlights the common features and differences between PWR and WWER fuel, and may serve as a general basis for the safety evaluation of these fuels. Therefore, it will be very beneficial for licensing activities for PWR and WWER plants, as it

  16. 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

  17. 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

  18. 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.

  19. WWER-440/230 reactor pressure vessel integrity. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-08-01

    This report was prepared with the objective of integrating all aspects involved and to provide plant specific information on the issue of reactor pressure vessel integrity including pressurized thermal shock assessment. Areas of the thermal hydraulic analysis including selection of transients, of the structural analysis including fracture mechanics assessment and of the material properties including embrittlement, annealing and re-embrittlement behaviour are addressed. The report also provides related recommendations and conclusions as well as detailed information on the plant specific status for operating WWER-440/230 nuclear power plants. 10 refs, 9 figs, 9 tabs

  20. 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)

  1. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    International Nuclear Information System (INIS)

    Solyany, V.I.; Bibilashvili, Yu.K.; Sukhanov, G.I.; Pimenov, Yu.V.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-01-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness. (author)

  2. 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.

  3. 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

  4. Fracture toughness of WWER Uranium dioxide fuel pellets with various grain size

    International Nuclear Information System (INIS)

    Sivov, R.; Novikov, V.; Mikheev, E.; Fedotov, A.

    2015-01-01

    Uranium dioxide fuel pellets with grain sizes 13, 26, and 33 μm for WWER were investigated in the present work in order to determine crack formation and the fracture toughness.The investigation of crack formation in uranium oxide fuel pellets of the WWER-types showed that Young’s modulus and the microhardness of polycrystalline samples increase with increasing grain size, while the fracture toughness decreases. Characteristically, radial Palmqvist cracks form on the surface of uranium dioxide pellets for loads up to 1 kg. Transgranular propagation of cracks over distances several-fold larger than the length of the imprint diagonal is observed in pellets with large grains and small intragrain pores. Intergranular propagation of cracks along grain boundaries with branching occurs in pellets with small grains and low pore concentration on the grain boundaries. Blunting on large pores and at breaks in direction does not permit the cracks to reach a significant length

  5. Applied model of through-wall crack of coolant vessels of WWER-type reactors

    International Nuclear Information System (INIS)

    Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

    2010-01-01

    We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

  6. Calculation models of pressure wave propagation within the WWER-440 primary circulating loop

    International Nuclear Information System (INIS)

    Adamik, V.; Tkach, A.

    1982-01-01

    Computer codes SHOCK, LOVE, BAREL are described that can be used for the study of pressure wave propagation within the reactor and pipeline system during a LOCA as well as for mechanical loads identification in various parts of the system. SHOCK code is applicable to one-dimensional pressure wave propagation analysis in any hydraulic network containing a compressible nonviscous liquid with a constant (within the considered transient process period) density. LOVE code allows to calculate non-symmetrical mechanical loads on the WWER shaft in case of the main circulation pipeline cold branch rupture. BAREL code is an advanced modification of SHOCK code. It is fitted for two-dimensional pressure wave propagation analysing in the downstream section of a pressurised water reactor in case of the main circulation pipeline cold branch rupture. The calculation results for B-213 type WWER-440 reactor are presented that have been obtained under the assumption of perfect structure rigidity [ru

  7. Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

    Energy Technology Data Exchange (ETDEWEB)

    Debarberis, L. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: luigi.debarberis@cec.eu.int; Acosta, B. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands)]. E-mail: beatriz.acosta-iborra@jrc.nl; Zeman, A. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Sevini, F. [Joint Research Centre of the European Commission, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Ballesteros, A. [Tecnatom, Avd. Montes de Oca 1, San Sebasitan de los Reyes, E-28709 Madrid (Spain); Kryukov, A. [Russian Research Centre Kurchatov Institute, Kurchatov Square 1, 123182 Moscow (Russian Federation); Gillemot, F. [AEKI Atomic Research Institute, Konkoly Thege M. ut 29-33, 1121 Budapest (Hungary); Brumovsky, M. [NRI, Nuclear Research Institute, Husinec-Rez 130, 25068 Rez (Czech Republic)

    2006-04-15

    It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.

  8. 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.

  9. Twenty years of operation of WWER 440/230 units in Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Tomek, J.

    1998-01-01

    It is twenty years this year since the first unit WWER 440 of Slovak Nuclear Power Plant Jaslovske Bohunice was commissioned. There are four units WWER 440 in operation Jaslovske Bohunice site. First two units of older soviet PWR design V-230 (also known as V-1) and other two units of newer V-213 type (also known as V-2). The goal of this presentation is to summarize and evaluate the operation of Unit 1 and 2 for this period of time and mainly to describe what has been done and what is planned to be done to increase the nuclear safety and operational reliability of both units. The operating organization and regulatory authority assume that an internationally acceptable level of safety will be reached by accomplishing of the upgrading program.(author)

  10. Comment on the in-core measurement in the WWER nuclear power plant

    International Nuclear Information System (INIS)

    Krett, V.; Dach, K.; Erben, O.

    1985-01-01

    The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

  11. WWER expert system for fuel failure analysis using the RTOP-CA code

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.

    2008-01-01

    The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)

  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. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Solyany, V I; Bibilashvili, Yu K; Sukhanov, G I; Pimenov, Yu V [Vsesoyuznyj Nauchno-Issledovatel' skij Inst. Neorganicheskikh Materialov, Moscow (USSR); Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-12-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness.

  14. The use of the codes from MCU family for calculations of WWER type reactors

    International Nuclear Information System (INIS)

    Abagijan, L.P.; Alexeyev, N.I.; Bryzgalov, V.I.; Gomin, E.A.; Glushkov, A.E.; Gorodkov, S.S.; Gurevich, M.I.; Kalugin, M.A.; Marin, S.V.; Shkarovsky, D.A.; Yudkevich, M.S.

    2000-01-01

    The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)

  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. Seismic analysis of the safety related piping and PCLS of the WWER-440 NPP

    International Nuclear Information System (INIS)

    Berkovski, A.M.; Kostarev, V.V.; Schukin, A.J.; Boiadjiev, Z.; Kostov, M.

    2001-01-01

    This paper presents the results of seismic analysis of Safety Related Piping Systems of the typical WWER-440 NPP. The methodology of this analysis is based on WANO Terms of Reference and ASME BPVC. The different possibilities for seismic upgrading of Primary Coolant Loop System (PCLS) were considered. The first one is increasing of hydraulic snubber units and the second way is installation of limited number of High Viscous Dampers (HVD). (author)

  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)

    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.

  18. Key developments in the advanced NPP with WWER-640/V-407 reactor plant design

    International Nuclear Information System (INIS)

    Dragunov, Yu.G.; Mokhov, V.A.; Nikitenko, M.P.; Afrov, A.M.

    1999-01-01

    The report covers the main design features of advanced NPP equipped with WWER-640 reactor, that take into account the up-to-date approaches in the process of forming safety concepts. An approach to accident management has been analysed, beyond design-basis accidents included. A description of principal safety systems has been presented as well as the interrelation of their operation. The principal features of the systems design have been shown. (author)

  19. Calculation study of the WWER-440 fuel performance for extended burnup

    International Nuclear Information System (INIS)

    Kujal, J.; Pazdera, F.; Barta, O.

    1984-01-01

    The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

  20. Source convergence problems in the application of burnup credit for WWER-440 fuel

    International Nuclear Information System (INIS)

    Hordosy, Gabor

    2003-01-01

    The problems in Monte Carlo criticality calculations caused by the slow convergence of the fission source are examined on an example. A spent fuel storage cask designed for WWER-440 fuel used a sample case. The influence of the main parameters of the calculations is investigated including the initial fission source. A possible strategy is proposed to overcome the difficulties associated by the slow source convergence. The advantage of the proposed strategy that it can be implemented using the standard MCNP features. (author)

  1. 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.

  2. 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

  3. Scientific and technological basis for maintenance optimization, planning, testing and monitoring for NPP with WWER

    International Nuclear Information System (INIS)

    Kovrizhkin, Yu.L.; Skalozubov, V.I.; Kochneva, V.Yu.

    2009-01-01

    The main results of the developments in the sphere of NPPs with WWER production efficiency increasing by the way of the maintenance optimization planning, testing and monitoring of the equipment and systems are shown. The attention is paid to the metal control during maintenance period of Power Unit. The realization methods of the transition concept at the repair according to the technical condition are resulted

  4. Computer techniques for experimental work in GDR nuclear power plants with WWER

    International Nuclear Information System (INIS)

    Stemmler, G.

    1985-01-01

    Nuclear power plant units with WWER are being increasingly equipped with high-performance, programmable process control computers. There are, however, essential reasons for further advancing the development of computer-aided measuring systems, in particular for experimental work. A special structure of such systems, which is based on the division into relatively rigid data registration and primary handling and into further processing by advanced programming language, has proved useful in the GDR. (author)

  5. Analyses and results from standard surveillance programmes of WWER 440/V-213C reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Falcnik, M; Brumovsky, M; Pav, T [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    In Czech and Slovak republics, six units of WWER 440/C type reactors are monitored by surveillance specimens programmes; the specimens are determined for static tensile testing, impact notch toughness testing and fracture toughness evaluation. Results of mechanical properties of these specimens after irradiation in intervals between 1 and 5 years of operation, are summarized and discussed with respect to the effect of individual heats and welded joints, radiation embrittlement, and annealing recovery. (authors). 3 refs., 11 figs., 2 tabs.

  6. Impact of changed fuel performances on safety barrier effectiveness at normal operation of NPP with WWER

    International Nuclear Information System (INIS)

    Zhurbenko, A. V.; Semchenkov, Y. M.; Slavyagin, P.D.

    2007-01-01

    The paper presents the analysis of adopted safety barriers against propagation of fission product released from WWER core of active power plants. Relationship between system and equipment performances and safety barriers is demonstrated. The fundamental principles of methodological approach to the operational limit determination based on the assessment of iodine-131 specific activity in the primary circuit are discussed. Problems of substantiating the operational limit for primary coolant activity are analyzed for conditions of growing burnup (Authors)

  7. Automatic vibration monitoring system for the diagnostic inspection of the WWER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Hollo, E.; Siklossy, P.; Toth, Zs.

    1982-01-01

    In the Hungarian Research Institute for Electric Power Industry (VEIKI) an automatic vibration monitoring system for diagnostics and inspection of nuclear power plants of type WWER-440 was developed. The paper summarizes the results of this work and investigates the use of mechanical vibrations and oscillations induced by flow for fault diagnosis. The design of the hardware system, the present software possibilities, the laboratory experiments and the guidelines for future software developments are also described in detail. (A.L.)

  8. 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)

  9. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    International Nuclear Information System (INIS)

    Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

    2007-01-01

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  10. Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

    International Nuclear Information System (INIS)

    Danilov, V.; Dobrov, V.; Semishkin, V.; Vasilchenko, I.

    2006-01-01

    The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

  11. WWER-440 control assembly local power peaking investigation on LR-0 reactor

    International Nuclear Information System (INIS)

    Mikus, J.

    2002-01-01

    This paper presents information concerning the local power peaking problem induced by the WWER-440 control assembly and the investigation possibilities on the light water, zero power reactor LR-0 at the Nuclear Research Institute (NRI) Rez plc. A brief description is given about the disposable control assembly model, experimental arrangement and conditions on the LR-0 reactor with regard to the earlier performed investigations as well as to the relevant measurements to be realized in the near future.(abstract)

  12. On the heat exchange tube failures in steam generators at NPPs with WWER reactors

    International Nuclear Information System (INIS)

    Titov, V.F.; Banyuk, G.F.; Brykov, S.I.

    1992-01-01

    Data on dynamics of failed heat exchanging tube closing in steam generators of NPPs with WWER type reactors for the whole period of their operation are presented. It is shown that the main cause of the tube failures consists in their corrosion cracking under stresses. The effect of chlorine ions on tubes is intensified by the presence of porous sediments on heat exchaning surfaces in quantities exceeding 150 g/m 2

  13. 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

  14. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  15. 11. International conference on WWER fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    Manolova, M.; Boneva, S.; Mitev, M.

    2015-01-01

    This publication is a compilation of the papers presented in 11th International Conference on WWER Fuel Performance, Modeling and Experimental Support, organized by the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Sciences in co-operation with the International Atomic Energy Agency (IAEA), Vienna, Austria, supported by the Kozloduy Nuclear Power Plant (KNPP), the Bulgarian Nuclear Regulatory Agency, and TVEL Fuel Company, Russia. The Conference took place in hotel Bolero, Golden Sands Resort, Bulgaria, from 26 September 2015 to 3 October 2015. It was attended by 117 participants, among them more than 100 experts and specialists from 22 countries, including representatives of 3 international organizations, 16 Russian organizations and other 36 foreign institutes, nuclear fuel plants, nuclear power plants and organizations responsible for WWER and PWR fuel design, manufacturing and research, and 3 Bulgarian organizations, working for the Bulgarian nuclear industry. 70 papers have been presented in the Conference in 6 oral and 1 poster session, covering: (1) general overview lectures; (2) fuel performance and operational experience; (3) fuel modeling and experimental support; (4) fuel safety and QA; (5) spent fuel performance and management; (6) specific issues of WWER-1000 fuel reliability. The proceedings provide Summary, Conclusions and Recommendations of the Conference, together with the full text of the presentations. IAEA Technical Meeting (TM) “Achieving zero fuel failure rates: challenges and perspectives”, 1 – 2 October 2015 was organized in conjunction with the 11th International Conference on WWER Fuel Performance, Modelling and Experimental Support. The reports presented on TM sessions are included in the Conference Proceedings too

  16. 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

  17. 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

  18. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Haralampieva, Z [National Electric Company Ltd., Kozloduy (Bulgaria); Scheglov, A; Proselkov, V [Institute of Nuclear Reactors, RSC Kurchatov Inst., Moscow (Russian Federation)

    1997-08-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab.

  19. Ranking of safety issues for WWER-440 model 230 nuclear power plants

    International Nuclear Information System (INIS)

    1992-02-01

    In response to requests from Member States operating Soviet designed WWER-440/230 nuclear power plants (NPPs) for assistance through the IAEA's nuclear safety services, a major international project was established to evaluate these first generation reactors as a complement to relevant ongoing national, bilateral and multilateral activities. The objective is to assist countries operating WWER-440/230 NPPs in performing comprehensive safety reviews aimed at identifying design and operational weaknesses. The scope of the project includes a review of the conceptual design of WWER-440/230 NPPs, safety review missions to each one of the operating reactors to review design and operational aspects and studies to resolve issues of generic safety concern. This report was prepared by a group of international experts and the IAEA staff and discussed by the Project Steering Committee, December 9-13, 1991 in Vienna. An overview of the safety issues identified is presented indicating their effect on the performance of the basic safety functions. Conceptual recommendations related to design issues are given as a technical basis for the safety modifications required

  20. The Westinghouse approach - an I and C modernization program for WWERs

    International Nuclear Information System (INIS)

    Werner, C.L.; Wassel, W.W.; Novak, V.

    1993-01-01

    When entering into a design program that is a marriage between two designs it is very difficult to separate self imposed design criteria from the requirements of the program. Therefore, the criteria of and the requirements for the Westinghouse modernization program will be discussed as one. These are outlined below: 1) The OSART Mission that was conducted by the IAEA at the Temelin Plant in 1990 identified the need to provide a new comprehensive Safety Analysis to verify the various aspects of the WWER safety system design. This recommendation is one that Westinghouse will provide as part of the WWER I and C Modernization Program. The design, no matter how well proven or verified from a hardware design point of view, is only as good as the basis for the system design; 2) Minimize the impact on the civil design aspects of the plant where possible and where this requirements do not affect the safety features of the design; 3) Ensure compatibility of the design to meet the latest US NRC requirements and those of the implementing country, applicable to the systems functional and hardware designs. This is a Westinghouse standard corporate requirement for all nuclear plant and systems design whether they be foreign or domestic; 4) Provide the most modern, proven design for the I and C systems. Application of the Westinghouse Instrumentation and Control microprocessor based design to the WWER Modernization Program will provide the basis for upgrading plants to meet western standards. (author) 6 figs., 1 ref

  1. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    International Nuclear Information System (INIS)

    Stefanova, S.; Vitkova, M.; Simeonova, V.; Passage, G.; Manolova, M.; Haralampieva, Z.; Scheglov, A.; Proselkov, V.

    1997-01-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab

  2. Evaluated experimental database on critical heat flux in WWER FA models

    International Nuclear Information System (INIS)

    Artamonov, S.; Sergeev, V.; Volkov, S.

    2015-01-01

    The paper presents the description of the evaluated experimental database on critical heat flux in WWER FA models of new designs. This database was developed on the basis of the experimental data obtained in the years of 2009-2012. In the course of its development, the database was reviewed in terms of completeness of the information about the experiments and its compliance with the requirements of Rostekhnadzor regulatory documents. The description of the experimental FA model characteristics and experimental conditions was specified. Besides, the experimental data were statistically processed with the aim to reject incorrect ones and the sets of experimental data on critical heat fluxes (CHF) were compared for different FA models. As a result, for the fi rst time, the evaluated database on CHF in FA models of new designs was developed, that was complemented with analysis functions, and its main purpose is to be used in the process of development, verification and upgrading of calculation techniques. The developed database incorporates the data of 4183 experimental conditions obtained in 53 WWER FA models of various designs. Keywords: WWER reactor, fuel assembly, CHF, evaluated experimental data, database, statistical analysis. (author)

  3. Temperature and velocity field of coolant at inlet to WWER-440 core - evaluation of experimental data

    International Nuclear Information System (INIS)

    Jirous, F.; Klik, F.; Janeba, B.; Daliba, J.; Delis, J.

    1989-01-01

    Experimentally determined were coolant temperature and velocity fields at the inlet of the WWER-440 reactor core. The accuracy estimate is presented of temperature measurements and the relation is given for determining the resulting measurement error. An estimate is also made of the accuracy of solution of the system of equations for determining coefficients B kn using the method of the least square fit. Coefficients B kn represent the relative contribution of the mass flow of the k-th fuel assembly from the n-th loop and allow the calculation of coolant temperatures at the inlet of the k-th fuel assembly, when coolant temperatures in loops at reactor inlet are known. A comparison is made of the results of measurements on a hydrodynamic model of a WWER-440 reactor with results of measurements made at unit 4 of the Dukovany nuclear power plant. Full agreement was found for 32 model measurements and 6 reactor measurements. It may be assumed that the results of other model measurements obtained for other operating variants will also apply for an actual reactor. Their applicability may, however, only be confirmed by repeating the experiment on other WWER-440 reactors. (Z.M.). 5 figs., 7 refs

  4. 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

  5. Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

    International Nuclear Information System (INIS)

    El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

    2004-01-01

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

  6. Selected safety aspects of containments for nuclear power plants with WWER-440 reactors

    International Nuclear Information System (INIS)

    Jankowski, M.W.; Kulig, M.J.; Strupczewski, A.; Balabanov, E.D.

    1996-01-01

    Considerable attention has been and continues to be focused on the design and operational features that prevent the release of radioactive materials to the environment for a spectrum of accidents for the two classes of WWER-440 reactors: the older 230 model and the more recently designed 213 models. This paper, based on published and unpublished information, aims to clarify the perceptions of the Russian WWER-440 models 230 and 213 nuclear power plant containment system designs and their relevance to selected aspects of accident mitigation. It should be noted that these are unclearly and often negatively perceived, primarily because of a lack of reliable information and a poorly assembled experimental database. Conflicting statements have been made regarding the nature and the features of the plant's containment system. The paper presents a brief outline of the design of both WWER-440 models with respect to their confinement functions. Selected safety-related aspects of the accident localization systems are discussed, and the recognized shortcomings and safety merits are pointed out. The older 230 units experience high leak rates and are designed to withstand medium-size pipe breaks. The possible implications for safety are pointed out in the paper. The on going studies that concentrate on improving the system are highlighted. (orig.)

  7. 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.

  8. 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.

  9. Strength analyses of the bubbler condenser structure of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia). Report of the IAEA Technical Co-operation Project RER/9/004 on evaluation of safety aspects of WWER-440 model 213 nuclear power plants

    International Nuclear Information System (INIS)

    1995-06-01

    This publication addresses the topic of mechanical strength of the bubbler condenser applied in the WWER-440 model 213 plants and is intended to assist WWER-440/213 operators in the re-assessment of the bubbler condenser performance. It is hoped that it will also be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, figs and tabs

  10. Strength analyses of the bubbler condenser structure of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia). Report of the IAEA Technical Co-operation Project RER/9/004 on evaluation of safety aspects of WWER-440 model 213 nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    This publication addresses the topic of mechanical strength of the bubbler condenser applied in the WWER-440 model 213 plants and is intended to assist WWER-440/213 operators in the re-assessment of the bubbler condenser performance. It is hoped that it will also be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, figs and tabs.

  11. 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

  12. Optimization of design solutions on safety and economy for power unit of NPP with WWER reactor of new generation. Annex 8

    International Nuclear Information System (INIS)

    Krushelnitsky, V.N.; Berkovich, V.M.; Shvyrayev, Yu.; Podshebaykin, A.K.; Fil, N.S.

    2002-01-01

    Development of new generation WWER reactors is being carried out in Russia. These new projects with WWER reactors aim to achieve increased levels of safety and reduced costs. This paper describes these designs and discusses the main factors leading to the safety level increase and the improved economics. (author)

  13. 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

  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. 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

  16. The decommissioning of WWER type nuclear power plants. Final report on an IAEA regional technical co-operation project

    International Nuclear Information System (INIS)

    2000-01-01

    Numerous WWER-440 nuclear power plants are in operation in central and eastern Europe and a small number have already been shut down. In addition to reactors already shut down, many other reactors will reach the end of their design lifetime in a few years and become candidates for decommissioning. It is unfortunate that little consideration was devoted to decommissioning of WWER-440 reactors at the plant design and construction stage, and little emphasis was placed on planning for decommissioning. It is within this context that the IAEA launched a regional technical co-operation project in 1994 with the aim of providing guidance on planning and management of decommissioning for WWERs. The project, which had a duration of four years (1995-1998), included the organization of workshops and scientific visits to countries having WWERs and other countries where active decommissioning projects were under way. Eventually, participants suggested the consolidation of expert guidance and collective opinions into a TECDOC, which was drafted by both designated participants from project recipient countries and invited experts. The TECDOC has the aim of serving as a stimulus for all concerned parties in central and eastern European countries to initiate concrete decommissioning planning, including assessment of existing and required resources for the eventual implementation of decommissioning plans. In addition, the regional technical co-operation project has managed to bring together in this TECDOC a number of good practices that could be useful in WWER-440 decommissioning

  17. Procedures for analysis of accidents in shutdown modes 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)

    1997-07-01

    Operational events occurring during shutdown conditions contribute significantly to the NPP risk due to the fact that both preventive and mitigatory capabilities of the plant are somehow degraded. The need for detailed information in the performance and review of accident analysis for WWER type NPPs was identified as a priority within IAEA Extrabudgetary Program on Safety of WWER and RBMK NPPs. The present guidelines were developed through two consultants meetings in 1995 and 1996. The guidelines establish a set of criteria for performing deterministic analysis of accidents, initiated by events occurring under shutdown conditions. This report is mostly relevant for licensing type calculations, and may to a certain extent, also used for development, improvement or justification of the plant limits and conditions, emergency operating procedures, operator training programs and probabilistic safety studies. The guidelines apply to all WWER plants in operation and/or under construction

  18. 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

  19. Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

    International Nuclear Information System (INIS)

    Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

    2014-01-01

    The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

  20. 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.

  1. 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)

  2. 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

  3. 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)

  4. 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

  5. OPAL- the in-core fuel management code system for WWER reactors

    International Nuclear Information System (INIS)

    Krysl, V.; Mikolas, P.; Sustek, J.; Svarny, J.; Vlachovsky, K.

    2002-01-01

    Fuel management optimization is a complex problem namely for WWER reactors, which at present are utilizing burnable poisons (BP) to great extent. In this paper, first the concept and methodologies of a fuel management system for WWER 440 (NPP Dukovany) and NPP WWER 1000 (NPP Temelin) under development in Skoda JS a.s. are described and followed by some practical applications. The objective of this advanced system is to minimize fuel cost by preserving all safety constraints and margins. Future enhancements of the system will allow is it to perform fuel management optimization in the multi-cycle mode. The general objective functions of the system are the maximization of EOC reactivity, the maximization of discharge burnup, the minimization of fresh fuel inventory / or the minimization of feed enrichment, the minimization of the BP inventory. There are also safety related constraints, in which the minimization of power peaking plays a dominant role. The core part of the system requires meeting the major objective: maximizing the EOC Keff for a given fuel cycle length and consists of four coupled calculation steps. The first is the calculation of a Loading Priority Scheme (LPS). which is used to rank the core positions in terms of assembly Kinf values. In the second step the Haling power distribution is calculated and by using fuel shuffle and/or enrichment splitting algorithms and heuristic rules the core pattern is modified to meet core constraints. In this second step a directive/evolutionary algorithm with expert rules based optimization code is used. The optimal BP assignment is alternatively considered to be a separate third step of the procedure. In the fourth step the core is depleted in normal up to 3D pin wise level using the BP distribution developed in step three and meeting all constraints is checked. One of the options of this optimization system is expert friendly interactive mode (Authors)

  6. Modeling of corrosion product migration in the secondary circuit of nuclear power plants with WWER-1200

    Science.gov (United States)

    Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.; Motkova, E. A.; Zelenina, E. V.; Prokhorov, N. A.; Gorbatenko, S. P.; Tsitser, A. A.

    2016-04-01

    Models of corrosion and mass transfer of corrosion products in the pipes of the condensate-feeding and steam paths of the secondary circuit of NPPs with WWER-1200 are presented. The mass transfer and distribution of corrosion products over the currents of the working medium of the secondary circuit were calculated using the physicochemical model of mass transfer of corrosion products in which the secondary circuit is regarded as a cyclic system consisting of a number of interrelated elements. The circuit was divided into calculated regions in which the change in the parameters (flow rate, temperature, and pressure) was traced and the rates of corrosion and corrosion products entrainment, high-temperature pH, and iron concentration were calculated. The models were verified according to the results of chemical analyses at Kalinin NPP and iron corrosion product concentrations in the feed water at different NPPs depending on pH at 25°C (pH25) for service times τ ≥ 5000 h. The calculated pH values at a coolant temperature t (pH t ) in the secondary circuit of NPPs with WWER-1200 were presented. The calculation of the distribution of pH t and ethanolamine and ammonia concentrations over the condensate feed (CFC) and steam circuits is given. The models are designed for developing the calculation codes. The project solutions of ATOMPROEKT satisfy the safety and reliability requirements for power plants with WWER-1200. The calculated corrosion and corrosion product mass transfer parameters showed that the model allows the designer to choose between the increase of the correcting reagent concentration, the use of steel with higher chromium contents, and intermittent washing of the steam generator from sediments as the best solution for definite regions of the circuit.

  7. Experience from operation of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia). Report of the IAEA technical co-operation project RER/9/004 on evaluation of safety aspects of WWER-440 model 213 nuclear power plants

    International Nuclear Information System (INIS)

    1995-07-01

    This TECDOC provides a comprehensive review of the operational experience with WWER-440/213 plants. It is hoped that it will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, figs and tabs

  8. Experience from operation of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia). Report of the IAEA technical co-operation project RER/9/004 on evaluation of safety aspects of WWER-440 model 213 nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This TECDOC provides a comprehensive review of the operational experience with WWER-440/213 plants. It is hoped that it will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, figs and tabs.

  9. Experimental design verification of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia). Report of the IAEA technical co-operation project RER/9/004 on evaluation of safety aspects of WWER-440 model 213 nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This publication addresses the experimental research supporting the design of WWER-440 model 213 plants. it is hoped that the material presented will be useful for experts working in the field of WWER safety, and in particular to those planning, executing or reviewing studies related to the subject. Refs, figs and tabs.

  10. Experimental design verification of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia). Report of the IAEA technical co-operation project RER/9/004 on evaluation of safety aspects of WWER-440 model 213 nuclear power plants

    International Nuclear Information System (INIS)

    1995-07-01

    This publication addresses the experimental research supporting the design of WWER-440 model 213 plants. it is hoped that the material presented will be useful for experts working in the field of WWER safety, and in particular to those planning, executing or reviewing studies related to the subject. Refs, figs and tabs

  11. Safety aspects of neutron noise diagnostics and loose parts monitoring in WWER reactors

    International Nuclear Information System (INIS)

    Por, G.

    1997-01-01

    In this paper our aim is to give very short introduction on different types of well selected noise diagnostics methods and then mentioning their occurrence in WWER reactors we analyze what impact they might have to operational safety and for ageing (which also affects on safety of the installations). We do not deny, that one of our main aim is to call the attention of management staff of NPP, which deals with safety, safety culture, maintenance and operation proving, that such methods and system can give not only benefit to economy but also impact on safety of nuclear installations. 14 refs, 6 figs

  12. The behaviour of WWER nuclear power stations under abnormal operational conditions

    International Nuclear Information System (INIS)

    Ackermann, G.; Dreger, P.; Prasser, H.M.; Reichenbach, D.

    1985-01-01

    The big power WWER type reactors can show the volumetric oscillations of the power density into the reactor core because of xenon poisoning. The ununiform xenon distribution occurs in the case of nonideal mixing of the coolant into the reactor vessel. This effect usually leads to decreasing of the reactor power. The theoretical and experimental investigations with the goal to evaluate the volumetric distribution of the temperature near the coolant input into the reactor core is discussed in this paper as well as the three dimensional model of the reactor core together with the primary coolant circuit model to calculate volumetric xenon oscillations

  13. 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

  14. 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

  15. Experience in decontamination of the equipment of NPP's with the WWER-440 reactors

    International Nuclear Information System (INIS)

    Balaban-Irmenin, Yu.V.

    1981-01-01

    Different methods of decontamination at NPPs are briefly characterized. Decontamination of the removable part of the main circulation pump (MCP) of the WWER-440 reactor is considered as an example of removable equipment decontamination. A design of the decontamination bath of the removable MCP elements and the applied chemical agents are described. A decontamination flowsheet of the Novovoronezh NPP steam generator (SG) is considered as an example of the autonomic decontamination system. The SG decontamination modes, principal flowsheets of a hydromonitor, steam-ejection sprayer and steam-emulsion device are described [ru

  16. 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

  17. Implementation of the KASKAD computer code system for WWER-440 at Kozloduy NPP

    International Nuclear Information System (INIS)

    Antonov, A.; Georgieva, N.; Spasova, V.

    2003-01-01

    Since 2002 at Kozloduy NPP - EP1 the code package KASKAD is used for WWER-440 reactor core calculations. The main codes entering this package are: BIPR-7A: 3-D diffusion and core analysis code; PERMAK-A: 2-D fine mesh diffusion code. The burnup calculations were performed for all cycles of the Kozloduy NPP Unit 1, Unit 2, Unit 3 and Unit 4. For the last 4-5 cycles of the Units were calculated control rods worth, critical boron concentration at zero power, reactivity coefficients and linear power. These results were analysed and were compared with experimental data. Some results were given in this paper

  18. Radioactive waste management in nuclear power plants with WWER-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dlouhy, Z; Napravnik, J; Safar, O

    1975-05-01

    The possibilities of radioactive waste solidification in nuclear power plants with LWR reactors (of the WWER type) and the problems of their safe storage in Czechoslovakia are discussed. The most suitable method for the treatment of emitted sorbents and concentrates seems to be their incorporation in bitumen or concrete. In the disposal of solidified blocks all requirements should be met including the selection of suitable sites and of convenient methods of transportation. A preliminary economic estimate shows that the storage of bitumen-incorporated wastes in trenches seems to be less expensive from the point of view of exploitation of the storage facility as well as from the point of view of investment.

  19. SEJV2 software package for radiation monitoring system of WWER 440 NPP

    International Nuclear Information System (INIS)

    Kapisovsky, V.; Jancik, O.; Kubik, I.; Bena, J.

    1993-01-01

    The main part of the radiation monitoring system at a WWER-440 (213 reactor type) nuclear power plant is the centralized 400-channel monitoring system 'SEJVAL' servicing twin reactor units. The SEJV2 software package is described developed to run on a PC with an IFS2 interface to the SEJVAL radiation monitoring system. It provides enhanced data presentation, record keeping and report generation, thus improving the efficiency of the health physics shift. The system was for the first time implemented at the Jaslovske Bohunice V-2 nuclear power plant with encouraging results. (Z.S.) 3 refs

  20. Procedure for validating the life extension for the WWER-440 internals at NV NPP unit 3

    International Nuclear Information System (INIS)

    Filatov, V.M.; Evropin, S.V.

    2001-01-01

    The design lifetime (30 years) of the 1st generation WWER-440 reactor facilities is nearing completion in Russia. One of the major problems is to validate the life extension (LE) of the reactor internals ensuring the core arrangement and free passage of the control and protection system components during different operating modes, emergency modes included. The internals at the 1st generation units are designed so that to enable their replacement. But it requires a lot of funds and time. The work has been done to demonstrate that the internals may be further safely operated without their components being replaced provided their strength, longevity and serviceability are sufficiently validated. (author)

  1. 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.

  2. 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.)

  3. 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

  4. Renovation of the sealing planes of WWER-400 reactors pressure vessel

    International Nuclear Information System (INIS)

    Jablonicky, P.; Pilat, P.

    2007-01-01

    An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

  5. International workshop on WWER-440 reactor pressure vessel embrittlement and annealing. Working material

    International Nuclear Information System (INIS)

    1994-01-01

    The purpose of the Workshop was essentially to discuss the WWER 440 model 230 reactor pressure vessel integrity in terms of the measures already taken, current activities and future plans. The meeting was arranged in two parts, namely, the Scientific programme followed by the consideration, review and revision of the IAEA Consultancy report on RPV Embrittlement and Annealing. This particular report covers the first part of the meeting i.e., the Scientific Programme, in the form of proceedings of the meeting, while the re-drafted Consultancy report will be issued later. The meeting was attended by sixty-six representatives from thirteen countries. Refs, figs and tabs

  6. Status and future perspectives of PWR and comparing views on WWER fuel technology

    International Nuclear Information System (INIS)

    Weidinger, H.

    2003-01-01

    The main purpose of this paper is to give an overview on status and future perspectives of the Western PWR fuel technology. For easer understanding and correlating, some comparing views to the WWER fuel technology are provided. This overview of the PWR fuel technology of course can not go into the details of the today used designs of fuel, fuel rods and fuel assemblies. However, it tries to describe the today achieved capability of PWR fuel technology with regard to reliability, efficiency and safety

  7. Study on regimes of nuclear power plants with WWER-type reactors

    International Nuclear Information System (INIS)

    Akkerman, G.; Khampel', R.; Khentshel', G.; Kertsher, F.; Lyuttsov, K.

    1976-01-01

    The problems are considered of optimization of nuclear fuel loading, the peculiarities of the NPP operation at decreased power, and also the problem of stability operation of NPP with WWER type reactors taking into account specific features of these reactors (partial fuel overloads, change in reactor reactivity with power changes). The two particular interconnected problems discussed are: choice of such a sequence of partial rechargings which ensures the minimum cost of the electric power generated, and increasing the reactor operating time by reducing its power output. Besides the technical and economic estimates, much attention is given to analysing the stability of NPP operation

  8. Modelling of WWER fuel rod during LOCA conditions using FEM code ANSYS

    International Nuclear Information System (INIS)

    Bogatyr, S. M.; Krupkin, A. V.; Kuznetsov, V. I.; Novikov, V. V.; Petrov, O. M.; Shestopalov, A. A.

    2013-01-01

    The report presents the results of the computer simulation of the IFA-650.6 experiment, the sixth test in Halden LOCA test project series, performed in May 18, 2007 with a pre-irradiated WWER-440 fuel with maximum burnup of 56 MWd/kgU. The thermo-mechanical analysis was fulfilled with the license finite element ANSYS code package.The calculation was carried out with the 2D axisymmetric and 3D problem definitions. Analysis of the calculational results shows that the ANSYS code can adequately simulate thermo-mechanical behavior of cladding under IFA-650.6 test conditions. (authors)

  9. 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.

  10. In-service diagnostics of main circulating circuit pipes of WWER nuclear power plants

    International Nuclear Information System (INIS)

    Svoboda, V.; Merta, J.; Merta, V.

    1982-01-01

    The application is discussed of the acoustic emission method for testing the integrity of the components of the main circulating circuit of the WWER 440 nuclear power plant. A description is given of the main circulating circuit and a stress analysis on the basis of strength computations considering operating modes is presented. An analysis is also presented of the possible damage of the pipe material as related to the application of the acoustic emission method for in-service inspection of the pipes. Certain practical problems of application are discussed. (author)

  11. Calculation of spatial weight functions for WWER-440 ex-core neutron detectors

    International Nuclear Information System (INIS)

    Csom, Gy.; Czifrus, Sz.; Feher, S.; Berki, T.

    2001-01-01

    The objective of the work presented in this paper was determination of a spatial weight function for WWER-440 ex-core detectors to be used for the interpretation of reload startup rod drop measurements. In view of the complexity of the geometry of the core as well as the detector, furthermore the presence of a cavity between the vessel and the concrete shield, Monte Carlo calculations were applied. In spite of the fact that in the corresponding literature the use of adjoint methods dominates, in the present case the forward method was chosen and implemented using MCNP4C (Authors)

  12. Probabilistic safety assessment of WWER440 reactors prediction, quantification and management of the risk

    CERN Document Server

    Kovacs, Zoltan

    2014-01-01

    The aim of this book is to summarize probabilistic safety assessment (PSA) of nuclear power plants with WWER440 reactors and  demonstrate that the plants are safe enough for producing energy even in light of the Fukushima accident. The book examines level 1 and 2 full power, low power and shutdown PSA, and summarizes the author's experience gained during the last 35 years in this area. It provides useful examples taken from PSA training courses the author has lectured and organized by the International Atomic Energy Agency. Such training courses were organised in Argonne National Laboratory (

  13. Development and using computer codes for improvement of defect assembly detection on Russian WWER NPPs

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.; Zborovskii, V.; Kanukova, V.; Sorokin, A.; Taran, M.; Ugrumov, A.; Riabinin, Y.

    2009-01-01

    Diagnostic methods of fuel failure detection for improving the radiation safety and shortening of fuel reload time at Russian WWERs are currently in development . The works include creation new computer means for increase of effectiveness of fuel monitoring and reliability of leakage tests. Reliability of failure detection can be noticeably improved when we apply an integrated approach including the following methods. The first is fuel failure analysis under operating conditions. Analysis is performed with the pilot version of the expert system, which has been developed on the basis of the mechanistic code RTOP-CA. The second stage of failure monitoring is 'sipping' tests in the mast of the refueling machine. The leakage tests are the final stage of failure monitoring. A new technique with pressure cycling in the specialized casks was introduced to meet the requirements of higher reliability in detection/confirmation of the leakages. Measurements of the activity release kinetics during the pressure cycling and handling of the acquired data with the RTOP-LT code enable to evaluate a defect size in leaking fuel assembly. The mechanistic codes RTOP-CA and RTOP-LT were verified on a base of specialized experimental data and currently the code were certified by Russian authorities Rostechnadzor. Now the pressure cycling method in the specialized casks has official status and is utilized at the all Russian WWER units. Some results of application of the integrated approach to fuel failure monitoring at several Russian NPPs with WWER units are reported in the present paper. Predictions of the current version of the expert system are compared with the results of the leakage tests and with the estimations of the defect size by the pressure cycling technique. Using the RTOP-CA code the level of activity is assessed for the following fuel campaign if the leaking fuel assembly was decided to be reloaded into the core. A project of the automated computer system on the basis of

  14. Proposal of criteria for evaluation of engineering safety factors of WWER core parameters

    International Nuclear Information System (INIS)

    Shishkov, L.; Tsyganov, S.; Dementiev, V.

    2009-01-01

    The paper states that the regulatory documentation, as a rule, do not give explicit recommendations on formation techniques of engineering safety factors for design limited parameters of normal operation. The AER countries use different approaches to evaluation (sometimes even one country in relation of various power units). The paper suggests the development of uniform rules to be used in calculation of engineering safety factor for all WWER reactors. The paper presents principal problems that must be solved in the course of the discussion, and in the form of an exercise suggests the way of their solution. (Authors)

  15. Simulation of the WWER-440/213 maximum credible accident at the EhNITs stand

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, O.I.; Melikhov, V.I.; Davydov, M.V.; Sokolin, A.V.; Shchepetil'nikov, Eh.Yu.

    2000-01-01

    The calculations of thermohydraulic processes through the ATHLET code for determining optimal conditions for modeling the coolant leakage at the EhNITs stand by the maximum credible accident at the NPP with WWER-440/213 reactor are presented. The diameters of the nozzle at the stand, whereby the local criterion of coincidence with the data on the NPP (by the maximum flow) and integral criterion of coincidence (by the mass and energy of the coolant, effluent during 10 s) are determined in the process of parametric calculations [ru

  16. Modelling the WWER-type reactor dynamics using a hybrid computer. Part 1

    International Nuclear Information System (INIS)

    Karpeta, C.

    Results of simulation studies into reactor and steam generator dynamics of a WWER type power plant are presented. Spatial kinetics of the reactor core is described by a nodal approximation to diffusion equations, xenon poisoning equations and heat transfer equations. The simulation of the reactor model dynamics was performed on a hybrid computer. Models of both a horizontal and a vertical steam generator were developed. The dynamics was investigated over a large range of power by computing the transients on a digital computer. (author)

  17. Experience in lifetime extension of the first generation WWER-440 power units

    International Nuclear Information System (INIS)

    Medvedev, P.

    2002-01-01

    In connection with the expiration of the lifetime for the first generation WWER-440 reactors in Russian Federation (Novo Voronezh and Kola NPP), the legal procedures and Life Time Extension (LTE) Program are discussed. The LTE Program includes: development of regulation basis; economic efficiency studies; power unit modernization; power unit comprehensive examination and justification od equipment resource; in-dept safety assessment; operational license acquisition. As a result from the LTE Program the safety level of the unit 3 of the Novo Voronezh NPP is significantly increases, the operational period has been justified and a 5-year license has been issued

  18. Computer simulation of WWER - 440 normal and emergency transient operating conditions

    International Nuclear Information System (INIS)

    Izbeshesku, M.; Rajka, V.; Untaru, S.; Dumitresku, A.; Paneh, M.; Turku, I.

    1976-01-01

    Results of computer realization of a model for studying transient process in the nuclear system of vapour production at the WWER - 40 peactor nuclear power plant are presented. The first circuit model consists of a number of modules, corresponding to its main parts: for each module derived were the equations describing neutron and thermohydraulic parameters. The second circuit effect is taken into account by heat quantity accepted from a steam generator. The equations are mostly differential with constant coefficients. Coefficient values and initial values of physical quantities are evaluated according to the technical literature. Both manual and automatic operations are modelled [ru

  19. Improvement of methods to evaluate brittle failure resistance of the WWER reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Popov, A A; Parshutin, E V [Engineering Center of Nuclear Equipment Strength, Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Rogov, M F; Dragunov, U G [Experimenter` s and Designer` s Office ` ` Hydropress` ` (Russian Federation)

    1997-09-01

    At the next 10 years a number of Russian WWER nuclear power plants will complete its design lifetime. Normative methods to evaluate brittle failure resistance of the reactor pressure vessels used in Russia have been intended for design stage. The evaluation of reactor pressure vessel lifetime in operation stage demands to create new methods of calculation and new methods for experimental evaluation of brittle failure resistance degradation. The main objective of the study in this type of reactor is weldment number 4. In this report an analysis is made of methods to determine critical temperature of reactor materials including the results of instrumented Charpy testing. 12 figs.

  20. 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

  1. 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

  2. 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

  3. Evaluation of radiation exposures to personnel during maintenance operations and fuel recharging at NPP with the WWER-440 reactor

    International Nuclear Information System (INIS)

    Beskrestnov, N.V.; Vasil'ev, Eh.S.; Kozlov, V.F.; Odinokov, Yu.Yu.; Romanov, V.P.; Tsypin, S.G.

    1983-01-01

    A unified data acquisition and analysis system is presented. The system is intended to assess radiation exposures to personnel and perform radiation monitoring during periodic maintenance operations sna fuel recharging at NPPs with WWER-440 reactors. The basic principles of developing this system, patterns of danita collection are considered, points of radiation motoring chosen with account of the NPP operating experience are pointed out

  4. 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.

  5. 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)

  6. Investigation of large grain and Gd-doped WWER fuels behaviour at BOL in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.

    2008-01-01

    In this paper the following issues have been discussed: 1) WWER fuel tests in the HBWR; 2) Main objectives of the test with large grains and Gd-doped WWER fuel; 3) Analysis of of the the data at BOL focus on: Gd-doped fuel thermal behaviour, fuel elongation and dimension stability as well as cladding elongation early in life. At the end authors concluded that: 1) No indication of substantial effect of large grains on fuel thermal performance at BOL; 2) Densification observed in large grain fuel is similar to the ordinary uranium dioxide fuel with 95-96 % of theoretical density; 3) Dimension stability of large grain fuel is similar or even better than that in reference WWER fuel; 4) More stable dimension behaviour of large grain fuel at power could be attributed to its lower creep or densification at high temperature in the centre part of the fuel; 5) Cladding elongation detectors indicated identical early-in-life PCMI in both large grain and reference fuel rods, which reflected an accommodation effect of fuel pellets in claddings during first rise to power; no residual strains in either fuel types were observed; subsequent cladding elongation measurements show a trend to irradiation growth; 6) No clear evidence for densification of Gd-doped WWER fuel is observed during first irradiation cycle

  7. 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.

  8. Influence of decontamination of the WWER-440 primary circuit equipment on pressure drop in the reactor

    International Nuclear Information System (INIS)

    Kritsky, V.; Rodionov, Y.; Beresina, I.

    2003-01-01

    Over 40 reactor cycles at four WWER-440 type reactors have been analyzed in order to explain the increase of the pressure drop under certain combination of conditions. It is shown that the staff radiation exposure and the dose rate at first circuit segments are inversely correlated with the value of the pressure drop at the reactor, which is connected with the mechanism of redistribution of deposits and radioactive nuclides between the reactor and the rest part of the circuit. The influence of pH T on the formation of the dose rate from equipment and the change of pressure drop in the reactor WWER-440 is studied. The optimal range of pH T values for these parameters is determined to be 6.95-7.05 and these values are within the range of the water chemistry standards. The correlation between the changes of pressure drop and the number of decontaminated steam generators is established. This correlation shows that the pressure drop at the reactor grows with the increase of steam generators decontaminated during a preventive maintenance

  9. Local linear heat rate ramps in the WWER-440 transient regimes

    International Nuclear Information System (INIS)

    Brik, A.N.; Bibilashvili, Ju.L.; Bogatyr, S.M.; Medvedev, A.V.

    1998-01-01

    The operation of the WWER-440 reactors must be accomplished in such a way that the fuel rods durability would be high enough during the whole operation period. The important factors determining the absence of fuel rod failures are the criteria limiting the core characteristics (fuel rod and fuel assembly power, local linear heat rate, etc.). For the transient and load follow conditions the limitations on the permissible local linear rate ramp are also introduced. This limitation is the result of design limit of stress corrosion cracking of the fuel cladding and depends on the local fuel burn-up. The control rod motion is accompanied by power redistribution, which, in principle, can result in violating the design and operation limitations. Consequently, this motion have to be such as the core parameters, including the local ramps of the linear heat generation rates would not exceed the permissible ones.The paper considers the problem of WWER-440 reactor control under transient and load follow conditions and the associated optimisation of local linear heat generation rate ramps. The main factors affecting the solution of the problem under consideration are discussed. Some recommendations for a more optimal reactor operation are given.(Author)

  10. Trends and results in In-Core management for the Kozloduy NPP WWER-440 reactors

    International Nuclear Information System (INIS)

    Haralampieva, Tz.; Antov, A.; Georgieva, N.; Spasova, V.

    2001-01-01

    The paper presents the experience gained during the design and operation of the last fuel cycles of the four WWER-440/V-230 units at Kozloduy NPP. High efficiency and economy of the fuel utilization requires very precise procedures for fuel in-core management, including calculations and analyses for reloading scheme design, compared with results from operational measurements and fuel cycle efficiency. The paper describes the main stages of implementation of advanced fuel assemblies in the Kozloduy NPP WWER-440 reactors. New advanced fuel has been implemented after the completion of comprehensive neutron-physical, thermal-hydraulic and thermal-mechanical analyses by using advanced computer codes. As a general task of the fuel cycle improvements it is pointed the increasing of the final fuel burnup and decreasing of the number of spent fuel assemblies. Series of calculations and analyses, related to the introducing of the advanced fuel assemblies and improvement of the fuel cycle characteristics have been carried out to guarantee the safe operation and fuel reliability

  11. 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

  12. Qualification of the APOLLO2 lattice physics code of the NURISP platform for WWER hexagonal lattices

    International Nuclear Information System (INIS)

    Hegyi, G.; Kereszturi, A.; Tota, A.

    2011-01-01

    The experiments performed at the ZR-6 zero critical reactor by the Temporary International Collective and a numerical assembly burnup benchmark specified for depletion calculation of a WWER-440 assembly containing gadolinium burnable poison were used to qualify the APOLLO2 (APOLLO2.8-E3) code as a part of its ongoing validation activity. The work is part of the NURISP project, where KFKI Atomic Energy Research Institute undertook to develop and qualify some calculation schemes for hexagonal problems. Concerning the ZR-6 measurements, single cell, macro cell and two-dimensional calculations of selected regular and perturbed experiments are being used for the validation. In the two-dimensional cases the radial leakage is also taken into account in the calculations together with the axial leakage represented by the measured axial buckling. Criticality parameter and reaction rate comparisons are presented. Although various sets of the experiments have been selected for the validation, good agreement of the measured and calculated parameters could be found by using the different options offered by APOLLO2. An additional mathematical benchmark-presented in the paper - also attests for the reliability of APOLLO2. All the test results prove the reliability of APOLLO2 for WWER core calculations. (Authors)

  13. Safety of reactors built according to earlier standards (WWER 440/V230 type)

    International Nuclear Information System (INIS)

    Misak, J.; Rohar, S.

    1995-01-01

    The problems of safety of WWER-440/V-230 type reactors are discussed, and the following conclusions are made. (1) The reactors have a very good operational record. (2) The reactors have serious design shortcomings, which should be eliminated by safety upgrading. Core damage frequency should be further reduced. (3) PSA methods constitute an appropriate tool for assessment of plant vulnerability to some initiating events and malfunctions, for prioritization of upgrading measures and for tolerability of deviations from current safety standards. (4) The most important safety merits, such as a large thermal inertia and low rupture probability, should be properly taken into account in the analysis. (5) Extensive safety upgrading is feasible and can lead to a considerable risk reduction. In certain circumstances such upgrading is the least expensive option even though the total cost is much higher than the initial plant construction cost. (6) Properly upgraded, the reactor units may be operable until better power resources are available within the country. (7) The existing gap between the technological and political judgements of nuclear safety should be reduced continuously by information exchange improvements. (8) A unified approach to nuclear safety should be adopted for all nuclear reactors (not just WWERs) built to earlier standards. 5 tabs., 1 fig

  14. 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.

  15. 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.

  16. Power release estimation inside of a fuel pin neighbouring a WWER-440 control rod

    International Nuclear Information System (INIS)

    Mikus, J.

    2006-01-01

    This work presents an estimation of the control rod (CR) influence in the WWER-440 core on the power release inside of a fuel pin neighbouring CR, that can have some consequences due to possible static and cyclic loads, for example fuel pin / fuel assembly bowing. For this purpose detailed (usual) axial power distribution measurements were performed in a WWER-440 type core on the light water, zero-power research reactor LR-0 in fuel pins near to an authentic CR model at zero boron concentration in moderator, modelling the conditions at the end of fuel cycle. To demonstrate the CR influence on power distribution inside of one fuel pin neighbouring CR, results of above measurements were used for estimation of the: 1) Axial power distribution inside of the investigated fuel pin in both opposite positions on its pellets surface that are situated to- and outwards CR and corresponding gradient of the (r, z) - power distribution in above opposite positions and 2) Azimuthal power distributions on pellet surface of the investigated fuel pin in horizontal planes at selected axial coordinates. Similar information can be relevant from the viewpoint of the fuel pin failures occurrence investigation

  17. 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)

  18. Methodology and conclusions of activation calculations of WWER-440 type nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Babcsány, Boglárka, E-mail: boglarka.babcsany@reak.bme.hu; Czifrus, Szabolcs; Fehér, Sándor

    2015-04-01

    Highlights: • Activation calculation of two WWER-440 type nuclear power plants. • Detailed description of the applied activation calculation methodology. • Graphical results for total activity and waste index categorization. • General conclusions for activation applicable in the case of PWR reactors. - Abstract: Activation calculations for two nuclear power plants of WWER-440 type have been performed by the authors in order to assist the decommissioning planning by assessing the radioactive inventory present at the time of and at different times after the final shutdown. According to related international literature and studies performed earlier by the authors, considering the activity more than 99% of this inventory is concentrated in the materials directly surrounding the reactor core, where the predominant evolution of radionuclides is generated by neutron induced nuclear reactions. In order to obtain the highest possible accuracy in modelling, three-dimensional Monte Carlo neutron transport calculations were performed. Besides the methods and models applied to these analyses, the paper also summarizes the results that can be generally applied to such nuclear power plant types. At the time of shutdown, the total activity of the stainless steel components is about 6 × 10{sup 16} Bq and 1.3 × 10{sup 17} Bq for the two NPPs considered. The biological shielding concrete constitutes approximately 7 × 10{sup 13} Bq and 1.1 × 10{sup 14} Bq.

  19. WWER-440 local power peaking experiment with/without Hf inserts in the LR-0 reactor

    International Nuclear Information System (INIS)

    Josek, R.; Hudec, F.; Rypar, V.

    2006-01-01

    One of the known issues of the WWER-440 reactors is the control rod coupler induced power peaking in the neighbouring fuel assemblies. The effect has been discovered some years ago and is believed to be the cause of several fuel failures during operation in WWER-440 reactors. The effect itself is due to over-moderation and small absorption in the region of control rod coupler, leading to increase in thermal neutron flux and hence to power flare-up in the neighbouring fuel pins. The fuel vendor tackled the problem by attaching hafnium inserts on the inside of the control assembly box. The experiment performed in the LR-0 reactor was focused on the axial and radial power profiles in the vicinity of the control assembly with and without the hafnium inserts. The results of measurements with zero boron concentration are presented. The hafnium insert causes a decrease in peaking factor of about 30% in selected pins close to the control assembly. The measurements are compared with calculations performed with the MCNP-4C code. The compared variables are: the axial fission density distributions; peking factors and peaking factor decrease due to Hf insert. The MCNP results are accurate with respect to the experimental results. A series of benchmarks is being prepared on the basis of these measurements

  20. Report of a consultants meeting on draft guidelines for WWER 440/213 containment evaluation

    International Nuclear Information System (INIS)

    Andrieu, R.; Barre, F.; Birbraer, A.N.

    1995-01-01

    This report has been prepared at two Consultants' Meetings of experts from France, Germany, Russia, Slovakia, Spain and Ukraine, organized within the framework of the IAEA Extrabudgetary Programme on the Safety of WWER NPPs. In view of the limited time available and the complexity of the task the present report should be treated as a draft, which will be sent to the national regulatory bodies for review and further improvement. The next meeting on the subject will be held in June 1996 with the participation of representatives of all countries operating WWER 440/213 reactors with bubbler condenser containments. During the meeting, the acceptance criteria and review requirements formulated in the IAEA NUSS Series and those in force in the Russian Federation, the United States of America, Germany, France, the Czech Republic, and Slovakia have been taken into consideration. The equivalence of the technical goals which are aimed to be achieved by applying comparable national criteria has been checked in order to avoid multiple addressing the issues and to assure the completeness of the acceptance criteria necessary to assess the design of the containment systems. 48 refs, 12 tabs

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. Experimental and computation method for determination of burnup and isotopic composition of the WWER-440 fuel using the 134Cs and 137Cs concentrations

    International Nuclear Information System (INIS)

    Babichev, B.A.; Kozharin, V.V.

    1990-01-01

    An experimental and computational method for determination of burnup and actinoid concentrations in WWER fuel elements using 134 Cs and 137 Cs concentrations in fuel is considered. It is shown that the error in calculation of fuel burnup and U and Pu isotope concentrations in WWER-440 fuel elements is 1.3-4.9% provided that the error in 134 Cs and 137 Cs concentration measurements does not exceed 1.7 and 1.2%. 9 refs.; 10 figs.; 4 tabs

  6. 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

  7. Post mortem investigations of the NPP Greifswald WWER-440 reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Schuhknecht, Jan; Viehrig, Hans-Werner; Weiss, Frank-Peter; Rindelhardt, Udo [Forschungszentrum Dresden-Rossendorf e.V., Dresden (Germany). Inst. for Safety Research; Keller, Werner [Studsvik GmbH und Co. KG, Stutensee (Germany)

    2008-07-01

    The paper presents first results of the post mortem investigations performed at the reactor pressure vessels (RPV) of the Russian WWER-440 type reactors. Trepans were taken from the core weld SN0.1.4 and base metal of the unit 1 RPV. This RPV was annealed after 15 years of operation and operated for two more years. At first the trepan of the core welding seam was investigated by Master Curve (MC) testing. Specimens from 5 locations through the thickness of the welding seam were tested according to ASTM E1921-05. The reference temperature T{sub 0} was calculated with the measured fracture toughness values, K{sub Jc}, at brittle failure of the specimen. Generally the K{sub Jc} values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follow the course of the Master Curve. The K{sub Jc} values show a remarkable scatter. More values than expected lie below the 5% fractile. In addition the MC SINTAP procedure was applied to determine T{sub 0}{sup SINTAP} of the brittle fraction of the data set. There are remarkable differences between T{sub 0} and T{sub 0}{sup SINTAP} indicating macroscopic inhomogeneous weld metal. The highest T{sub 0} was about 50 C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T{sub 0} at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material may not represent the most conservative condition. The results presented in this paper show that the Master Curve approach as adopted in the test standard ASTM E1921-05 is applicable to the investigated WWER-440 multilayer weld metal. The results are of direct importance for an advanced WWER-440 RPV integrity assessment. On the other hand the data pool is broadened for a general introduction of the MC based RPV integrity assessment in the national codes. Furthermore general experiences in

  8. Post mortem investigations of the NPP Greifswald WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    Schuhknecht, Jan; Viehrig, Hans-Werner; Weiss, Frank-Peter; Rindelhardt, Udo

    2008-01-01

    The paper presents first results of the post mortem investigations performed at the reactor pressure vessels (RPV) of the Russian WWER-440 type reactors. Trepans were taken from the core weld SN0.1.4 and base metal of the unit 1 RPV. This RPV was annealed after 15 years of operation and operated for two more years. At first the trepan of the core welding seam was investigated by Master Curve (MC) testing. Specimens from 5 locations through the thickness of the welding seam were tested according to ASTM E1921-05. The reference temperature T 0 was calculated with the measured fracture toughness values, K Jc , at brittle failure of the specimen. Generally the K Jc values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follow the course of the Master Curve. The K Jc values show a remarkable scatter. More values than expected lie below the 5% fractile. In addition the MC SINTAP procedure was applied to determine T 0 SINTAP of the brittle fraction of the data set. There are remarkable differences between T 0 and T 0 SINTAP indicating macroscopic inhomogeneous weld metal. The highest T 0 was about 50 C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T 0 at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material may not represent the most conservative condition. The results presented in this paper show that the Master Curve approach as adopted in the test standard ASTM E1921-05 is applicable to the investigated WWER-440 multilayer weld metal. The results are of direct importance for an advanced WWER-440 RPV integrity assessment. On the other hand the data pool is broadened for a general introduction of the MC based RPV integrity assessment in the national codes. Furthermore general experiences in the cutting of irradiated RPV steels are collected

  9. On the reliability of steam generator performance at nuclear power plants with WWER type reactors

    International Nuclear Information System (INIS)

    Styrikovich, M.A.; Margulova, T.Kh.

    1974-01-01

    The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

  10. 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)

  11. 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)

  12. Influence of delayed neutron parameter calculation accuracy on results of modeled WWER scram experiments

    International Nuclear Information System (INIS)

    Artemov, V.G.; Gusev, V.I.; Zinatullin, R.E.; Karpov, A.S.

    2007-01-01

    Using modeled WWER cram rod drop experiments, performed at the Rostov NPP, as an example, the influence of delayed neutron parameters on the modeling results was investigated. The delayed neutron parameter values were taken from both domestic and foreign nuclear databases. Numerical modeling was carried out on the basis of SAPFIR 9 5andWWERrogram package. Parameters of delayed neutrons were acquired from ENDF/B-VI and BNAB-78 validated data files. It was demonstrated that using delay fraction data from different databases in reactivity meters led to significantly different reactivity results. Based on the results of numerically modeled experiments, delayed neutron parameters providing the best agreement between calculated and measured data were selected and recommended for use in reactor calculations (Authors)

  13. Calculation of the local power peaking near WWER-440 control assemblies with Hf plates

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Hordosy, G.; Kereszturi, A.; Maraszy, Cs.; Temesvari, E.

    2003-01-01

    The original coupler design of the WWER-440 assemblies had the following well known deficiency: The relatively large amount of water in the coupler between the absorber and fuel port of the control assembly can cause undesirably sharp power peaking in the fuel rods next to the coupler. The power peaking can be especially high after control rod withdrawal when the coupler reached low burnup level region of the adjacent assembly. The modernized coupler design overcomes the original problem by applying a thin Hf plate in the critical region. The very complicated structure of the coupler requires the verification of the core design methods by high precision 3D Monte Carlo calculations. The paper presents an MCNP reference calculation on the control rod coupler benchmark with Hf absorber plates. The benchmark solution with the KARATE-440 code system is also presented. The need for treating the Hf burnout in the reflector region is investigated (Authors)

  14. 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)

  15. Study on the power control system for NPP power unit with the WWER-440 reactor

    International Nuclear Information System (INIS)

    Aleksandrova, N.D.; Naumov, A.V.

    1981-01-01

    Results of model investigations into basic version of the power control systems (PCS) conformably to the WWER-440 NPP power unit are stated. Transient processes in the power unit system when being two PCS versions during perturbations of different parameters: unit power, vapour pressure or position of control rods have been simulated. Investigations into the different PCS versions show that quality of operation of a traditional scheme with a turbine power controller and reactor pressure controller can be significantly improved with the introduction of a high-speed signal of pressure into the reactor controller. The PCS version with the compensation of interrelations between the turbine and reactor controllers constructed according to the same principles as the standard schemes of power units of thermal electric power plant is perspective as well [ru

  16. Analysis of coolant flow in central tube of WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Zsiros, G.; Toth, S.; Attila Aszodi, A.

    2011-01-01

    Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

  17. Conception of programs for evaluating balance indices of operation of WWER-440 nuclear power plants

    International Nuclear Information System (INIS)

    Zadrazil, J.

    1986-01-01

    The procedures are discussed of writing computation programs for the evaluation of basic technical and economic parameters of the operation of WWER-440 nuclear power plants. The criterion of the evaluation is the maximum power supply to the grid together with the required supply of heat, with the observance of safe operating conditions. Previous procedures of evaluation are compared with present procedures based on the use of a monitoring and evaluation system of the KOMPLEX-URAN 2 M type. The mathematical model of the program is based on balance equations and relations derived for actual measuring systems and is completed with an assessment of the reliability of input and output data. The flow chart is shown of the algorithm for the evaluation of technical and economic parameters, and methods are suggested for improving and extending the program. (J.C.)

  18. 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.)

  19. Experimental verification of lifetime of bolting joints for WWER reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1992-01-01

    This paper presents results from experimental verification of cyclic lifetime of bolting joints of M 140x6 mm type used for WWER-440 MW reactor pressure vessels. Bolting joints or real dimensions were tested in a special testing equipment ZS 1000 in Skoda Concern. Stud bolts are made from 25Kh1MF or 38KhN3MFA type of steels. Tests were carried out at operating as well as at room temperatures with coefficient of asymmetry equal to 0.1; one tests was realized with given bending moment. Experimental results have been compared with calculated lifetimes according to ASME, Soviet and CMEA Codes. In all cases calculations give conservative assessments. (orig.)

  20. 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

  1. Long-term WWER-440 dynamics in cyclic power output changes

    International Nuclear Information System (INIS)

    Petruzela, I.

    1989-01-01

    Xenon poisoning is one of the main factors limiting the operation of a nuclear power plant with a WWER-440 reactor in the variable load mode, when long-term dynamics applies to cyclic power output changes. An analysis of the xenon poisoning linearized transfer shows that a phase shift of 180deg takes place between the summed-up reactivity change due to a power change and the reactivity change due to xenon poisoning, this for a sine-wave power change with a period of 24 hours. Thus, the requirements are minimized for the change in reactivity of the control elements, and the maximum value can be achieved of released reactivity that can be utilized before the end of the campaign. (B.S.). 6 figs., 4 tabs., 9 refs

  2. 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)

  3. Part-task simulator of a WWER-440 type nuclear power plant unit

    International Nuclear Information System (INIS)

    Palecek, P.

    1990-01-01

    In the present paper the design of a part-task simulator for WWER-440 type nuclear power plant units by the CEZ (Czech Power Works) Concern is reported. This part-task simulator has been designed for the training of NPP operating personnel. It includes a central computer that is coupled with the training work places and the trainer place. Interchange of information is performed by functional keyboards and semigraphical colour displays. The process is simulated, also in real time scale, on the basis of dynamic models. In addition to the precision of the models used, great importance has primarily been attached to plasticity of information presentation. The part-task simulator may be applied to simulation and demonstration as well as to teaching purposes. The paper presents the achieved state of implementation of the part-task simulator and points out some further stage of evolution. (author)

  4. Operation results of 3-rd generation nuclear fuel WWER-440 in initial period

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.

    2011-01-01

    On unit 4 of Kola NPP trial operation of 3-rd generation's fuel began in 2010. Fuel assemblies of 3-rd generation (FA-3) have a number of design features that provide better operational characteristics. Concise description of a design and the basic advantages of fuel of 3-rd generation are described in articles. Increasing of efficiency of nuclear fuel usage will be achieved by reduction of the parasitic capture of thermal neutrons in constructional materials (weight of zirconium is reduced), optimization of uranium-water relation (increase in fuel elements step), increasing of uranium loading (usage of fuel pellets with increased diameter and without central hole in them). By results of trial operation mass transition to use of given type of assemblies in WWER-440 is possible. This report presents the basic outcomes of the trial operation, a brief survey of the obtained data. The basic characteristics of the reactor core with fuel of 3-rd generation are resulted in work. (authors)

  5. Automated WWER steam generator eddy current testing and plugging control system

    International Nuclear Information System (INIS)

    Gorecan, I.; Gortan, K.; Grzalja, I.

    2004-01-01

    The structural architecture of the system contains three main components which are described as follows: Manipulator Guidance System; Eddy Current Testing System; Plugging System. The manipulator system has the task to position the end-effectors to the desired tube position. When the final position is reached, the Eddy Current testing system performs data acquisition. In case defects are found, the plugging system performs tube plug installment. Each system is composed of 3 layers. The first layer is the hardware layer consisting of motors driving the effectors along with sensors needed to obtain the positioning data, pusher motors used to push the test probes into tubes of the WWER steam generator, and plugging hardware tool. The second layer is the control box performing basic monitoring and control routines as an interconnection between first and third layer. The highest layer is the control software, running on the PC, which is used as a human-machine-interface.(author)

  6. Safety improvement and results of commissioning of Mochovce NPP WWER 440/213

    International Nuclear Information System (INIS)

    Lipar, M.

    1998-01-01

    Mochovce NPP is the last one of this kind and compared to its predecessors, it is characterized by several modifications which contribute to the improvement of the safety level. In addition based on Nuclear Regulatory Authority requirements and based on documents: - IAEA - Safety Issues and their ranking for NPP WWER 440/213, - IAEA - Safety Improvement of Mochovce NPP Project Review Mission, - Riskaudit - Evaluation of the Mochovce NPP Safety Improvements. Additional safety measures have been implemented before commissioning. The consortium EUCOM (FRAMATOME - SIEMENS), SKODA Praha, ENERGOPROJEKT Praha, Russian organizations and VUJE Trnava Nuclear Power Plants research institute were selected for design and implementation of the safety measures. The papers summarized, safety requirements, safety measures implemented, results of commissioning and results of safety analysis report evaluation. (author)

  7. A small angle neutron study of irradiation induced microstructures in Cr-Mo-V WWER steels

    International Nuclear Information System (INIS)

    Levit, Vladimir I.; Santos, Ari S.; Louzada, Ana R.R.; Silveira, Cristina M.; Vaniel, Ana Paula H.; Odette, George R.; Mader, Eric

    2000-01-01

    Small angle neutron scattering (SANS) has proven to be a very effective technique for characterizing the ultrafine (∼1 nm) irradiation induced microstructures which are responsible for hardening and the concomitant embrittlement of reactor pressure vessel steels. SANS measurement were carried out on three irradiated and unirradiated weld materials of WWER- type on 8 m instrument at the National Institute of Standards and Technology, Washington, USA. Small (r m < 1 nm) irradiation induced features were found for all three materials. Were found volume fractions, number densities and ratios of magnetic to nuclear scattering. Some analyses of the irradiation induced precipitation nature and possible chemical composition were made by comparison of the results with other reactor materials SANS and Atom Probe Field Ion Microscopy data. (author)

  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. Failed fuel diagnosis during WWER reactor operation using the RTOP-CA code

    International Nuclear Information System (INIS)

    Likhanskii, V.; Afanasieva, E.; Sorokin, A.; Evdokimov, I.; Kanukova, V.; Khromov, A.

    2006-01-01

    The mechanistic code RTOP-CA is developed for objectives of failed fuel diagnosis during WWER reactor operation. The RTOP-CA code enables to solve a direct problem: modelling the failed fuel behavior and prediction of primary coolant activity if characteristics of failures in the reactor core are known. Results of verification of the RTOP-CA code are presented. Separate physical models were verified on small-scale in-pile and out-of-pile experiments. Integral verification cases included data obtained at research reactors and at nuclear power plants. The RTOP-CA code is used for development of a neural-network approach to the inverse problem: detection of failure characteristics on the base of data on primary coolant activity during reactor operation. Preliminary results of application of the neural-network approach for evaluation of fuel failure characteristics are presented. (authors)

  10. 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)

  11. A possible way for conservatism reduction at the modelling of WWER operational modes

    International Nuclear Information System (INIS)

    Shishkov, L. K.

    2010-01-01

    The paper discusses the way to ensure safety operation of the WWER by meet ing the technological Safety Criteria that are defined in the documents of the state level. That Safety Criteria are ensured by-turn by implementation of the design limits for the number of core parameters at normal operation. Verification of the Safety Criteria adherences for all design modes of NPP operation is conducted in assumption that all (if possible) core parameters that have the design limits reach these limiting values. Additionally, to take into account possible error of modeling, a reactor operates in conditions for the core parameters to not reach the limits with probability greater than 95%. This method is deliberately conservative and it may be substituted with 'probabilistic method'. Probabilistic method supposes the operational parameters are defined with accounting of the possible random deviation to ensure the probability of technological Safety Criteria to be met. (Author)

  12. Specific features of the WWER Uranium-Gadolinium fuel behavior at BOL

    International Nuclear Information System (INIS)

    Shcheglov, A.; Proselkov, V.; Volkov, B.

    2013-01-01

    The calculated-experimental analysis of the WWER fuel behavior with 5%wt of gadolinium oxide at the beginning of life (BOL) is presented. The results are based on the data on fuel centerline temperature measurements, gas media pressure inside the cladding and fuel elongation obtained during irradiation of the test fuel rods in HBWR (Halden). Computer analysis of experimental data is performed with TOPRA-2, version 2 code. It is shown that specific features of the uranium-gadolinium fuel behavior at the early of life is due to presence of burnable absorber influencing the average linear heat rating, radial power distribution and lower thermal conductivity. In particular, the analysis of “late” relocation effect on the maximum Gd fuel temperature is presented. (authors)

  13. 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)

  14. 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.

  15. WWER fuel performance, development, QA and future prospects at Loviisa NPS

    Energy Technology Data Exchange (ETDEWEB)

    Loesoenen, P [Imatran Voima Oy, Vantaa (Finland)

    1994-12-31

    The essential characteristics of Loviisa fuel service are presented. A brief description is given of the steps in developing the performance values including: burnup dependent linear heat rate for fuel rods, comparison of measured and calculated cladding creep down of some rods, development of average discharge burnup for unit 1 and unit 2, results from experimental irradiation of test rods in research reactors. Some noticeable design changes performed and in-reactor behaviour of the lead fuel assemblies are also discussed. The quality assurance of the fuel procurement and operation is explained. Future prospects are connected with power raise of Loviisa reactors, discontinuing of spent fuel transportation to Russia for final disposal and new WWER fuel deliverers. 4 figs., 1 tab., 8 refs.

  16. Testing of acoustic emission method during pressure tests of WWER-440 steam generators and pressurizers

    International Nuclear Information System (INIS)

    Wuerfl, K.; Crha, J.

    1987-01-01

    The results are discussed of measuring acoustic emission in output pressure testing of steam generators and pressurizers for WWER-440 reactors. The objective of the measurements was to test the reproducibility of measurements and to find the criterion which would be used in assessing the condition of the components during manufacture and in operation. The acoustic emission was measured using a single-channel Dunegan/Endevco apparatus and a 16-channel LOCAMAT system. The results showed that after the first assembly, during a repeat dismantle of the lids and during seal replacement, processes due to seal contacts and bolt and washer deformations were the main source of acoustic emission. A procedure was defined of how to exclude new acoustic emission sources in such cases. The acoustic emission method can be used for the diagnostics of plastic deformation processes or of crack production and propagation in components during service. (Z.M.)

  17. 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.

  18. Development and validation of natural circulation based systems for new WWER designs

    International Nuclear Information System (INIS)

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

    2002-01-01

    Elaboration and introduction of NPP designs with improved technical and economic parameters are defined as an important element of the National Program of nuclear power development approved by the Russian Federation Government in 1998. This Program considers the designs of WWER-1000/V-392 and WWER-640/ V-407 power units as the priority projects of the new generation NPPs with increased safety. A number of passive systems based on natural circulation phenomena are used in V-392 and V-407 designs to prevent or mitigate severe accidents. Design basis, configuration and effect of some naturally driven systems of V-392 design sited at Novovoronezh are mainly reflected in the present paper. One of the most important mean for severe accident prevention in V-392 design is so called SPOT - passive heat removal system designed to remove core decay heat in case of station blackout (including failure of all diesel generators). This system extracts the steam from the steam generator, condenses it and returns water to steam generator by natural circulation. The SPOT heat exchangers are cooled by atmospheric air coming by natural circulation through a special direct action control gates which operate passively as well. Extensive experimental investigation of the different aspects of this system operation has been carried out to validate its functioning under real plant conditions. In particular, full-scale section of air heat exchanger-condenser has been tested with natural circulation steam, condensate and air paths modeled. The environment air temperature and steam pressure condensing were varied in the wide range, and the relevant experimental results are being discussed in this paper. The effect of wind velocity and direction to the containment is also checked by the experiments. (author)

  19. 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.

  20. 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)

  1. 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.

  2. '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)

  3. Development of destructive methods of burn-up determination and their application on WWER type nuclear fuels

    International Nuclear Information System (INIS)

    Hermann, A.; Stephan, H.; Nebel, D.

    1984-03-01

    Results are described of a cooperation between the Central Institute of Nuclear Research Rossendorf and the Radium Institute 'V.G. Chlopin' Leningrad in the field of destructive burn-up determination. Laboratory methods of burn-up determination using the classical monitors 137 Cs, 106 Ru, 148 Nd and isotopes of heavy metals (U, Pu) as well as the usefulness of 90 Sr, stable isotopes of Ru and Mo as monitors are dealt with. The analysis of the fuel components uranium (spectrophotometry, potentiometric titration, mass-spectrometric isotope dilution) and plutonium (spectrophotometry, coulometric titration, mass- and alpha-spectrometric isotope dilution) is fully described. Possibilities of increasing the reproducibility (automatic adjusting of measurement conditions) and the sensibility (ion impuls counting) of mass-spectrometric measurements are proposed and applied to a precise determination of Am and Cm isotopic composition. The methods have been used for burn-up analysis of spent WWER (especially WWER-440) fuel. (author)

  4. The requirements to informativity of FFD Methods on NPP with WWER at acceptance of the concept of zero refusal

    International Nuclear Information System (INIS)

    Miglo, V.; Girchenko, A.

    2015-01-01

    The paper reviewed current approaches to detect fuel failures during reactor operation and during refueling outages in WWER s. Generally, the diagnosis of leaking fuel is performed in three steps. First, failure parameters are estimated by coolant activity during reactor operation. Second, leaking fuel is detected by sipping in the mast of the refueling machine. Third, additional leakage test is performed in most WWER units to confirm the leak and sometimes to evaluate failure parameters (equivalent hydraulic size of the defect in cladding may be estimated during this activity). These additional leakage tests are performed in the special casks in the spent fuel pool. Uncertainties and limitations of analytical diagnosis of failure parameters during reactor operation are mainly due to a variety of operating conditions of fuel assemblies in the core (e.g. different and broader range of the heat rates, variety of fuel enrichments, and different amounts of Gd, mixed cores)

  5. Computerization of the nuclear material accounting system for safeguards purposes at nuclear power plants with WWER-440 reactors

    International Nuclear Information System (INIS)

    Antonov, V.P.; Konnov, Yu.I.; Semenets, A.N.

    1983-01-01

    The paper sets forth the basic principles underlying nuclear material accounting at nuclear power plants with WWER-440 reactors. It briefly describes the general structure and individual units in a program for computerized accounting. The use of this program is illustrated by the actual accounting data from the fifth unit of the Novovoronezh nuclear power station. The NUMIS program seems to be of interest both for the purposes of IAEA safeguards and for nuclear power plant operators in countries where power plants with WWER-440 reactors subject to IAEA safeguards are either in operation or under construction. The research in question was conducted initially under an IAEA research contract; the system is now being developed further and tested under the IAEA-USSR technical and scientific co-operation programme on safeguards. (author)

  6. 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

  7. Computer-based system for inspection of water chemistry regimes in WWER-type nuclear power plants

    International Nuclear Information System (INIS)

    Burcl, R.; Novak, M.; Malenka, P.

    1993-01-01

    The unsatisfactory situation in water chemistry testing at nuclear power plants with WWER type reactors is described. The testing primarily relies on laboratory analyses of manually taken samples. About 40 samples from one unit are tested per shift, which comprises approximately 250 determinations of various parameters. The time between two determinations is no shorter than 4 to 6 hours, thus rapid parameter changes between two determinations fail to be monitored. A novel system of automated chemistry monitoring is outlined, feasible for WWER type reactors. The system comprises 10 sets of sensors for monitoring all the relevant chemistry parameters of both the primary and secondary coolant circuits. Each sensor set has its own autonomous computer which secures its function even in case of loss of the chemical information network. The entire system is controlled by a master computer which also collects the results and provides contact with the power plant's information system. (Z.S.). 1 fig

  8. MOST-7 program for calculation of nonstationary operation modes of the nuclear steam generating plant with WWER

    International Nuclear Information System (INIS)

    Mysenkov, A.I.

    1979-01-01

    The MOST-7 program intended for calculating nonstationary emergency models of a nuclear steam generating plant (NSGP) with a WWER reactor is considered in detail. The program consists of the main MOST-7 subprogram, two main subprograms and 98 subprograms-functions. The MOST-7 program is written in the FORTRAN language and realized at the BESM-6 computer. Program storage capacity in the BESM-6 amounts to 73400 words. Primary information input into the program is carried out by means of information input operator from punched cards and DATA operator. Parameter lists, introduced both from punched cards and by means of DATA operator are tabulated. The procedure of calculational result output into printing and plotting devices is considered. Given is an example of calculating the nonstationary process, related to the loss of power in six main circulating pumps for NSGP with the WWER-440 reactor

  9. 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)

  10. 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)

  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. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report.

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V.

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  13. In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V

    1999-04-15

    The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

  14. Investigation of neutron physical features of WWER-440 assembly containing differently enriched pins and Gd burnable poison

    International Nuclear Information System (INIS)

    Nemes, Imre

    2000-01-01

    In this paper different pin-distributions of WWER-440 fuel assembly are examined. Assemblies contain 3 Gd-doped pins (Hungarian design), 6 Gd-doped pins near the assembly corners (Russian design) and differently profiled U5-enrichment in different pins. The main neutron physical characteristics of this assemblies - as the function of burnup - are calculated using HELIOS code. The calculated parameters of different assembly designs are analyzed from the standpoint of fuel cycle economy and refueling design practice. (Authors)

  15. Study of multiplication factor sensitivity to the spread of WWER spent fuel isotopics calculated by different codes

    International Nuclear Information System (INIS)

    Markova, L.

    2001-01-01

    As a sensitivity study the impact on the system reactivity was studied in the case that different calculational methodologies of spent fuel isotopic concentrations were used for WWER spent fuel inventory computations. The sets of isotopic concentrations obtained by calculations with different codes and libraries as a result of the CB2 international benchmark focused on WWER-440 burnup credit were used to show the spread of the calculated spent fuel system reactivity. Using the MCNP 4B code and changing the isotopics input data, the multiplication factor of an infinite array of the WWER-440 fuel pin cells was calculated. The evaluation of the results shows the sensitivity of the calculated reactivity to different calculational methodologies used for the spent fuel inventory computation. In the studied cases of the CB2 benchmark, the spread of the reference k-results relative to the mean was found less or about ±1% in spite of the fact that the data of isotopic concentrations were spread much more. (author)

  16. 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)

  17. 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)

  18. 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

  19. Reduction of waste arising as an option for improvement of waste management systems at NPPs with WWER type reactors

    International Nuclear Information System (INIS)

    Dultchenko, A.; Mikolaitchouk, H.

    1995-01-01

    After the USSR breakdown Ukraine inherited five NPPs with 12 WWER type reactor units and 4 RBMK type reactor units and no selected disposal site for NPP operational waste and just a few waste treatment facilities which had not been licensed or certified and could not be considered as complying safety requirements and NPP needs. At the same time the lack of competent designer organizations in Ukraine and the overall economical situation including the payment crisis resulted in significant delays in the development of radioactive waste management infrastructure and brought to the foreground a reduction of waste arisings and implementation of waste recycling technologies. In order to evaluate efficiency of waste management systems at Ukrainian NPPs in comparison with current practices at western NPPs and fix main deficiencies and optimum upgrading measures the comparative analyses of waste management systems at Ukrainian NPPs was initiated within the R and D program supported by the Ukrainian State Committee for Nuclear and Radiation Safety (UkrSCNRS). In carrying out the analyses the results of IAEA Technical Assistance Regional project on Advice on Waste Management at WWER type Reactors were used. Taking into account an influence of the Chernobyl accident consequences on the waste management system of Chernobyl NPP the case of Chernobyl NPP was set apart and cannot be considered typical so the authors confine their analysis to the WWER type reactors. For the purposes of comparison the related information about Kozlodui, Paks, Loviisa and Russian NPPs provided under the above-mentioned IAEA Regional Project was used

  20. Design basis and design features of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia)

    International Nuclear Information System (INIS)

    1994-05-01

    The prime objective of the IAEA Technical Co-operation Project on Evaluation of Safety Aspects of WWER-440 model 213 NPPs is to co-ordinate and to integrate assistance to national organizations in studying selected aspects of safety for the same type of reactors. Consequently, the study integrated the results generated by national activities carried out in the Czech Republic, Hungary, Slovakia and Ukraine and co-ordinated through the IAEA. Valuable assistance in carrying out the tasks was also provided by Bulgaria and Poland. A set of publications is being prepared to present the results of the project. The publications are intended to facilitate the review and utilization of the results of the project. They are also providing assistance in further refinement and/or extension of plant specific safety evaluation of model 213 NPPs. This Technical Document addressing the design basis and safety related design features of WWER-440 model 213 plants is the first of the series to be published. It is hoped that this document will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, 36 figs, tabs

  1. 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

  2. Decommissioning costs of WWER-440 nuclear power plants. Interim report: Data collection and preliminary evaluations

    International Nuclear Information System (INIS)

    2002-11-01

    Based on the interest in decommissioning costs within Member States, especially in WWER- 440 operating countries that face the complex decision about continued operation vs. decommissioning in the near future, the IAEA launched the task to prepare a technical document on decommissioning costs of WWER-440 nuclear power plants. The main objectives of this publication were to present the decommissioning costs of WWER-440 NPPs in a uniform manner, i.e. using the cost item and cost group system of the Interim Technical Document on Nuclear Decommissioning 'A Proposed Standardised List of Items for Costing Purposes' developed jointly by the EC, the IAEA and the OECD Nuclear Energy Agency (NEA), and providing, as such, a basis for understanding decommissioning costs differences. Member States operating WWER-440 NPPs or having such units under shutdown or even under decommissioning conditions have been requested to provide cost estimates and other input data in order to facilitate understanding of their cost figures. Both decommissioning options, i.e. immediate decommissioning and safe enclosure, have been considered. In the aforementioned joint Interim Technical Document, cost items related to activities that are carried out with a similar emphasis, whether or not tied to a similar time schedule for decommissioning, or that are based on overall activities that cannot be categorised in a specific time period, are grouped as follows: pre-decommissioning actions; facility shutdown activities; procurement of general equipment and material; dismantling activities; waste processing, storage and disposal; site security, surveillance and maintenance; site restoration, cleanup and landscaping; project management, engineering and site support; research and development; fuel and nuclear material; other costs. Before starting implementation of the study, agreement was obtained on general financial, technical and social boundary conditions that should be used in order to facilitate

  3. Verification of thermal-hydraulic computer codes against standard problems for WWER reflooding

    International Nuclear Information System (INIS)

    Alexander D Efanov; Vladimir N Vinogradov; Victor V Sergeev; Oleg A Sudnitsyn

    2005-01-01

    Full text of publication follows: The computational assessment of reactor core components behavior under accident conditions is impossible without knowledge of the thermal-hydraulic processes occurring in this case. The adequacy of the results obtained using the computer codes to the real processes is verified by carrying out a number of standard problems. In 2000-2003, the fulfillment of three Russian standard problems on WWER core reflooding was arranged using the experiments on full-height electrically heated WWER 37-rod bundle model cooldown in regimes of bottom (SP-1), top (SP-2) and combined (SP-3) reflooding. The representatives from the eight MINATOM's organizations took part in this work, in the course of which the 'blind' and posttest calculations were performed using various versions of the RELAP5, ATHLET, CATHARE, COBRA-TF, TRAP, KORSAR computer codes. The paper presents a brief description of the test facility, test section, test scenarios and conditions as well as the basic results of computational analysis of the experiments. The analysis of the test data revealed a significantly non-one-dimensional nature of cooldown and rewetting of heater rods heated up to a high temperature in a model bundle. This was most pronounced at top and combined reflooding. The verification of the model reflooding computer codes showed that most of computer codes fairly predict the peak rod temperature and the time of bundle cooldown. The exception is provided by the results of calculations with the ATHLET and CATHARE codes. The nature and rate of rewetting front advance in the lower half of the bundle are fairly predicted practically by all computer codes. The disagreement between the calculations and experimental results for the upper half of the bundle is caused by the difficulties of computational simulation of multidimensional effects by 1-D computer codes. In this regard, a quasi-two-dimensional computer code COBRA-TF offers certain advantages. Overall, the closest

  4. Mechanical properties and microstructure of long term thermal aged WWER 440 RPV steel

    Energy Technology Data Exchange (ETDEWEB)

    Kolluri, M., E-mail: kolluri@nrg.eu [Nuclear Research & Consultancy Group (NRG), P.O. Box 25, 1755 ZG Petten (Netherlands); Kryukov, A. [Scientific and Engineering Centre for Nuclear and Radiation Safety, 107140 Moscow (Russian Federation); Magielsen, A.J. [Nuclear Research & Consultancy Group (NRG), P.O. Box 25, 1755 ZG Petten (Netherlands); Hähner, P. [European Commission, Joint Research Centre, Directorate G – Nuclear Safety and Security, P.O. Box 2, 1755 ZG Petten (Netherlands); Petrosyan, V. [Armenian Scientific Research Institute for Nuclear Plant Operation (ARMATOM), 0027 Yerevan (Armenia); Sevikyan, G. [Armenian Nuclear Power Plant (ANPP), 0911, Metsamor, Armavir Marz (Armenia); Szaraz, Z. [European Commission, Joint Research Centre, Directorate G – Nuclear Safety and Security, P.O. Box 2, 1755 ZG Petten (Netherlands)

    2017-04-01

    The integrity assessment of the Reactor Pressure Vessel (RPV) is essential for the safe and Long Term Operation (LTO) of a Nuclear Power Plant (NPP). Hardening and embrittlement of RPV caused by neutron irradiation and thermal ageing are main reasons for mechanical properties degradation during the operation of an NPP. The thermal ageing-induced degradation of RPV steels becomes more significant with extended operational lives of NPPs. Consequently, the evaluation of thermal ageing effects is important for the structural integrity assessments required for the lifetime extension of NPPs. As a part of NRG's research programme on Structural Materials for safe-LTO of Light Water Reactor (LWR) RPVs, WWER-440 surveillance specimens, which have been thermal aged for 27 years (∼200,000 h) at 290 °C in a surveillance channel of Armenian-NPP, are investigated. Results from the mechanical and microstructural examination of these thermal aged specimens are presented in this article. The results indicate the absence of significant long term thermal ageing effect of 15Cr2MoV-A steel. No age hardening was detected in aged tensile specimens compared with the as-received condition. There is no difference between the impact properties of as-received and thermal aged weld metals. The upper shelf energy of the aged steel remains the same as for the as-received material at a rather high level of about 120 J. The T{sub 41} value did not change and was found to be about 10 °C. The microstructure of thermal aged weld, consisting carbides, carbonitrides and manganese-silicon inclusions, did not change significantly compared to as-received state. Grain-boundary segregation of phosphorus in long term aged weld is not significant either which has been confirmed by the absence of intergranular fracture increase in the weld. Negligible hardening and embrittlement observed after such long term thermal ageing is attributed to the optimum chemical composition of 15Cr2MoV-A for high

  5. Implementation of safety parameter display system on Russian NPPs with WWER reactors

    International Nuclear Information System (INIS)

    Dounaev, V.G.; Neboyan, V.T.

    1996-01-01

    This report gives a short overview of the status of safety parameter display systems (SPDS) implementation on Russian NPPs with WWER reactors and also discusses the SPDS, which is being developed for Kalinin NPP. The assessment of the safety status of the plant is done by the continuous monitoring of six critical safety functions and the corresponding status trees. Besides, a number of additional functions are realized within the scope of KlnNPP, aimed at providing the operator and the safety engineer in the main control room with more detailed information in accidental situation as well as during the normal operation. In particular, these functions are: archiving, data logs and alarm handling, safety actions monitoring, mnemonic diagrams indicating the state of main technological equipment and basic plant parameters, reference data, etc. Also, the operator support function ''computerized procedures'' is included in the scope of SPDS. The basic SPDS implementation platform is ADACS of SEMA GROUP design. The system architecture includes two workstations in the main control room: one is for reactor operator and the other one for safety engineer. Every station has two CRT screens which ensures computerized procedures implementation and provides for extra services for the operator. Also, the information from the SPDS is transmitted to the local crisis centre and to the crisis centre of the State utility organization concern ''Rosenergoatom''. (author). 3 refs

  6. EddyOne automated analysis of PWR/WWER steam generator tubes eddy current data

    International Nuclear Information System (INIS)

    Nadinic, B.; Vanjak, Z.

    2004-01-01

    INETEC Institute for Nuclear Technology developed software package called Eddy One which has option of automated analysis of bobbin coil eddy current data. During its development and on site use, many valuable lessons were learned which are described in this article. In accordance with previous, the following topics are covered: General requirements for automated analysis of bobbin coil eddy current data; Main approaches to automated analysis; Multi rule algorithms for data screening; Landmark detection algorithms as prerequisite for automated analysis (threshold algorithms and algorithms based on neural network principles); Field experience with Eddy One software; Development directions (use of artificial intelligence with self learning abilities for indication detection and sizing); Automated analysis software qualification; Conclusions. Special emphasis is given on results obtained on different types of steam generators, condensers and heat exchangers. Such results are then compared with results obtained by other automated software vendors giving clear advantage to INETEC approach. It has to be pointed out that INETEC field experience was collected also on WWER steam generators what is for now unique experience.(author)

  7. 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

  8. Bubble-vacuum system of accident localization of reference nuclear power plant with two WWER's

    International Nuclear Information System (INIS)

    Sykora, D.; Sykorova, I.

    1988-01-01

    Higher efficiency of the safety system for removing the consequences of project design accidents and higher radiation safety of a nuclear power plant with two WWER-440 units is the subject of Czechoslovak patent document 243961. The principle consists in interconnecting air chambers which are the end parts of safety systems for the two units. The air chamber is separated from the other parts of the safety system by double swing-check valves or closures. The connecting pipes of the two air chambers do not in any way reduce the reliability of the safety system thanks to their high technical safety and totally passive function. The benefits of the interconnection of the air chambers are given by the fact that it reduces maximum accident overpressure both in the air chambers and in the airtight zones. The reduction of the overpressure reduces the total leakage of radioactive substances and the radiation burden of the environment in case of a nuclear power plant accident. (Z.M.). 2 figs

  9. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kocik, J.; Keilova, E.

    1993-01-01

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

  10. Safeguarding the nuclear safety of WWER-440 reactor pressure vessels at SKODA Plzen

    International Nuclear Information System (INIS)

    Hrbek, Z.

    1986-01-01

    The approach is described of the SKODA enterprise to safety assurance and to providing the reliability of WWER-440 reactor pressure vessels. The philosophy is analyzed of in-service inspection and determination of the residual service life of pressure vessels. This follows up on the so-called conception of basic safety whose main aim is to preclude failures at production stage by the selection of suitable material, namely by optimizing the choice of raw materials, of metallurgical procedures such as will lead to high purity of the pressure vessel material, by introducing multiple inspection in production, reducing the sensitivity of materials to technological operations, and by high-quality welds. The quality of in-service inspections is given by the use of technical diagnostic instruments of peak quality and of modern methods of nondestructive materials testing. The instruments and methods used are described. It is stated that the experience gained with in-service inspection will make it possible to draw up operating regulations and safety criteria for nuclear installations and own inspection regulations, this with regard to technical and economic factors. (Z.M.)

  11. Mounting of localization shaft by enlarged structures at the NPP with WWER-440

    International Nuclear Information System (INIS)

    Naumenko, S.V.

    1982-01-01

    A technique of mounting of a localization system at the WWER-440 NPP is described. The localization system consists of air-lift devices located in pressurized building (shaft) 12.6 thousand m 3 volume. Air-lift devices are placed in 12 bayers with 3.37 m spacing over the height of localization shaft. Every layer of air-lift devices consists of 18 supporting H-beams number 60 of 8.5 m in length. The total host of air-lift devices and metal works of servicing platforms is equal to 725 t. The air-lift device consists of the large number of details (660 pieces of 500-2500 kg mass and above 2500 pieces of 500 kg mass), which causes the necessity of accomplishing a large volume of assembling and welding works. To reduce the labour content in the mounting zone and volume of work accomplished at the height the method of large-structure mounting of air-lift devices was suggested. The method lies in ground assembly of air-lift structures on the basis of several supporting beams and their following lifting to the corresponding layer. The large-structure mounting of localization shaft enables to reduce by 25-30% the length of joint welds made during the mounting as well as the volume of transport and cordage works; to reduce the time of building crane usage and 1.5-1.7 times the total periods of works

  12. 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

  13. 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

  14. 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

  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. 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)

  17. Seismic assessment of Kozloduy WWER-440, model 230 nuclear power plant

    International Nuclear Information System (INIS)

    Monette, P.; Baltus, R.; Yanev, P.; Campbell, R.

    1993-01-01

    A preliminary report is given of the findings of an IAEA sponsored walkdown of the WWER-440, model 230 at Kozloduy, in May 1991. The scope of the IAEA mission was to determine the lower bound seismic capacity of the plant and to make recommendations for improvements to increase the earthquake resistance. The methodology utilized in the assessment is that used for evaluation of the seismic margin in U.S. nuclear power plants subjected to earthquakes beyond their design basis. Included in the assessment is the establishment of a safe shutdown path which would include the capacity to mitigate a small break in the primary system, performance of a walkdown of the safe shutdown path and calculation of the high-confidence-of-low-probability-of-failure (HCLPF) of the safe shutdown path. Excluding system design deficiency relative to U.S. and Western Europe standards, it was found that the plant has many seismic vulnerabilities similar to those that existed in many of the U.S. plants prior to about 1979 when the Systematic Evaluation Program was initiated. (Z.S.) 1 tab., 1 fig

  18. Criteria for seismic evaluation and potential design fixes for WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-01-01

    The purpose for this document is to provide a criteria for the seismic evaluation and development of potential design fixes for structures, systems and components for the WWER type Nuclear power plants. The design fixes are divided into two categories, detailed and easy fixes. Detailed fixes are typically applicable to building structures, components for which there is little or no seismic capacity information, large tanks and vital systems and components which make up the reactor cooling system and components which perform support or auxiliary functions. In case of the design of 'easy fixes', the criteria presented may be used for both the seismic design as well as for the evaluation of structures, systems and components to which easy fix design applies. Easy fixes are situations where seismic capacities of structures, systems and components can be significantly increased with relatively minor, inexpensive fixes usually associated with anchorage modification of safety related structures, systems and components or those that could interact with safety related structures, systems and components. Often these fixes can be accomplished while the plant is in operation

  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. WWER440 few group data library preparation and its application in MOBY-DICK modular system

    International Nuclear Information System (INIS)

    Krysl, V.; Mikolas, P.; Svarny, J.

    2002-01-01

    Paper provides summary of methodology of few-group library preparation with emphasis on new features of assemblies, like fuel assemblies with Gd burnable poison or Control Fuel Assembly with Hf plates. Special attention is devoted to the transient part (coupler) of Control Fuel Assembly from the point of view of boundary conditions preparation. Based of this methodology prepared library is implemented into macro code for different number of axial meshes for both coarse and fine mesh diffusion calculations. Problems with local power peaking calculations in WWER-440 cores are closely connected with the correct modeling of the power perturbations in the neighbourhood of Control Fuel Assembly coupler. The new version of MOBY-DICK provided with the new few group data library can assess most of effects induced by insertion of Control Fuel Assembly in the core including effects of newly designed Hf plates in the coupler of Control Fuel Assembly. Compatibility of transport and diffusion calculation was taken into account in the analysis of Control Fuel Assembly movement without/with Hf plates (Authors)

  1. 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)

  2. Kozloduy NPP WWER-440/230 reactor pressure vessel radiation lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Vodenicharov, S [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. po Metaloznanie i Tekhnologiya na Metalite; Kamenova, Tz [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. po Metaloznanie i Tekhnologiya na Metalite; Tzokov, P; Videnov, A [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Pekov, B [Committee on the Use of Atomic Energy for Peaceful Purposes, Sofia (Bulgaria)

    1996-12-31

    The processes of metal embrittlement induced by neutron irradiation embrittlement (NIE) and neutron re-irradiation embrittlement (NRE) of the rector pressure vessel (RPV) are investigated. Radiation lifetime is calculated using two approaches for re-embrittlement: a conservative law and a lateral shift of the critical transition temperature curve after neutron irradiation. In order to prevent NIE the following measures have been taken at the Kozloduy NPP: loading of dummy elements into core periphery; heating the water for emergency core cooling to 55{sup o} C; fast acting valves in the main steam piping, etc. The critical embrittlement temperature, the residual part of temperature shift and the radiation lifetime have been calculated for units 1 - 4 using the two approaches and updated information on P and Cu impurities content. It is concluded that if the lateral re-embrittlement law is adopted and the P content does not exceed 0.05%, all RPV should reach their design lifetime. The NIE in WWER-440/230 is related to C and P content in the weld 4 and is negligible for the Unit 4 in particular, which has low impurities content. In order to reach the design lifetime of the Units 1 - 3 it is necessary to install MSIV. A verification of chemical composition of the Unit 1 RPV weld 4 metal is recommended. 7 refs., 3 figs., 6 tabs.

  3. Assesment criteria improvement for operational economical efficiency of NPP with WWER typr reactors

    International Nuclear Information System (INIS)

    Abagyan, A.A.; Matveev, A.A.; Ignatenko, E.I.; Pshechenkova, T.V.

    1983-01-01

    A new technique for calculating fuel component of the cost of NPP electric power generation is suggested. To calculate the variable part of fuel component it is suggested to consider the acquisition cost of fuel assemblies, unloaded from the reactor on fuel cycle completion instead of acquisition cost of fresh fuel assemblies, loaded into the reactor for organization of this cycle; it is also suggested to include the acquisition cost of fuel assemblies, remaining in reactop core on completion of the last fuel cycle+ in constant fuel expenses. The fuel component of the cost of WWER-440 reactor electric power generation for desigh operating conditions with 700 full power days period of steady-state fuel cycle was calculated. The suggested technique enables to reveal the deviation of the real fuel cycling conditions from the standard ones and calculate the value of this deviation, establish the reasons, disturbing the economical conditions of reactor operation, to approximate the real conditions of fuel cycling to the optimal ones by influencing on technological process, resulting in the change of factors, determining the fuel cycling efficiency during electric power generation and refueling in considered cycle

  4. 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)

  5. Kozloduy NPP WWER-440/230 reactor pressure vessel radiation lifetime

    International Nuclear Information System (INIS)

    Vodenicharov, S.; Pekov, B.

    1995-01-01

    The processes of metal embrittlement induced by neutron irradiation embrittlement (NIE) and neutron re-irradiation embrittlement (NRE) of the rector pressure vessel (RPV) are investigated. Radiation lifetime is calculated using two approaches for re-embrittlement: a conservative law and a lateral shift of the critical transition temperature curve after neutron irradiation. In order to prevent NIE the following measures have been taken at the Kozloduy NPP: loading of dummy elements into core periphery; heating the water for emergency core cooling to 55 o C; fast acting valves in the main steam piping, etc. The critical embrittlement temperature, the residual part of temperature shift and the radiation lifetime have been calculated for units 1 - 4 using the two approaches and updated information on P and Cu impurities content. It is concluded that if the lateral re-embrittlement law is adopted and the P content does not exceed 0.05%, all RPV should reach their design lifetime. The NIE in WWER-440/230 is related to C and P content in the weld 4 and is negligible for the Unit 4 in particular, which has low impurities content. In order to reach the design lifetime of the Units 1 - 3 it is necessary to install MSIV. A verification of chemical composition of the Unit 1 RPV weld 4 metal is recommended. 7 refs., 3 figs., 6 tabs

  6. Characterisation of the inventory of radioisotopes induced in the biological shield a WWER-440 reactor

    International Nuclear Information System (INIS)

    Feher, S.; Czifrus, Sz.; Zsolnay, E.M.; Szondi, E.

    2001-01-01

    A significant part of the radwaste originating from the decommissioning of NPPs is made up of the activated concrete and steel components of the biological shield. The paper presents the results of studies aimed at the determination of the amount of radionuclides accumulating in the serpentinous and ordinary concrete shield around the WWER-440 reactors of the Paks NPP. For the calculations, the reactor, vessel and shield were modelled in detail both in terms of geometry and material composition. The spatial and energy distribution of the activating neutron spectrum was determined by certain modules of SCALE 4.3 and the code TORT in two and three dimensions, while the activation was calculated using ORIGEN-S for 22 geometrical regions. The results showed that the activity of the concrete structures at final shutdown after 30 years of operation is approximately 50 TBq, which decreases to 20, 12, 1.1 TBq and 27 GBq after 1 month, 1 year, 10 and 100 years, respectively (Authors)

  7. 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

  8. Effect of increased fuel exploitation on the main characteristics of spent WWER 440 fuel

    International Nuclear Information System (INIS)

    Zib, A.

    2001-01-01

    The article deals with the effect of a higher fuel exploitation on the main characteristics (particularly radioactivity and decay heat power) of spent WWER 440 fuel. The main characteristics were calculated by using the Origen code. The study was implemented as a three-stage process. In the first stage, the radioactivity and residual thermal power time evolution values were calculated for the 'typical fuel', i. e. fuel assembly with initial enrichment of 3.6% U-235, 3 years in reactor, and burnup of 30 MWd/kg U. In the second stage, ceteris paribus radioactivity and thermal power analyses of sensitivity to changes in the fuel burnup, initial fuel enrichment, and time in reactor were carried out for the typical fuel assembly. In the third stage, the effect of changes in all three variables was investigated for fuel assemblies possessing parameters that approach those applied at the Dukovany NPP. The effect of a higher fuel exploitation on the interim fuel storage is also mentioned. (author)

  9. 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)

  10. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.

  11. A strategic approach to short- and long-term irradiated WWER fuel management

    International Nuclear Information System (INIS)

    Wilcox, P.; Conboy, T.M.

    1994-01-01

    A methodology is presented for comparison of alternative options for short-term and long-term irradiated fuel management. The value of this methodology is that all interested parties can take part in the analysis and derive the basis for decision-making. The methodology can answer questions as: When can uranium and plutonium recovered by reprocessing be recycled cost effectively in WWER? If reprocessing is not the short-term option chosen, is storage of irradiated fuel at the original licensed nuclear reactor site preferable to a separate storage-only-site? Are modular vault dry stores and cooling ponds which necessitate significant capital investment prior to deployment, more costly overall than their options? Should the most suitable form of irradiated fuel management be determined only by cost constraints? The key stages of the methodology are: 1) Assessing the current situation; 2) Assessing priorities; 3) Option/possible solutions; 4) Generic storage systems; 5) Cost/funding analysis; 6) Selection criteria; 7) Optioneering/evaluation. The conclusions that can be reached from this methodological approach lead to firm recommendations based on objective assessments. The methodology builds on existing expertise. It is not an imposed solution allowing an excellent exchange of knowledge and skills between the people involved

  12. Thermal-hydraulic analysis of total loss of steam generator feed water in WWER-440

    International Nuclear Information System (INIS)

    Sabotinov, L.; Cadet-Mercier, S.

    2001-01-01

    The analysis is carried out for a WWER-440/V270 with upgraded primary safety valves (replacement of the existing PRZ safety valves with Pilot Operated Relief Valves (PORV) of the type SEBIM (France)) The current analysis is focused on the scenario 'Total Loss of SGs Feed Water' with application of the operator action of primary system 'Feed and Bleed' in order to check the effectiveness of the installed pressurizer SEBIM valves and to verify that the operator can cool down the reactor system and cope with this accident. The calculations have been performed at the Institute of Protection and Nuclear Safety (IPSN) in Fontenay-aux-Roses with the computer code CATHARE 2 Version 1.3L1. CATHARE is a French best estimate thermal-hydraulic program for accident analysis in the light water nuclear reactors, developed with the participation of the IPSN (Institut de Protection et Surete Nucleaire), CEA (Commissariat a l'Energie Atomique), Framatome and EdF (Electricite de France). (author)

  13. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kupca, L [VUJE Inc. (Slovakia)

    2012-07-01

    In the Slovak Republic are under operation 6 units (4 in the Jaslovske Bohunice site, and 2 in the Mochovce), 2 units are under construction in Mochovce site. All units are WWER-440 type. The fresh fuel is imported from the Russian Federation. The spent fuel assemblies are stored in wet conditions in Bohunice Interim Storage Spent Fuel Facility (SFIS). By 15 July 2008, there were 8413 assemblies in SFIS. The objectives are: 1) Wet AR storage of spent fuel from the NPP Bohunice and Mochovce: Surveillance of conditions for spent fuel storage in the at-reactor (AR) storage pools of both NPP's (characteristics of pool water, corrosion product data); Visual control of storage pool components; Evaluation of storage conditions with respect to long-term stability (corrosion of fuel cladding, structural materials); 2) Wet SFIS storage at Bohunice: Measurement of spent fuel conditions during the long-term wet storage, activity data in the storage casks and amount of crud; Surveillance program for SFIS structural materials.

  14. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Takats, F [TS Enercon Kft. (Hungary)

    2012-07-01

    Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. The fresh fuel is imported from Russia so far. The spent fuel assemblies were shipped back to Russia until 1997 after about 6 years cooling at the plant. A dry storage facility (MVDS type) has been constructed and is operational since then. By 1 January 2008, there were 5107 assemblies in dry storage. The objectives are: 1) Wet AR storage of spent fuel from the NPP Paks: Measurements of conditions for spent fuel storage in the at-reactor (AR) storage pools of Paks NPP (physical and chemical characteristics of pool water, corrosion product data); Measurements and visual control of storage pool component characteristics; Evaluation of storage characteristics and conditions with respect to long-term stability (corrosion of fuel cladding, construction materials); 2) Dry AFR storage at Paks NPP: Calculation and measurement of spent fuel conditions during the transfer from the storage pool to the modular vault dry storage (MVDS) on the site; Calculation and measurement of spent fuel conditions during the preparation of fuel for dry storage (drying process), such as crud release, activity build-up; Measurement of spent fuel conditions during the long-term dry storage, activity data in the storage tubes and amount of crud.

  15. Analysis of some fuel characteristics deviations and their influence over WWER-440 fuel cycle design

    International Nuclear Information System (INIS)

    Stoyanova, I.; Kamenov, K.

    2001-01-01

    The aim of this study is to estimate the influence of some deviations in WWER-440 fuel assemblies (FA) characteristics upon fuel core design. A large number of different fresh fuel assemblies with enrichment of 3.5 t % are examined related to the enrichment, mass of initial metal Uranium and assembly shroud thickness. Infinite multiplication factor (Kinf) in fuel assembly has been calculated by HELIOS spectral code for basic assembly and for different FA with deviation of a single parameter. The effects from single parameter deviation (enrichment) and from two parameter deviations (enrichment and wall thickness) on the neutron-physics characteristics of the core are estimated for different fuel assemblies. Relatively week burnup dependence on Kinf is observed as result of deviation in the enrichment of the fuel and in the wall thickness of the assembly. An assessment of a FA single and two parameter deviations effects on design fuel cycle duration and relative power peaking factor is also considers in the paper. As a final conclusion can be settled that the maximum relative shortness of fuel cycle can be observed in the case of two FA parameters deviations

  16. Criteria for seismic evaluation and potential design fixes for WWER type nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Stevenson, J D [Stevenson and Associates, Cleveland, OH (United States)

    1995-07-01

    The purpose for this document is to provide a criteria for the seismic evaluation and development of potential design fixes for structures, systems and components for the WWER type Nuclear power plants. The design fixes are divided into two categories, detailed and easy fixes. Detailed fixes are typically applicable to building structures, componentsfor which there is little or no seismic capacity information, large tanks and vital systems and components which make up the reactor cooling system and components which perform support or auxiliary functions. In case of the design of 'easy fixes', the criteria presented may be used for both the seismic design as well as for the evaluation of structures, systems and components to which easy fix design applies. Easy fixes are situations where seismic capacities of structures, systems and components can be significantly increased with relatively minor, inexpensive fixes usually associated with anchorage modification of safety related structures, systems and components or those that could interact with safety related structures, systems and components. Often these fixes can be accomplished while the plant is in operation.

  17. Radioactive 55Fe contamination in the primary circuit of WWER-440

    International Nuclear Information System (INIS)

    Ruskov, T.; Ruskov, R.; Dobrevski, I.; Konnova, S.; Zaharieva, N.; Menut, P.

    2001-01-01

    The isotope 55 Fe generation in the steel construction materials of the reactor and the mechanism of internal irradiation and blood affection by the 55 Fe are briefly discussed in this paper. The paper also presents the results from calculation of direct generation of 55 Fe due to neutron irradiation of different iron-contained parts of the reactor system such as the steel shell of the reactor core, the core basket, the steel shaft of the reactor vessels. Calculations are performed with specially developed program code DIRGEN. Another type of contamination, considered in the paper is due to the corrosion of materials and erosion-dissolution processes in the primary circuit of WWER with their subsequent deposition-precipitation on the inner surface of the primary circuit. The real time calculations of the 55 Fe activity are performed, by using of the updated computer code MIGA-RT. The obtained results show that the 55 Fe activity deposition on the inner surfaces of the primary circuits reaches the values of 103 kBq/cm 2 for the reactor core surfaces and 102 kBq/cm 2 for the out-of-core surfaces. The activity values are in one order of magnitude higher than the corresponding activity values due to 60 Co buildup

  18. Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

    International Nuclear Information System (INIS)

    Boros, I.; Aszodi, A.

    2003-01-01

    The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

  19. Optimization of steam generators of NPP with WWER in operation with variable load

    Science.gov (United States)

    Parchevskii, V. M.; Shchederkina, T. E.; Gur'yanova, V. V.

    2017-11-01

    The report addresses the issue of the optimal water level in the horizontal steam generators of NPP with WWER. On the one hand, the level needs to be kept at the lower limit of the allowable range, as gravity separation, steam will have the least humidity and the turbine will operate with higher efficiency. On the other hand, the higher the level, the greater the supply of water in the steam generator, and therefore the higher the security level of the unit, because when accidents involving loss of cooling of the reactor core, the water in the steam generators, can be used for cooling. To quantitatively compare the damage from higher level to the benefit of improving the safety was assessed of the cost of one cubic meter of water in the steam generators, the formulated objective function of optimal levels control. This was used two-dimensional separation characteristics of steam generators. It is demonstrated that the security significantly shifts the optimal values of the levels toward the higher values, and this bias is greater the lower the load unit.

  20. Results of calculation of WWER-440 fuel rods (Kol`skaya-3 NPP) at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Scheglov, A; Proselkov, V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Panin, M; Pitkin, Yu [Kol` skaya NPP, (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation)

    1994-12-31

    Thermal-physical characteristics of fuel rods of two fuel assemblies which were operated within 5 - 8 and 5 - 9 core fuel loadings of the Unit 3 of the Kol`skaya NPP are calculated. They have achieved deep burnup during 4-year (> 46 Mwd/kg U) and 5-year (> 48 Mwd/kg U) fuel cycle. Fuel assemblies have been unloaded off the reactor and subjected to a post-irradiation testing. PIN-mod2 code originally designed for modelling of WWER fuel rod behaviour in a quasi-steady-state operation is used. The average fuel rod in the fuel assembly and the fuel rod with maximum burnup are selected. The preliminary comparison of the calculation results with those of the post-irradiation examination shows a satisfactory agreement. On the basis of the results obtained in the post-irradiation experiments an improvement of the model for calculation of fission gas release and creep of the cladding is planned. The results of the analysis performed indicate that the fuel rod completely preserves its working ability; fuel temperature does not exceed 1300{sup o} C; fission gas release does not exceed 4%; maximum gas pressure inside the cladding at the end of campaign does not exceed 2 MPa. 2 tabs., 11 figs., 5 refs.